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 Nuclear Regulatory Commission 10 CFR Parts 1, 2, 10, 11, 19, 20, 21, 25, 26, 30, 40, 50, 51, 53, 70, 72, 73, 74, 75, 95, 140, 150, 170, and 171
[NRC-2019-0062]
RIN 3150-AK31
 AGENCY: Nuclear Regulatory Commission.
 ACTION: Proposed rule.
 SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is proposing to revise the NRC's regulations by adding a risk-informed, performance-based, and technology-inclusive regulatory framework for commercial nuclear plants in response to the Nuclear Energy Innovation and Modernization Act (NEIMA). The NRC plans to hold a public meeting to promote full understanding of the proposed rule and facilitate public comments.
 DATES: Submit comments by December 30, 2024. Comments received after this date will be considered if it is practical to do so, but the NRC is able to ensure consideration only for comments received before this date.
 ADDRESSES: You may submit comments by any of the following methods however, the NRC encourages electronic comment submission through the Federal rulemaking website: *Federal Rulemaking website:*
                             Go to 
                            *https://www.regulations.gov*
                             and search for Docket ID NRC-2019-0062. Address questions about NRC dockets to Helen Chang; telephone: 301-415-3228; email: 
                            *Helen.Chang@nrc.gov.*
                             For technical questions contact the individuals listed in the 
                            **FOR FURTHER INFORMATION CONTACT**
                             section of this document.
                        *Email comments to: Rulemaking.Comments@nrc.gov.*
                             If you do not receive an automatic email reply confirming receipt, then contact us at 301-415-1677.
                        *Fax comments to:*
                             Secretary, U.S. Nuclear Regulatory Commission at 301-415-1101.
                        *Mail comments to:*
                             Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, ATTN: Rulemakings and Adjudications Staff.
                        *Hand deliver comments to:*
                             11555 Rockville Pike, Rockville, Maryland 20852, between 7:30 a.m. and 4:15 p.m. eastern time, Federal workdays; telephone: 301-415-1677. You can read a plain language description of this proposed rule at 
                            *https://www.regulations.gov/​docket/​NRC-2019-0062.*
                             For additional direction on obtaining information and submitting comments, see “Obtaining Information and Submitting Comments” in the 
                            **SUPPLEMENTARY INFORMATION**
                             section of this document.
 FOR FURTHER INFORMATION CONTACT: Robert Beall, Office of Nuclear Material Safety and Safeguards, telephone: 301-415-3874; email: 
                            *Robert.Beall@nrc.gov;*
                             or Anders Gilbertson, Office of Nuclear Reactor Regulation, telephone: 301-415-1541; email: 
                            *Anders.Gilbertson@nrc.gov.*
                             Both are staff of the U.S. NRC, Washington, DC 20555-0001.


SUPPLEMENTARY INFORMATION:  Executive Summary A. Need for the Regulatory Action On January 14, 2019, the President signed the Nuclear Energy Innovation and Modernization Act (NEIMA) into law (Pub. L. 115-439). NEIMA section 103(a)(4) directs the NRC to “complete a rulemaking to establish a technology-inclusive, regulatory framework for optional use by commercial advanced nuclear reactor applicants for new reactor license applications.” NEIMA defines a “technology-inclusive regulatory framework” as one that is “developed using methods of evaluation that are flexible and practicable for application to a variety of reactor technologies, including, where appropriate, the use of risk-informed and performance-based techniques.” NEIMA, as further amended by the Accelerating Deployment of Versatile, Advanced Nuclear for Clean Energy Act of 2024 (ADVANCE Act), defines the term “advanced nuclear reactor” as “a nuclear fission reactor or fusion machine, including a prototype plant (as defined in sections 50.2 and 52.1 of title 10, 
                        *Code of Federal Regulations*
                         (as in effect on the date of enactment of [NEIMA])), with significant improvements compared to commercial nuclear reactors under construction as of the date of enactment of [NEIMA].” The NRC initially considered establishing the scope of proposed part 53, “Risk-Informed, Technology-Inclusive Regulatory Framework for Commercial Nuclear Plants,” of title 10 of the 
                        *Code of Federal Regulations*
                         (10 CFR) as being for “advanced nuclear plants” consisting of one or more “advanced nuclear reactors” as defined in NEIMA. Based on public discussions on the use of the term, the NRC determined that the NEIMA definition, although broad, did not define “significant improvements” with enough specificity to implement in NRC regulations. Additionally, a number of stakeholders suggested that the descriptor, “advanced,” implied enhanced safety, while the NEIMA definition includes “significant improvements” in areas other than safety enhancements. In response to this feedback, and to be technology inclusive, the NRC determined that the broader term “commercial nuclear plant” would be preferable. The current application and licensing requirements in 10 CFR part 50, “Domestic Licensing of Production and Utilization Facilities,” and 10 CFR part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” were primarily developed to address license requests concerning water-cooled reactors, and to address operational requirements for those types of reactors. This proposed rule responds to NEIMA by creating an alternative regulatory framework for licensing future commercial nuclear plants. The new alternative requirements and implementing guidance would adopt technology-inclusive approaches and use risk-informed and performance-based techniques to ensure an equivalent level of safety to that of operating commercial nuclear plants while providing flexibility for licensing and regulating a variety of technologies and designs for commercial nuclear reactors. B. Major Provisions Major provisions of this proposed rule, supported by accompanying guidance, include the following: A new alternative technology-inclusive, risk-informed, performance-based framework that includes requirements for licensing and regulating nuclear plants during the various stages of their life cycles.
                    A new alternative technology-inclusive, risk-informed, and performance-based framework in10 CFR part 26, “Fitness for Duty Programs,” developed from existing requirements in subpart K, “FFD Programs for Construction,” of part 26.
                    A new alternative technology-inclusive and performance-based security framework in10 CFR part 73, “Physical Protection of Plants and Materials,” that includes requirements for protection of licensed activities at commercial nuclear plants. C. Costs and Benefits The NRC prepared a draft regulatory analysis to determine the expected quantitative costs and benefits of this proposed rule and associated guidance as well as qualitative factors to be considered in the NRC's rulemaking decision. The conclusion from the 
                        ( print page 86919)
                        analysis is that this proposed rule and associated guidance would result in net averted costs to the industry and the NRC ranging from $53.6 million using a 7-percent discount rate to $68.2 million using a 3-percent discount rate, using an assumption of one applicant under 10 CFR part 53. As the number of applicants increases, so do the estimated averted costs. The draft regulatory analysis also considers qualitative factors, such as greater regulatory stability, predictability, and clarity to the licensing process. These benefits would result from incorporating advances in probabilistic risk assessment (PRA) and other risk-informed analyses and codifying regulatory enhancements that currently exist in regulatory guides (RGs). Another qualitative factor is promoting a performance-based regulatory framework that specifies requirements to be met and provides flexibility to an applicant or licensee regarding the information or approach needed to satisfy those requirements. For more information, please see the draft regulatory analysis (available in the NRC's Agencywide Documents Access and Management System (ADAMS) Accession No. ML21165A112). Table of Contents I. Obtaining Information and Submitting Comments A. Obtaining Information B. Submitting Comments II. Background A. NRC Advanced Reactor Readiness B. Stakeholder Views on Part 53 Preliminary Proposed Rule Language III. Discussion A. Objective and Applicability B. Need for Changes to the Existing Regulatory Framework C. 10 CFR Part 53: Framework IV. Part 53: Framework Subpart A—General Provisions A. Discussion of Definitions in Proposed Part 53 B. Other General Provisions Subpart B—Technology-Inclusive Safety Requirements Subpart C—Design and Analysis Requirements Subpart D—Siting Requirements Subpart E—Construction and Manufacturing Requirements Subpart F—Requirements for Operation Subpart G—Decommissioning Requirements Subpart H—Licenses, Certifications, and Approvals Subpart I—Maintaining and Revising Licensing Basis Information Subpart J—Reporting and Other Administrative Requirements Subpart M—Enforcement V. Changes to Other Parts of 10 CFR Chapter I 10 CFR Part 26 A. Introduction B. Proposed Changes to Part 26, Subparts A Through E and I C. Proposed Requirements for Part 26, Subpart M D. Proposed Changes to Part 26, Subpart N E. Proposed Changes to Part 26, Subpart O 10 CFR Part 50 A. Section 50.160: Emergency Preparedness for Small Modular Reactors, Non-Light-Water Reactors, and Non-Power Production or Utilization Facilities B. Appendix B to Part 50: Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants 10 CFR Part 73 A. Section 73.100: Technology-Inclusive Requirements for Physical Protection of Licensed Activities at Commercial Nuclear Plants Against Radiological Sabotage B. Section 73.110: Technology-Inclusive Requirements for Protection of Digital Computer and Communication Systems and Networks C. Section 73.120: Access Authorization Program for Commercial Nuclear Plants VI. Specific Requests for Comments VII. Section-by-Section Analysis VIII. Regulatory Flexibility Certification IX. Regulatory Analysis X. Backfitting and Issue Finality XI. Cumulative Effects of Regulation XII. Plain Writing XIII. Environmental Assessment and Proposed Finding of No Significant Environmental Impact XIV. Paperwork Reduction Act XV. Criminal Penalties XVI. Voluntary Consensus Standards XVII. Availability of Guidance XVIII. Public Meeting XIX. Availability of Documents I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2019-0062 when contacting the NRC about the availability of information for this action. You may obtain publicly available information related to this action by any of the following methods: *Federal Rulemaking Website:*
                         Go to 
                        *https://www.regulations.gov*
                         and search for Docket ID NRC-2019-0062.
                    *NRC's Agencywide Documents Access and Management System (ADAMS):*
                         You may obtain publicly available documents online in the ADAMS Public Documents collection at 
                        *https://www.nrc.gov/​reading-rm/​adams.html.*
                         To begin the search, select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, at 301-415-4737, or by email to 
                        *PDR.Resource@nrc.gov.*
                         For the convenience of the reader, instructions about obtaining materials referenced in this document are provided in the “Availability of Documents” section.
                    *NRC's PDR:*
                         The PDR, where you may examine and order copies of publicly available documents, is open by appointment. To make an appointment to visit the PDR, please send an email to 
                        *PDR.Resource@nrc.gov*
                         or call 1-800-397-4209 or 301-415-4737, between 8 a.m. and 4 p.m. eastern time, Monday through Friday, except Federal holidays. B. Submitting Comments The NRC encourages electronic comment submission through the Federal rulemaking website (
                        *https://www.regulations.gov*). Please include Docket ID NRC-2019-0062 in your comment submission. To facilitate NRC review, please distinguish between comments on the proposed rule and comments on the proposed guidance. The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at 
                        *https://www.regulations.gov*
                         as well as enter the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information. If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment into ADAMS. II. Background A. NRC Advanced Reactor Readiness In its “Policy Statement on the Regulation of Advanced Nuclear Power Plants,” dated July 8, 1986, the Commission stated that it considered the term “advanced” to apply to reactors that are significantly different from current (
                        *i.e.,*
                         current in 1986) generation light-water reactors (LWRs) then under construction or in operation, and that “advanced” includes reactors that provide enhanced margins of safety or utilize simplified inherent or other innovative means to accomplish their safety functions. At the time, certain high temperature gas-cooled reactors, liquid metal reactors, and LWRs of innovative design were considered to be “advanced.” The 1986 policy statement 
                        ( print page 86920)
                        provided the Commission's policy regarding the review of, and desired characteristics associated with, advanced reactors. The NRC updated this statement in the “Policy Statement on the Regulation of Advanced Reactors,” dated October 14, 2008 (Advanced Reactor Policy Statement). The agency has undertaken many activities related to advanced reactors, including issuing an advance notice of proposed rulemaking titled, “Approaches to Risk-Informed and Performance-Based Requirements for Nuclear Power Reactors,” dated May 4, 2006 (71 FR 26267). These efforts were often done in parallel, and sometimes interwoven, with the NRC's efforts to improve risk-informed and performance-based approaches within the agency (
                        *e.g.,*
                         the Commission's policy statement, “Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities,” dated August 16, 1995 (PRA Policy Statement)). In 2016, the NRC issued “NRC Vision and Strategy: Safely Achieving Effective and Efficient Non-Light-Water Mission Readiness” (Advanced Reactor Vision and Strategy Document), in response to increasing interest in advanced reactor designs. The NRC considered the Department of Energy's (DOE's) advanced reactor deployment goals in developing the Advanced Reactor Vision and Strategy Document. Since publication of the document, the NRC continues to manage its activities to support the DOE's deployment goals. The Advanced Reactor Vision and Strategy Document identified initiating and developing a new risk-informed and performance-based regulatory framework as a possible long-term goal. However, the NRC staff's initial efforts were focused on resolving policy issues and developing guidance for licensing non-LWR technologies under the existing regulatory frameworks (parts 50 and 52). The NRC staff issues annual Commission papers on the status and progress of the NRC staff's activities related to advanced reactors (
                        *e.g.,*
                         SECY-24-0020, “Advanced Reactor Program Status,” dated February 27, 2024). These Commission papers provide status updates for advanced reactor activities undertaken both prior to and after initiation of this rulemaking. In 2017, the NRC staff prioritized activities to support the development of technology-inclusive, risk-informed, and performance-based licensing approaches that could be implemented under the existing regulatory framework in parts 50 and 52. One key element of these efforts was the Licensing Modernization Project (LMP), a cost-shared initiative led by nuclear utilities and supported by DOE. The LMP is a technology-inclusive, risk-informed, and performance-based methodology developed for non-LWR designs. The LMP provides a systematic and reproducible process for licensing-basis event (LBE) selection and evaluation; classification of structures, systems, and components (SSCs); and assessment of defense in depth. The LMP refined the DOE's Next Generation Nuclear Plant Program methodologies to reflect interactions with the NRC, to address feedback from industry, and to broaden the scope of the approach to ensure applicability to various non-LWR technologies. The LMP activities led to the publication and submittal of Nuclear Energy Institute (NEI) 18-04, Revision 1, “Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development,” issued August 2019. The document indicates that controlling the frequencies and potential consequences of a wide spectrum of events is the primary focus of the LMP approach. The NRC endorsed the principles and methodology in NEI 18-04, with clarifications, in RG 1.233, “Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors.” The NRC staff sought Commission approval of the use of LMP and NEI-18-04 in SECY-19-0117, “Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors,” dated December 2, 2019. In that paper, the staff described the relationship between the LMP and NEI-18-04 and previous relevant Commission decisions, including those described in SECY-93-092, “Issues Pertaining to the Advanced Reactor (PRISM, MHTGR, and PIUS) and CANDU 3 Designs and their Relationship to Current Regulatory Requirements,” dated April 8, 1993. The Commission approved the use of the LMP methodology and NEI-18-04 as a reasonable approach for establishing key parts of the licensing basis and content of applications for licenses, certifications, and approvals for non-LWRs in Staff Requirements Memorandum (SRM) SRM-SECY-19-0117, dated May 26, 2020. Although the LMP approach is technology- inclusive, the industry and NRC staff initially focused the LMP's applicability on non-LWRs, both for efficiency and to support near-term non-LWR applications under the existing regulatory framework, such as the Advanced Reactor Demonstration Projects supported by DOE. As stated in the part 53 rulemaking plan, SECY-20-0032, the NRC staff developed part 53 by building upon recent and ongoing activities such as the LMP approach described in SECY-19-0117. Such an approach supports implementing the NEIMA requirement to use, where appropriate, risk-informed and performance-based techniques, and it also capitalizes on previous initiatives by the industry, DOE, and the NRC, including the LMP. This approach highlights the role of PRA in risk-informed and performance-based approaches to identifying enhanced safety margins that can be used to justify operational flexibilities. The proposed framework is largely based on the methodology described in SECY-19-0117 and includes a prominent role for PRA. As discussed in section II.B, “Stakeholder Views on Part 53 Preliminary Proposed Rule Language,” of this document, the NRC conducted extensive public outreach on early versions of the proposed rule text. Early versions of the draft proposed rule included two alternative regulatory frameworks. One framework (called “Framework A”) offered a licensing approach centered largely on risk analysis and the other framework (called “Framework B”) largely replicated the existing licensing approach in parts 50 and 52 but modified it to be technology neutral. In its SRM to SECY-23-0021, “Proposed Rule: Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors (RIN 3150-AK31),” the Commission disapproved the inclusion of Framework B in this proposed rule and directed the staff to provide them within one year an options paper for possible future use of the Framework B methodology. B. Stakeholder Views on Part 53 Preliminary Proposed Rule Language In SRM-SECY-20-0032, the Commission directed the NRC staff to prepare and release preliminary proposed rule language, followed by public outreach and dialogue, and then further revise the language until the NRC staff had established the rudiments of its proposed rule for Commission consideration. To implement the Commission's direction, the NRC staff undertook an unprecedented program of stakeholder engagement, recognizing the importance of this rulemaking to the advanced reactor community and 
                        ( print page 86921)
                        interested stakeholders from a broad range of backgrounds and organizations. On November 6, 2020, the NRC published a notification in the 
                        **Federal Register**
                         (85 FR 71002) describing plans for the periodic release of preliminary proposed rule language, meetings with stakeholders, and the ability of stakeholders to provide input during the development of this proposed rule. Sections of the preliminary proposed rule language were subsequently released, and the NRC held numerous public meetings to discuss the preliminary proposed rule language and obtain input from stakeholders. On December 10, 2021, the NRC published a second notification in the 
                        **Federal Register**
                         (86 FR 70423) announcing that the development of the proposed rule and related interactions with stakeholders were being extended until August 31, 2022. By the close of the public stakeholder interactions on August 31, 2022, the NRC staff had held 24 public meetings since September 2020. The NRC staff also met with the Advisory Committee on Reactor Safeguards (ACRS) in 16 public meetings during this period. By the close of the public engagement period on the preliminary proposed rule language, 126 letters were received on the preliminary proposed rule language. Of these 126 letters, 21 were from non-governmental organizations, 31 were from the public, one was from Congress, and the remaining 73 letters were from NRC licensees, the NEI, and other industry groups. In addition, the ACRS wrote four interim letter reports to the Chair on this rulemaking and issued its final letter report on November 22, 2022. The letters from stakeholders provided various points of view and suggestions for clarifications, additions, and deletions to the preliminary proposed rule language. Copies of these letters may be viewed and downloaded from the Federal rulemaking website 
                        *https://www.regulations.gov,*
                         under docket number NRC-2019-0062. The inputs received were considered in the development of this proposed rule. However, as described during the various public interactions related to this rulemaking and in supporting documents, the NRC will not formally disposition the questions and suggestions related to the preliminary proposed rule language as it will for public comments received following the publication of this proposed rule. III. Discussion A. Objective and Applicability The NRC is proposing to add a new, alternative part to its regulations that would set out a risk-informed, technology-inclusive framework for the licensing and regulation of commercial nuclear plants. This new approach would achieve the following: (1) continue to provide reasonable assurance of adequate protection of public health and safety and the common defense and security; (2) promote regulatory stability, predictability, and clarity; (3) reduce requests for exemptions from the current requirements in parts 50 and 52; (4) establish new requirements to address non-LWR technologies; (5) recognize technological advancements in reactor design; and (6) credit the possible response of some designs of commercial nuclear plants to postulated accidents, including slower transient response times and relatively small and slow release of fission products. This proposed rule would add 10 CFR part 53; subpart M, “Fitness for Duty Programs for Facilities Licensed Under 10 CFR Part 53,” to Part 26; § 73.100, “Technology-inclusive requirements for physical protection of licensed activities at commercial nuclear plants against radiological sabotage,” § 73.110, “Technology-inclusive requirements for protection of digital computer and communication systems and networks,” and § 73.120, “Access authorization program for commercial nuclear plants,” as well as make conforming changes throughout 10 CFR chapter I, “Nuclear Regulatory Commission.” B. Need for Changes to the Existing Regulatory Framework The NRC has long recognized that the licensing and regulation of a variety of nuclear reactor technologies would present challenges because the existing regulatory framework has evolved primarily to address the LWR designs that compose the current operating fleet (widely referred to as Generation II reactors). The NRC has had many interactions with designers of various reactor technologies under development, sometimes collectively referred to as advanced reactors (widely referred to as Generation III/III+ (
                        *i.e.,*
                         evolutionary light-water) and Generation IV (
                        *i.e.,*
                         non-light-water) reactors). The interactions have informed the development of policies and guidance to support the potential licensing of new and different types of reactor facilities, some of which may not utilize LWR designs. The NRC issued its Advanced Reactor Policy Statement to provide all interested parties, including the public, with the Commission's views concerning the desired characteristics of advanced reactor designs. The NRC further described its early efforts to establish a technology-inclusive approach to the regulation of nuclear reactors in the advance notice of proposed rulemaking published in 2006. The NRC acknowledged in its “Report to Congress: Advanced Reactor Licensing,” issued August 2012, that while the safety philosophy inherent in the current regulations applies to all reactor technologies, the specific and prescriptive aspects of those regulations clearly focus on the current fleet of LWR facilities. Congress similarly recognized the potential benefits of developing a regulatory infrastructure to support the development and commercialization of advanced nuclear reactors. Consequently, Congress passed NEIMA in late 2018, and the President signed it into law in January 2019. NEIMA directed the NRC to undertake a rulemaking to establish a technology-inclusive regulatory framework for optional use by applicants for new commercial advanced nuclear reactor licenses. In addition, on July 9, 2024, the President signed into law the Accelerating Deployment of Versatile, Advanced Nuclear for Clean Energy Act of 2024, also referred to as the ADVANCE Act. The NRC is evaluating its plans for implementing the ADVANCE Act, including how its regulations, as well as the proposed part 53 or future revisions to it, could be used to address provisions in the ADVANCE Act. The ADVANCE Act contains provisions on a variety of nuclear-related topics, such as micro reactors, nuclear reactor license application reviews, and nuclear fuel. In Section VI, “Specific Requests for Comments,” the NRC is requesting public input on how part 53 could be revised to better enable its potential use to implement the ADVANCE Act. The requirements in part 53 would support a wide variety of potential commercial nuclear reactor technologies. As noted in this discussion, the current regulatory framework in parts 50 and 52 evolved in the context of the current operating reactor fleet dominated by LWRs and as a result includes provisions specific to LWR technologies. While the NRC can license other reactor technologies under the current framework by using existing regulatory flexibilities and the exemption process, there is significant interest in developing a regulatory framework that is flexible enough to accommodate multiple technologies and robust enough to ensure a level of safety equivalent to parts 50 and 52, consistent with the Commission's Advanced Reactor Policy Statement. The Commission reiterated its safety 
                        ( print page 86922)
                        expectations for new reactors in the SRM for SECY-10-0121, “Modifying the Risk-Informed Regulatory Guidance for New Reactors,” dated March 2, 2011: Because new plant designs incorporate operating experience from current generation reactors, severe accident research, and risk insights from design probabilistic risk assessments, the Commission expects that the advanced technologies incorporated in new reactors will result in enhanced margins of safety. However, the Commission continues to expect (consistent with the 2008 Advanced Reactor Policy Statement), as a minimum, at least the same degree of protection of the public and the environment that is required for current-generation light-water reactors. New reactors with these enhanced margins and safety features should have greater operational flexibility than current reactors. However, developing a regulatory framework that can accommodate a wide range of technologies while maintaining an acceptable level of safety presents significant regulatory challenges. The existing regulations have been developed over the course of decades and reflect changes to address events discovered through operating experience. In contrast, part 53 is being developed to accommodate technologies that, in some cases, lack significant operating experience. To address these challenges, the NRC drew on well-developed approaches to licensing to produce a technology-neutral and robust regulatory framework. The proposed regulatory framework would use PRAs to assess risks, help establish technical requirements, and manage operations. The framework builds on the LMP, which is a technology-inclusive approach to licensing that leverages insights from a detailed PRA to provide applicants with significant design and operation flexibilities. C. 10 CFR Part 53: Framework This proposed rule consists of several major components, including a new part 53, to be added to 10 CFR chapter I, revisions for part 26, part 50, and part 73, and conforming changes throughout 10 CFR chapter I. Part 53 is comprised of subparts A through M. These provisions are organized to provide high-level performance criteria and to specify requirements to demonstrate compliance with those performance criteria throughout major stages of the life cycle of commercial nuclear plants. This organization reflects a systems-engineering style approach to the design, licensing, operation, and ultimately decommissioning of future commercial nuclear plants. Organizing requirements in this manner also supports performance-based approaches. Required programs (
                        *e.g.,*
                         radiation protection) and monitoring (
                        *e.g.,*
                         technical specification (TS) surveillance) during the operations phase that are similar to those required by part 50 would complement the design and analysis requirements in subpart C. The performance-based approach proposed in part 53 also includes regulatory requirements that would allow applicants to use a flexible and graded approach to the performance of safety functions based on the role of a particular SSC, human action, or program in limiting the overall risks to the public below accepted standards through balanced measures to prevent and mitigate possible events. Proposed subpart M of part 26 would be new and would be largely consistent with the objective-based fitness for duty (FFD) requirements in current subpart K, “FFD Programs for Construction,” of part 26 supplemented by select requirements from subparts A through I, N, and O of part 26. These requirements are designed to ensure program effectiveness, maintain protections afforded to individuals subject to the FFD program, and align with FFD program implementation by parts 50 and 52 licensees. The proposed requirements are not entirely equivalent because current subpart K of part 26 only applies during construction of the commercial nuclear plant, whereas proposed subpart M of part 26 would apply during construction, operation, and decommissioning. Furthermore, proposed subpart M of part 26 would allow the use of a variety of biological specimens for drug testing as well as innovative technologies for drug and alcohol screening and testing that are not described or allowed by the requirements in subparts A through K, N, and O of part 26, except under limited conditions. Proposed revisions to part 73 would establish a new technology-inclusive consequence-based approach for a range of security areas, including physical security, cybersecurity, and access authorization (AA) for commercial nuclear reactors. The NRC used operating experience to include additional regulatory flexibility for a part 53 licensee's implementation of security requirements. In addition, this proposed rule would make conforming changes throughout 10 CFR chapter I, by adding “and part 53” where appropriate to account for the addition of the proposed part 53. IV. Part 53: Framework Subpart A—General Provisions Subpart A would provide the general provisions applicable to all applicants and licensees that would be established in part 53 for the issuance, amendment, and termination of licenses, permits, certifications, and approvals for commercial nuclear plants licensed under Section 103 of the Atomic Energy Act of 1954, as amended (the Act) and title II of the Energy Reorganization Act of 1974 (88 Stat. 1242). Subpart A would include purpose, scope, definitions, written communications, employee protections, completeness and accuracy of information, exemptions, standards for review, jurisdictional limits, consideration of attacks and destructive acts by enemies of the United States, and information collection requirements. The requirements in subpart A would be largely equivalent to the general requirements in part 50 that are applicable to all part 50 applicants and licensees (specifically, §§ 50.1 through 50.13) but would reference the corresponding regulations in part 53 in place of references to part 50. A. Discussion of Definitions in Proposed Part 53 This proposed rule would include a definition section in § 53.020. The definitions of most terms in § 53.020 would be equivalent to the corresponding terms defined in: (1) §§ 50.2, 52.1, and other NRC regulations; (2) NEI 18-04, as endorsed by RG 1.233; or (3) American Society of Mechanical Engineers (ASME)/American Nuclear Society Risk Assessment Standard (RA-S)-1.4-2021, as endorsed for trial use by RG 1.247, “Acceptability of Probabilistic Risk Assessment Results for Non-Light-Water Reactor Risk-Informed Activities.” This is intended to provide clarity and consistency in terminology where possible and to utilize past and ongoing NRC initiatives to support the licensing of new reactors. Specific deviations from existing definitions are further explained in the following paragraphs. Regarding the definition of “
                        *Commercial nuclear plant*
                        ” and “
                        *Commercial nuclear reactor*
                        ” in proposed § 53.020, as noted previously, the NRC initially considered establishing the scope of part 53 as being for “advanced nuclear plants.” The preliminary proposed rule language defined “advanced nuclear plant” as “a utilization facility consisting of one or more advanced nuclear reactors” as defined in NEIMA. NEIMA defines the term “advanced nuclear reactor” as “a 
                        ( print page 86923)
                        nuclear fission reactor or fusion machine, including a prototype plant (as defined in sections 50.2 and 52.1 of title 10, 
                        *Code of Federal Regulations*
                         (as in effect on the date of enactment of this Act)), with significant improvements compared to commercial nuclear reactors under construction as of the date of enactment of this Act, including improvements such as—(A) additional inherent safety features; (B) significantly lower levelized cost of electricity; (C) lower waste yields; (D) greater fuel utilization; (E) enhanced reliability; (F) increased proliferation resistance; (G) increased thermal efficiency; or (H) ability to integrate into electric and nonelectric applications.” Based on public discussions on the use of the term, the NRC determined that the NEIMA definition, although broad, did not define “significant improvements” with enough specificity to implement in NRC regulations. Additionally, a number of stakeholders suggested that the descriptor, “advanced,” implied enhanced safety, while the NEIMA definition includes “significant improvements” in areas other than safety enhancements. In response to this feedback, and to be technology inclusive, the NRC determined that the broader term “commercial nuclear plant” would be preferable. The NEIMA definition of advanced nuclear reactor also includes fusion technologies. Fusion energy systems have not been included in the scope of part 53 but are the subject of a separate rulemaking activity, “Regulatory Framework for Fusion Systems.” See NRC docket ID NRC-2023-0017 on the Federal rulemaking website 
                        *http://www.regulations.gov.* The NRC proposes to allow use of part 53 by any “commercial nuclear plant.” The use of the term “plant” versus “reactor,” as used in existing regulations (
                        *i.e.,*
                         § 50.2), recognizes that co-located support facilities and radionuclide sources need to be considered in the licensing of a facility. The phrase “commercial purposes,” as used in the definition of “commercial nuclear plant,” includes purposes such as providing process heat for a variety of industrial applications (
                        *e.g.,*
                         desalination, oil refining, hydrogen production). The NRC has not compiled a complete list of such commercial purposes. The definition of “
                        *Commercial nuclear plant*
                        ” refers to a “
                        *Commercial nuclear reactor,*
                        ” which is defined based on the definition of “
                        *Nuclear reactor*
                        ” in § 50.2. However, the phrase “in a self-supporting chain reaction” was removed from the definition to enable applying part 53 to accelerator driven systems that use special nuclear material (SNM) but that do not involve self-sustaining chain reactions. Relatedly, “
                        *Utilization facility*
                        ” is also defined in § 53.020 based on the definition of that term in § 50.2 but is also revised to refer to a “
                        *Commercial nuclear plant*
                        ” as defined in § 53.020. The NRC proposes to include a definition of “
                        *Consensus code or standard*
                        ” in part 53 that is based on the use of these terms in the National Technology Transfer and Advancement Act of 1995 (NTTAA) (Pub. L. 104-113) and the Office of Management and Budget (OMB) Circular No. A-119, “Federal Participation in the Development and Use of Voluntary Consensus Standards and in Conformity Assessment Activities.” As required by NTTAA, the NRC undertakes the following activities: (i) consults with voluntary consensus standards bodies; (ii) participates with voluntary consensus bodies in the development of consensus standards; and (iii) uses consensus standards as a means to carry out the NRC's policy objectives. In part 53, the NRC is not proposing to incorporate by reference specific codes and standards as is done under the existing regulations in § 50.55a, “Codes and standards,” because some codes and standards are LWR-specific. Part 53 would require that design features must be designed using generally accepted consensus codes and standards but would not incorporate the specific code or standard into the NRC's regulations. During public meetings, significant discussions with stakeholders indicated that future reactor designers were interested in the use of international consensus standards that have not yet been endorsed by the NRC. The definition proposed in part 53 would allow for the use of international codes and standards not previously used in NRC licensing but recognizes that the use of any consensus code or standard would ultimately need to be found acceptable by the NRC, either through generic efforts to endorse a code or standard or on an application-specific basis during an individual licensing review. The proposed definition of “
                        *Construction*
                        ” is slightly different than the definition in § 50.10—it would cover the same concept but be applied to a slightly different scope of activities based on how SSCs are classified under part 53. In part 53, the definition of “
                        *Construction*
                        ” is based on the definition in § 50.10 but modified to apply to safety-related (SR) and non-safety-related but safety-significant (NSRSS) SSCs identified by the design and analysis requirements in subparts B and C to ensure the safety criteria are met. Section 53.020 would also add definitions for terms related to event selection (LBEs, design-basis accidents (DBAs), anticipated event sequences, unlikely event sequences, and very unlikely event sequences); equipment classifications (SR, NSRSS, and non-safety-significant SSCs); performance metrics (
                        *e.g.,*
                         safety criteria and functional design criteria); and special treatment. The regulation would define “
                        *Safety criteria*
                        ” in terms of the plant-level performance-based metrics that would be provided in §§ 53.210 and 53.220. The term “
                        *Functional design criteria*
                        ” would be defined as metrics for the performance of specific SSCs that are determined from the role of the SSC in meeting the safety criteria. These are new terms that have not previously been defined or used in NRC regulation. The term “
                        *Safety-related SSCs*
                        ” would refer to those SSCs needed to meet the safety criteria in § 53.210. The term “
                        *Non-safety-related but safety-significant SSCs*
                        ” would mean those SSCs that are not SR because they are not relied upon to perform any function necessary to demonstrate compliance with § 53.210 but warrant special treatment because they are relied on to achieve adequate defense in depth or perform risk-significant functions. The term “
                        *Special treatment*
                        ” would be defined as requirements, such as quality assurance and programmatic controls, identified for each design feature to ensure that the safety criteria are satisfied and the safety functions are fulfilled. These requirements would also ensure that SR and NSRSS SSCs will provide defense in depth, or perform risk-significant functions, under service conditions and with SSC reliabilities that are consistent with the analysis required in proposed subpart C. Structures, systems, and components designated as SR would also contribute to defense in depth and risk-significant functions and may warrant special treatments beyond those defined for the SR functions needed for compliance with § 53.210. The term “
                        *Non-safety-significant SSCs*
                        ” would mean those SSCs that are not SR or NSRSS. The terms “
                        *Design-basis accidents,*
                        ” “
                        *Anticipated event sequences,*
                        ” “
                        *Unlikely event sequences,*
                        ” and “
                        *Very unlikely event sequences*
                        ” would be defined to be different types of “
                        *Licensing-basis events*
                        ” and would also be largely equivalent to the LMP's definitions of DBAs, anticipated operational occurrences (AOOs), design-basis events (DBEs), and beyond-design-basis events, respectively. The term 
                        ( print page 86924)
                        “
                        *Design-basis accidents*
                        ” would be defined as postulated event sequences that are used to set functional design criteria and performance objectives for the design of SR SSCs through deterministic analyses. Design-basis accidents would be derived from the unlikely event sequences from the PRA and then analyzed in a conservative approach by prescriptively assuming that only SR SSCs are available to mitigate postulated accident scenarios. Within the LMP methodology, event sequences with mean frequencies of 1 × 10−2
                        /plant-year and greater would be classified as anticipated event sequences. Within the LMP methodology, infrequent event sequences with mean frequencies of 1 × 10−4
                        /plant-year to 1 × 10−2
                        /plant-year would be classified as unlikely event sequences. “
                        *Very unlikely event sequences*
                        ” would be less likely to occur than unlikely event sequences. Within the LMP methodology, rare event sequences with frequencies of 5 × 10−7
                        /plant-year to 1 × 10−4
                        /plant-year would be classified as very unlikely event sequences. While the proposed terminology for these event sequences would create some differences between part 53 and the LMP, part 53 would use new terms for these event sequences specifically to avoid conflicts with terms already used within part 50 and part 52 to represent different concepts. Further, because some stakeholder comments demonstrated confusion related to the history of beyond-design-basis accidents terminology, these definitions seek to clarify the event categories in part 53. The sections of this preamble related to subparts B and C provide additional discussion of LBEs. B. Other General Provisions Section 53.040 would govern written communications and how applications and other required information must be submitted to the NRC. These requirements would be equivalent to those in § 50.4. Section 53.050 would establish requirements for enforcement action to which a licensee, an applicant, or a licensee's or applicant's contractor or subcontractor, or an employee of any of them may be subject for engaging in deliberate misconduct. These requirements would be equivalent to those in § 50.5. Section 53.060 would prohibit discrimination against an employee of a holder or applicant for an NRC license, permit, design certification (DC), or design approval, or a contractor or subcontractor of a holder or applicant for an NRC license, permit, DC, or design approval for engaging in certain protected activities. Section 53.060 also would prescribe a procedure for seeking a remedy for employees who believe they have been discriminated against for engaging in such protected activities. These requirements would be equivalent to those in §§ 50.7 and 52.5. Section 53.070 would govern the completeness and accuracy of information provided to the NRC. These requirements would be equivalent to those in §§ 50.9 and 52.6. Section 53.080 would govern exemptions from the requirements of the regulations in part 53. These requirements would be equivalent to those in §§ 50.12 and 52.7. Paragraphs (a) through (d) of § 50.90 would establish requirements for standards that the NRC would consider in determining whether a construction permit (CP), operating license (OL), early site permit (ESP), combined license, or manufacturing license (ML) under part 53 would be issued to an applicant. These requirements would be equivalent to those in §§ 50.40, 50.42, 50.43 and 50.22, respectively. Requirements equivalent to those in §§ 50.41 and 50.21 would not be included in part 53 because they apply to Class 104 licenses, and part 53 would not apply to those licenses. Section 53.100 would require that no license issued under part 53 would cover activities which are not under or within the jurisdiction of the United States. These requirements would be equivalent to those in § 50.53. Section 53.110 would state that licensees and applicants would not be required to provide design features or other measures for the specific purpose of protection against the effects of attacks and destructive acts by enemies of the United States directed against the facility or deployment of weapons incident to U.S. defense activities. These requirements would be equivalent to those in § 50.13. Section 53.115 would establish requirements for rights related to SNM. These requirements would be equivalent to those in § 50.54(b) and (c). Section 53.117 would establish requirements for license suspension and rights of recapture of the material or control of the facility in a state of war or national emergency declared by Congress. These requirements would be equivalent to those in § 50.54(d). Section 53.120 would establish requirements for information collection requirements and OMB approval. These requirements would be equivalent to those in § 50.8. Subpart B—Technology-Inclusive Safety Requirements Proposed subpart B, “Technology-Inclusive Safety Requirements,” would provide technology-inclusive safety criteria that would serve as performance standards for the subsequent performance-based requirements used throughout part 53. Subsequent subparts would define how specific activities during various stages of the life cycle of a commercial nuclear plant contribute to satisfying these high-level performance standards. The performance standards in subpart B would also establish a means to determine appropriate regulatory controls for SSCs, human actions, and programs in the following subparts. For example, the classification of SR SSCs would be built upon the proposed safety criteria in § 53.210, “Safety criteria for design-basis accidents.” The more detailed requirements for those SSCs would then be further defined in the design and analysis requirements in subpart C, “Design and Analysis Requirements.” The activities for manufacturing, constructing, and maintaining the SR SSCs would be governed by subpart E, “Construction and Manufacturing Requirements,” and subpart F, “Requirements for Operation.” Requirements for NSRSS SSCs warranting special treatment would likewise be determined under § 53.220, “Safety criteria for licensing-basis events other than design-basis accidents,” in subpart B and § 53.460, “Safety categorization and special treatment,” in subpart C. Regulatory requirements related to the NSRSS SSCs would be distinguished from the regulatory requirements for SR SSCs throughout part 53. Part 53 would afford more flexibility to applicants and licensees regarding how NSRSS SSCs would be used in the design and maintained during plant operations, as compared to SR SSCs. The collective set of performance-based requirements in part 53 would be sufficient, if met, for the NRC to make the findings required to grant an application for a utilization facility under Section 182 of the Act that the utilization of SNM will be in accord with the common defense and security and will provide adequate protection to the health and safety of the public. This construct would be similar to existing NRC regulations, which the Commission has said on many occasions do not specifically define “adequate protection.” However, compliance with NRC regulations may be presumed to assure adequate protection at a 
                        ( print page 86925)
                        minimum. The requirements throughout part 53 that support demonstrating compliance with § 53.220 would be similar to current regulations that both contribute to assuring adequate protection of public health and safety and are desirable to promote the common defense and security or to protect health or to minimize danger to life or property under Section 161 of the Act. Consistent with historical practice, Sections 182 and 161 of the Act are cited as authorizing legislation within this proposed rule. However, specific language from the Act would not be incorporated into the safety objectives or safety criteria in part 53. This is because, again consistent with historical practice, the NRC would not be defining “adequate protection” through the individual safety requirements in part 53. Rather, part 53 would enable the NRC to make its required findings under the Act by providing sufficient performance standards, safety criteria, and related requirements on how applicants must demonstrate compliance with subpart B and other subparts. Section 53.210 would provide safety criteria for DBAs that would be required to be identified under § 53.240 and analyzed under § 53.450(f) in subpart C of part 53. Subsequent sections in part 53 would require that the SSCs relied upon to demonstrate compliance with the criteria in § 53.210 be classified as SR. The use of SR SSCs and the 25 rem reference values for potential radiological consequences would align with traditional deterministic approaches for LWRs from §§ 50.34, 52.79, and 100.11 for evaluating the effectiveness of plant design features with respect to postulated reactor accidents. A footnote similar to that included in § 50.34(a)(1)(ii)(D)(
                        *1*) and § 52.79(a)(1)(vi)(A) would be included in § 53.210 to explain that the use of the 25 rem value would not be intended to imply that this number constitutes an acceptable limit for an emergency dose to the public under accident conditions. Rather, this dose value has been set forth in this proposed section as a reference value that would be used in the evaluation of plant design features with respect to DBAs to verify that the proposed designs would provide assurance of low risk of public exposure to radiation in the event of an accident. The inclusion of the safety criteria for DBAs in subpart B would provide a logical structure supporting the identification and treatment of SR SSCs and establishing the corresponding functional design criteria for those SSCs. Section 53.220 would provide safety criteria for LBEs other than DBAs that would be required to be identified under § 53.240 and analyzed under § 53.450(e) in subpart C. Whereas § 53.210 and the related requirements for SR SSCs would provide that a defined success path exists for DBAs, the safety criteria for LBEs other than DBAs would establish the connections between SSC design, human actions, and programmatic controls and a broader set of potential internal and external hazards. These safety criteria would also address defense-in-depth matters such as a balanced consideration of prevention and mitigation. The safety criterion in § 53.220(b) would include a requirement to use a comprehensive risk metric or set of metrics and associated risk performance objectives against which calculated values of the risk metrics are compared. The comprehensive risk metrics or set of metrics and associated risk performance objectives would support a performance-based approach to developing an appropriate combination of design features and programmatic controls to prevent or mitigate LBEs other than DBAs. The applicant must propose the comprehensive risk metric or set of metrics and associated risk performance objectives, and the comprehensive risk metric or set of metrics and associated risk performance objectives must provide an appropriate level of safety. Comprehensive risk metrics should consist of a proposed plant risk metric or set of proposed risk metrics that approximate the total, overall risk from the facility and that address the range of possible plant configurations and associated internal and external hazards to the extent practicable. The associated risk performance objectives are preestablished, indicative values of the comprehensive risk metrics that are used as part of risk-informed decision-making. The methodology for developing and using proposed comprehensive risk metrics and associated risk performance objectives is defined by the proposed requirements for analyses in § 53.450. Therefore, the application must include a description of that methodology and, among other things, should explain the initial conditions, boundary conditions, and key assumptions used to develop and calculate the risk metrics. Screening tools and bounding or simplified methods may be used for any mode or hazard, provided that the applicant provides an acceptable technical basis. As with all risk-informed methodologies, treatment of uncertainties must be addressed. The risk performance objectives established under this methodology are likely to involve assessing and averaging the risks over a period of time (
                        *e.g.,*
                         plant year) and would not constitute a real-time requirement that must be continuously demonstrated by the licensee. The use of a comprehensive risk metric or set of risk metrics and risk performance objectives that reflect an average risk to establish performance goals for SR and NSRSS SSCs is consistent with current practices that use other risk assessment techniques to address short-term plant configurations during plant maintenance activities. It is worth noting that the evaluation of plant risks, as represented by a comparison of analysis results to acceptable risk performance objectives for comprehensive risk metrics, would be one of several performance standards used in subpart B. The proposed use of multiple performance standards, including deterministic criteria and defense-in-depth measures, reflects an integrated decision-making process similar to that described in RG 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,” Revision 3. The NRC's approval of using a comprehensive risk metric or set of metrics with associated risk performance objectives is not, by itself, an indicator of adequate protection. Rather, the comparison of comprehensive risk metrics to associated risk performance objectives that are acceptable to the NRC is part of a suite of regulatory requirements that, when considered holistically, form the basis for the NRC's decision-making. This is analogous to the approach used for plants licensed under part 50 and part 52, where no single regulatory requirement governs whether a plant is “safe enough.” The RG 1.233, “Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors,” describes an example of an acceptable approach for identifying and analyzing LBEs under part 50 and part 52, including the use of the quantitative health objectives (QHOs) stated in the NRC's policy statement, “Safety Goals for Nuclear Power Plant Operation,” dated August 4, 1986 (51 FR 28044), as corrected and republished August 21, 1986 (51 FR 30028) (Safety Goals Policy Statement), as acceptable performance objectives for 
                        ( print page 86926)
                        comprehensive risk metrics. The use of comprehensive risk metrics, such as the individual early fatality risk (IEFR) and the individual latent cancer fatality risk (ILCFR), and associated risk performance objectives, such as the QHOs, from the Safety Goals Policy Statement, could form the basis for one approach to meet § 53.220(b). The requirement for comprehensive risk metrics, in combination with the other proposed requirements in subparts B and C, would bring the approach endorsed in RG 1.233 for parts 50 and 52 into part 53. Additionally, the use of comprehensive risk metrics and associated risk performance objectives would provide a logical performance objective to support the risk management approaches in the various subparts comprising proposed part 53. The Commission stated in the introduction of the Safety Goals Policy Statement that improvements to then-current regulatory practices could lead to a more coherent and consistent regulation of nuclear power plants, a more predictable regulatory process, a better public understanding of the regulatory criteria that the NRC applies, and public confidence in the safety of operating plants. Accordingly, the Commission announced the safety goals with a focus on the risks to the public from nuclear power plant operation. Following the issuance of the Safety Goals Policy Statement, the NRC has used the comprehensive risk metrics and performance objectives provided in the safety goals within the criteria for many decisions involving safety judgments during the licensing and regulation of operating reactors and proposed nuclear reactor designs. Consistent with NUREG-0880, the proposed comprehensive risk metrics and associated risk performance objectives required under § 53.220(b) could be expressed in terms of a biologically average individual in terms of age and other risk factors. Although some comprehensive risk objectives such as the IEFR and ILCFR are defined in terms of fatality risks, the Commission continues to make clear that no death attributable to nuclear power plant operation will ever be “acceptable” in the sense that the Commission would regard it as a routine or permissible event. Comprehensive risk metrics and associated risk performance objectives as used in this proposed rule would establish acceptable risks, not acceptable deaths. Applicants under the proposed part 53 may choose to develop and seek NRC approval of comprehensive risk metrics or sets of risk metrics and associated risk performance objectives beyond those discussed above, including the use of surrogate measures for use in specific analyses to satisfy the proposed requirements in § 53.220(b). Such surrogate measures for comprehensive risk metrics and associated risk performance objectives could be used in a manner similar to the use of core damage frequency and conditional containment failure probability for LWRs within the safety goal evaluation process in NUREG/BR-0058, “Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commission,” and other assessments of LWRs using the NRC's safety goals. The NRC would, as appropriate, review novel approaches for comprehensive metrics and associated risk performance goals proposed by applicants, industry organizations, or standard development organizations and would engage stakeholders during the development of the related regulatory guidance or specific licensing actions. Section 53.230 would require safety functions needed to ensure that the safety criteria under §§ 53.210 and 53.220 can be met if an assumed LBE were to occur at a commercial nuclear plant. Section 53.230 would specify that limiting the release of radioactive materials from the facility is the primary safety function, and therefore, limiting potential offsite consequences (
                        *i.e.,*
                         dose to a hypothetical individual) would be used as the primary performance metric throughout part 53. The additional or subsidiary safety functions needed to limit the release of radionuclides may include, without limitation, controlling processes related to reactivity, heat generation, heat removal, and chemical interactions. This proposed rule provides flexibility to applicants and licensees in identifying, implementing, and maintaining the safety functions supporting retention of radionuclides for commercial nuclear plants of varying sizes and technologies. Proposed § 53.240 would require applicants to identify and address LBEs. LBEs are unplanned events, resulting from both internal and external hazards, that are used in the design and analyses required under part 53 for licensing commercial nuclear plants. This ensures estimates of offsite consequences from analyses performed under proposed § 53.450 are below the safety criteria identified under proposed §§ 53.210 and 53.220 and that SSCs, personnel, and programs address the safety functions from proposed § 53.230. Including a high-level performance requirement related to the identification and analysis of LBEs in subpart B would reflect the historical and continuing importance of evaluating unplanned events as part of the licensing of commercial nuclear plants. Proposed § 53.240 would require identification and analysis of LBEs under § 53.450, which would require a PRA. Examples of acceptable methods of using PRAs to identify and assess LBEs would be the methodology in RG 1.233, as discussed in Draft Regulatory Guide (DG)-1413, “Technology-Inclusive Identification of Licensing Events for Commercial Nuclear Plants.” Section 53.250 would establish defense-in-depth requirements based on the longstanding philosophy of providing defense in depth to address uncertainties about the design, operation, and performance of commercial nuclear plants. For example, parts 50 and 52 address defense in depth through layered prescriptive technical requirements (
                        *e.g.,*
                         fuel performance, cladding integrity, reactor coolant system integrity, containment performance) for LWRs. In contrast, the flexibility afforded to applicants in how they propose to demonstrate compliance with the high-level safety criteria within part 53 would necessitate this specific requirement to ensure defense in depth is provided. The requirements in this section would state that no single engineered design feature, human action, or programmatic control, no matter how robust, should be exclusively relied upon to address LBEs other than DBAs. The phrase “engineered design feature” would not preclude the possible crediting of inherent characteristics within the design and analysis for commercial nuclear reactors. While defense in depth would only be assessed for LBEs other than DBAs, the need to ensure dedicated success paths for DBAs would contribute to the overall defense in depth for each commercial nuclear plant under part 53. Section 53.260 would govern normal operations and would establish a level of safety based on current requirements in 10 CFR part 20, “Standards for Protection Against Radiation,” which limits doses to members of the public and dose rates in unrestricted areas. Section 53.270 would provide for the protection of plant workers and would establish a level of safety based on current requirements in 10 CFR part 20 which limits occupational dose. Subpart C—Design and Analysis Requirements This subpart would provide requirements for the design of commercial nuclear plants and the supporting analyses, including the analyses of LBEs, to demonstrate that the performance standards in proposed 
                        ( print page 86927)
                        subpart B can be satisfied. The sections within subpart C would reflect the overall hierarchy throughout part 53, which would cover: (1) plant-level safety criteria (§§ 53.210, 53.220, and 53.470); (2) safety functions (§ 53.230) needed to demonstrate compliance with the safety criteria; (3) design features (§ 53.400), human actions, and programmatic controls needed to fulfill the safety functions; and (4) functional design criteria (§§ 53.410 and 53.420) that must be defined for each design feature relied on to demonstrate the safety criteria (§§ 53.210, 53.220, and 53.470) are met. Subpart C would also contribute to the logic and structure of part 53 by distinguishing between SR SSCs and NSRSS SSCs and licensee-controlled programs that address LBEs other than DBAs. Specifically, SR SSCs, human actions, and programmatic controls needed to protect against DBAs are used to satisfy the safety criteria in § 53.210. Non-safety-related but safety-significant SSCs, human actions, and licensee-controlled programs that address LBEs other than DBAs generally contribute to the appropriate measures considering potential risks to public health and safety. Section 53.400 would establish a requirement that design features be provided for each commercial nuclear plant to satisfy the safety criteria and fulfill safety functions from proposed subpart B during LBEs. Other sections in subpart C would, in turn, further address the necessary capabilities and reliabilities for SSCs by establishing functional design criteria, fulfilling design requirements, performing analyses of LBEs, performing other supporting analyses, and categorizing SSCs based on their roles in preventing or mitigating LBEs. Section 53.410 would require that functional design criteria be defined for design features relied upon to demonstrate that the consequences from DBAs would be below the criteria in § 53.210 through analyses performed under § 53.450(f), which includes insights from both PRAs and deterministic analyses. Other sections within part 53 would establish appropriate controls on these design features (
                        *e.g.,*
                         safety classification, protection from external hazards, quality assurance, and TS) to ensure the functional design criteria are satisfied. The performance requirements for the SSCs needed to address DBAs and the corresponding human actions and programmatic controls would contribute to ensuring that a commercial nuclear plant licensed under part 53 would meet the safety criteria in § 53.210. Section 53.415 would require that SR SSCs be protected against or designed to withstand the effects of natural phenomena (
                        *e.g.,*
                         earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches) and constructed hazards (
                        *e.g.,*
                         from dams, transportation routes, and military or industrial facilities). Specifically, § 53.415 would require that SR SSCs remain capable of performing the safety functions stated in § 53.230 for which they are credited up to the design-basis external hazard levels as determined under § 53.510. As used in § 53.415 and subpart D of part 53, a hazard level would refer to such things as the magnitude and recurrence rate of an earthquake and the resultant ground motions, the height of a flood, the force of hurricane winds, or the concentrations of chemicals resulting from a release from a nearby facility. These requirements would support either traditional deterministic approaches for determining and protecting against external hazards or probabilistic approaches that are being developed for seismic and some other external hazards. Section 53.420 would require that functional design criteria be defined for design features that play a significant role in demonstrating that the safety criteria for LBEs other than DBAs are satisfied. The analyses required for this demonstration would be described in proposed § 53.450(e), which would require that those events be identified and assessed using a PRA methodology in combination with other generally accepted approaches for systematically evaluating engineered systems. The SSCs determined to be safety significant (
                        *i.e.,*
                         either SR or NSRSS) would have associated special treatment requirements as specified in § 53.460. Special treatment would be defined in subpart A of part 53 and generally refers to measures (
                        *e.g.,*
                         quality assurance, testing, monitoring) taken beyond the procurement and installation of commercial grade products to provide confidence that the SSC will comply with the applicable functional design criteria. The inclusion of a systematic approach to identifying the functional design criteria for SSCs and tailoring the special treatments to specific LBEs and safety functions is an important contributor to satisfy the proposed safety criteria in subpart B. Therefore, designers and licensees for commercial nuclear plants would be provided flexibility on how LBEs other than DBAs are either prevented or mitigated and how the calculated comprehensive plant risks satisfy the safety criterion established under § 53.220(b). Section 53.425 would establish requirements for design features and related functional design criteria limiting doses to members of the public during normal operations to satisfy the criteria in part 20. Section 53.430 would provide similar requirements for design features and related functional design criteria for protection of plant workers to meet the safety criteria in part 20. Similar to existing regulations, the NRC considers that licensees would generally comply with the requirements of part 20 to keep doses as low as reasonably achievable by meeting a design objective of keeping doses to the public from routine plant effluents less than 10 millirem per year. This goal is similar to that provided by appendix I to part 50 and would assist designers, applicants, and licensees in performing the evaluations of possible reductions in public dose from routine effluents when considering costs and other factors. As emphasized in existing regulations in part 50, the design objective of keeping doses to the public from routine plant effluents less than 10 millirem per year should not be construed as a radiation protection standard. The NRC anticipates that future guidance will continue to reflect this performance goal. The proposed requirements in §§ 53.425 and 53.430 for design features and functional design criteria to support radiation protection activities have parallels in existing regulations such as § 50.34(a) and (b)(3), which require in part that the means be provided for meeting the requirements of part 20 and General Design Criterion 60, 61, 63, and 64 in appendix A to part 50, which provide radiation protection related design criteria. Section 53.440 would address various design requirements that warrant specific mention to ensure that the design features required by § 53.400 comply with the functional design criteria required by §§ 53.410 and 53.420. These requirements would be met through design practices, consideration of testing and operating experience, and various assessments of LBEs and other potential challenges to commercial nuclear plants. Discussions of some of the key design requirements included in this section follow. § 53.440(a): An essential element to ensuring a proposed design can comply with the performance criteria in proposed part 53 would be that the abilities of design features to fulfill their safety functions are demonstrated by a combination of analyses, test programs, prototype testing, and operating experience. This requirement closely aligns with the language in § 50.43(e)( print page 86928)
                        and is proposed in part 53 as the same foundational requirement. In addition, the proposed § 53.440(a) would require the design processes for SSCs under this section to include administrative procedures for evaluating operating, design, and construction experience for considering applicable important industry experiences in the design of those SSCs. This proposed requirement corresponds to the existing requirement under § 50.34(f)(3)(i) that was developed in response to the 1979 accident at Three Mile Island Nuclear Generating Station.
                    § 53.440(b): The design and licensing of commercial nuclear plants should use generally accepted consensus codes and standards. Such codes and standards ensure sufficient testing and qualification of materials and equipment and provide defined processes, specifications, and acceptance criteria for use by designers and suppliers. The NRC would indicate acceptance of consensus codes and standards used in the design and licensing of a specific commercial nuclear plant either through the NRC's generic endorsement of a code or standard (*i.e.,*
                         through regulatory guidance), including any limitations or conditions, that can be referenced within an application, or through the review of a referenced code or standard as part of the review of a specific application.
                    § 53.440(c): The design requirements in subpart C would require the materials used for SR and NSRSS SSCs to be qualified for their service conditions over the design life of the SSC.
                    § 53.440(d): The requirements in § 53.440 would include the need to consider possible degradation mechanisms for materials and equipment to inform both the design process and the development of integrity assessment programs to be executed during plant operations in accordance with subpart F of part 53. The inclusion of requirements related to designing and monitoring for possible degradation mechanisms reflects important lessons learned from the history of LWRs as well as operating experience with structures and systems in countless other engineering endeavors.
                    § 53.440(e) and (f): The design requirements in subpart C would state specific design requirements similar to existing requirements in parts 50, 52, and 73 for protections against fires and explosions and consideration of safety and security together in the design process.
                    § 53.440(g) and (h): Specific design requirements are proposed to ensure that commercial nuclear reactors under part 53 have the capability to achieve and maintain subcriticality and long-term cooling. The requirements would be included to address the potential that some reactor designs may be able to achieve a stable end state for the purpose of event analyses but might need further actions to completely shut down and service the facility.
                    § 53.440(i): The design, analysis, and development of programmatic controls under part 53 would consider the number of reactor units and other significant inventories of radioactive materials contributing to the risks to public health and safety. This would reflect the definition of “*Commercial nuclear plant*
                        ” in subpart A and reinforce that the evaluation of LBEs is performed on a plant-wide basis. This aspect of part 53 would be different from parts 50 and 52, which generally define safety requirements on the assumption of events involving only individual reactor units.
                    § 53.440(j): A design requirement is proposed to provide a technology-inclusive requirement that would be equivalent to the requirements in § 50.150 to address the possible impact of a large commercial aircraft.
                    § 53.440(k): The inclusion of a specific proposed requirement to address the risks to public health from potential chemical hazards of licensed material is appropriate given the diversity of reactor technologies and designs that might be licensed under part 53. The requirement in part 53 would be similar to the existing requirements in10 CFR part 70, “Domestic Licensing of Special Nuclear Material,” that address both potential radiological and chemical hazards for licensed materials at fuel cycle facilities.
                    § 53.440(l): Provisions are proposed to require that measures be taken during the design of commercial nuclear plants to minimize contamination of the facility and the environment, facilitate eventual decommissioning, and minimize the generation of radioactive waste in accordance with § 20.1406.
                    § 53.440(m): A design requirement is proposed to provide a technology-inclusive equivalent to the requirements in § 50.68 by including options for commercial nuclear plants to either have a monitoring system capable of detecting a criticality as described in § 70.24 or to have restrictions on SNM handling and storage that would prevent inadvertent criticality events.
                    § 53.440(n): The design would need to reflect state-of-the-art human factors principles for safe and reliable performance in all settings that human activities are expected for performing or supporting the continued availability of plant safety or emergency response functions. Section 53.450 would establish analysis requirements and would center upon the use of a PRA in combination with other generally accepted approaches for systematically evaluating engineered systems. The reliance on PRAs as a key component in the proposed analysis requirements for part 53 would reflect the decades of improvements in PRA methodologies and the increasing use of PRA techniques in the design, licensing, and oversight of both operating and future nuclear reactors. Part of the Commission's PRA Policy Statement is that the use of PRA technology should be increased in all regulatory matters to the extent supported by the state of the art in PRA methods and data and in a manner that complements the NRC's deterministic approach and supports the NRC's traditional defense-in-depth philosophy. The need to supplement PRA insights with other engineering approaches and judgments reflects the NRC's longstanding policy described in the SRM to SECY-98-144, “Staff Requirements—SECY-98-144—White Paper on Risk-Informed and Performance-Based Regulations,” dated February 24, 1999, for regulatory decision-making to be risk-informed but not solely based on numerical results of a risk assessment (
                        *i.e.,*
                         not a risk-based approach). Part 53 would maintain a role for NRC's traditional deterministic approaches (particularly for DBAs) and defense-in-depth philosophy by including specific requirements utilizing these regulatory tools in subparts B and C. PRA would be used in combination with other techniques in part 53 to identify and categorize LBEs, classify SSCs, and evaluate defense in depth. This increased role for the PRA necessitates that it would be developed, performed, and maintained in accordance with NRC-approved standards and practices (see § 53.450(c) and (d)). The computer codes used to model the plant response and the behavior of the barriers to the release of radionuclides would need to be qualified for the range of conditions being simulated across a wide range of unplanned events. These analyses would need to use realistic approaches and address uncertainties associated with states of knowledge, modeling, and performance of SSCs. While industry consensus PRA standards and peer review processes endorsed in RGs 1.200 and 1.247 remain 
                        ( print page 86929)
                        acceptable for developing a PRA, they are not regulatory requirements and an application under part 53 need not follow every aspect of the applicable consensus PRA standard. Existing processes for defining the scope and capability of a PRA supporting an application offer flexibility in determining the degree to which the PRA needs to be developed and may be informed by other factors such as design complexity and the needed degree of realism and level of detail, consistent with the use of the PRA and substance of the application. Such processes are currently available for appropriately defining the scope of the PRA and determining applicability of supporting requirements in consensus PRA standards needed to satisfy the proposed regulatory requirements for the specific uses of analyses under § 53.450(b). Likewise, NRC determinations of the acceptability of such PRAs would include consideration of the appropriateness of the applicant-defined scope as part of determining the applicability of and conformance to consensus PRA standard supporting requirements consistent with the current state of practice. In addition, these determinations would include consideration of other aspects of the development of the PRA, such as PRA peer reviews. An NRC determination of the acceptability of a PRA includes but is not limited to assessing the initial and boundary conditions and key assumptions used in the analysis, treatment of uncertainties, and the use of screening tools and bounding or simplified methods for any mode or hazard, provided the use of those tools and methods is justified by an acceptable technical basis. In that regard, the consensus PRA standards would not be applied by the NRC as a strict checklist of requirements for part 53 PRA acceptability determinations. The proposed § 53.450(c) would require periodic maintenance and upgrading of the PRA to maintain an alignment between the supporting analyses and the design and performance of plant equipment, programs and procedures, and other factors associated with meeting the safety criteria of the proposed § 53.220 and the evaluation criteria of proposed § 53.450(e)(2). The periodic maintenance of the PRA would also be a means to consider new or revised information related to external hazards, industry operating experience, performance issues with or degradation of SSCs, and other contributors to the frequency and potential consequences of various event sequences. The periodic assessments performed by licensees to support the maintenance of the PRA and other requirements in the proposed part 53 would be complemented by NRC inspections and programs to assess new or revised information related to topics such as natural hazards, operating experience, and potential generic safety issues. The categories of LBEs used in part 53 would include anticipated event sequences, unlikely event sequences, and very unlikely event sequences. The unlikely event sequences would include those events with estimated frequencies well below the frequency of events expected to occur during the lifetime of a commercial nuclear plant. An important aspect of the analysis requirements is that, under proposed § 53.450(e), the analyses of LBEs other than DBAs would not only be used to show the performance criteria of § 53.220 are satisfied but to also show that evaluation criteria defined for each LBE or category of LBEs would also be satisfied. Such evaluation criteria for specific LBEs or categories of LBEs would be defined in terms of limits on the release of radionuclides or maintaining the integrity of one or more barriers used to limit the release of radionuclides and reflect a graded approach of allowing lesser potential consequences from more frequent events. An example of such evaluation criteria for a range of LBEs that could likely be expanded for part 53 is provided in RG 1.233. Another proposed requirement for the proposed § 53.450(e) analyses is that the methodology would need to include a means to identify event sequences deemed risk-significant such that those event sequences can be given special attention within other sections of part 53. Part 53 would maintain an important role for a deterministic analysis of DBAs in the performance criteria of § 53.210 and the related analytical requirements in § 53.450(f). The analysis of DBAs would be required to address event sequences drawn from those with estimated frequencies below the expected lifetime of a generation of reactors (
                        *e.g.,*
                         event sequences with frequencies as low as one in ten thousand years). As proposed in this section, DBAs would need to be analyzed using deterministic methods and ensure a safe, stable end state with reliance upon only SR SSCs and human actions, if needed, to be performed by operators licensed under the provisions of §§ 53.760 through 53.795. While the DBAs analyzed under part 53 would be similar to the traditional DBAs analyzed under parts 50 and 52, there are important distinctions between the overall role of DBA analyses in part 50 and proposed part 53. In part 53, the role of the DBA analysis would be more narrowly focused on selecting SR SSCs and determining functional design criteria for those SSCs to ensure the commercial nuclear plant meets the safety criteria in § 53.210. The overall control of risks posed by commercial nuclear plants under part 53 would be provided by the analyses of and measures taken for both DBAs and other LBEs, including very unlikely event sequences. This would contrast with the traditional deterministic approach in part 50 wherein the analyses of DBEs such as DBAs were used to provide bounding assessments, incorporate standard design rules such as assumptions related to single failures, and to define conservative performance requirements for SR SSCs. Limitations related to the traditional deterministic approach were addressed in part 50 through case-by-case assessments and specific actions for beyond-design-basis events such as anticipated transients without scram and station blackout. Section 53.450 would also include provisions to ensure that analyses are performed to support the design requirements of § 53.440(e) on fire protection, § 53.440(j) on aircraft impact assessments, and § 53.425 on using design features and plant programs to control doses to members of the public from routine effluents and direct radiation from contained sources. The proposed analysis requirements related to fire protection would support either a traditional, deterministic approach or a more risk-informed approach where the risks from fires are addressed within the identification and analyses of LBEs. Section 53.460 would establish criteria for the safety classification of SSCs and determination of appropriate special treatments. As noted in subpart A, the term “
                        *Special treatments*
                        ” would be defined to mean those items, such as measures taken to satisfy functional design criteria, quality assurance, and programmatic controls, which provide assurance that certain SSCs will provide defense in depth or perform risk-significant functions. These requirements would also provide confidence that the SSCs will perform under the service conditions and with the reliability credited in the analysis performed in accordance with § 53.450 to satisfy the safety criteria in §§ 53.210 and 53.220. The terminology used in part 53 would include the following categories for SSC classification: (1) SR; (2) NSRSS; and (3) non-safety significant. Requirements for SR SSCs would be defined in other sections of 
                        ( print page 86930)
                        part 53 and would include using TSs for controls during operation and the application of quality assurance requirements from appendix B of part 50. Requirements for NSRSS SSCs would include the need to identify necessary special treatments such as performance measures on reliability. Licensees would generally be afforded flexibility in maintaining and changing special treatments for SSCs categorized as NSRSS. Non-safety-significant SSCs would be addressed under normal licensee programs for commercial grade equipment and typical industry practices for general plant design and maintenance. Safety-related SSCs would also contribute to defense in depth and risk-significant functions and may warrant special treatments beyond those defined for their SR functions to reflect their role in meeting the safety criteria in § 53.220 and the evaluation criteria in § 53.450(e). Section 53.470 would allow an applicant or licensee to seek operational flexibilities by adopting more restrictive criteria than those provided in § 53.220 and that might otherwise be used in the analysis of LBEs under § 53.450(e). Such an approach might be taken to ensure sufficient safety margins to gain operational flexibilities in areas such as justifying siting in relation to population centers or staffing levels. As an example, an applicant or licensee could propose to justify siting proposals by adopting alternate criteria for very unlikely event sequences. Such alternate criteria could require calculated consequences for an individual at the exclusion area boundary to be less than one rem total effective dose equivalent (TEDE). This section would establish requirements to ensure that, if more restrictive evaluation criteria than those required by a methodology were used to justify operational flexibilities, then the analysis, design features, and programmatic controls would be established and maintained accordingly. Section 53.480 would establish seismic design considerations. This proposed section would relate to the safety criteria in subpart B, the analytical requirements related to external hazards in § 53.450, and subpart D, “Siting Requirements.” For licenses issued under part 53, this section in subpart C would support a variety of approaches to seismic design. For example, a design for a commercial nuclear plant could show that SSCs are able to withstand the effects of earthquakes by adopting an approach similar to that in appendix S to part 50. Alternatively, an applicant could follow the more recent risk-informed alternatives afforded by standards development organizations (
                        *e.g.,*
                         American Society of Civil Engineers (ASCE)/Structural Engineering Institute (SEI) 43-19, “Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities.”) Because the agency has not endorsed ASCE/SEI-43-19, an applicant can propose to use ASCE/SEI 43-19 on an application specific basis to meet § 53.480 and the NRC would evaluate the adequacy of the standard as applied in that application. The design could also be done with the full integration of seismic PRAs into the design and licensing of a particular commercial nuclear plant. This section has been developed to accommodate a variety of potential risk-informed, performance-based seismic design approaches. The analyses required by § 53.450 would need to address seismic hazards as well as other external hazards. The expected responses of SSCs to a range of seismic events would be included in the analyses when ensuring that the safety criteria defined under § 53.220 would be met. The potential SSC responses to seismic hazards could be addressed in the analyses using a fragility model (conditional probability of its failure at a given hazard input level), a high confidence of low probability of failure value, or other method endorsed or otherwise found acceptable by the NRC. Subpart D—Siting Requirements Proposed subpart D in part 53 would state requirements for the siting of commercial nuclear plants and would serve the role provided by 10 CFR part 100, “Reactor Site Criteria,” for nuclear reactors licensed under parts 50 and 52. As reflected in proposed § 53.500, the reason for establishing siting requirements would remain the same as it has been historically, which is to ensure that licensees and applicants assess what impact the site environs may have on a commercial nuclear plant (
                        *e.g.,*
                         external hazards) and, conversely, what potential adverse health and safety impacts a commercial nuclear plant may have on nearby populations in view of the site characteristics. Proposed § 53.510 would require that design-basis external hazard levels be identified and characterized based on site-specific assessments of natural and constructed hazards with the potential to adversely affect plant functions. The site-specific assessments would be used in the proposed § 53.415, which would require that SR SSCs be designed to withstand the effects of natural phenomena and constructed hazards of levels or severities up to design-basis external hazard levels. The design-basis levels for external hazards relevant to a site would need to account for uncertainties and variabilities in data, models, and methods used to characterize those hazards. Existing approaches could be used to demonstrate compliance with this requirement. The historical importance of assessing seismic events as risks to commercial nuclear plants and the associated development of risk-informed approaches to address seismic events would be reflected in proposed § 53.480, “Earthquake engineering,” and specific requirements in subpart C. The NRC is developing a graded approach for seismic design by grouping SSCs into different seismic design categories (SDCs) based on their risk significance. While the agency has not endorsed ASCE/SEI-43-19, an applicant can propose to use ASCE/SEI 43-19 on an application-specific basis to meet § 53.480 and the NRC will evaluate the adequacy of the standard as applied in that application. The NRC staff will continue to review ASCE/SEI-43-19 as part of its efforts to further develop guidance in this area. The approach described in RG 1.208, “A Performance-Based Approach to Define the Site-Specific Earthquake Ground Motion,” would be an acceptable way to develop site-specific ground motion response spectra for SSCs under appendix S to part 50, which corresponds to SSCs that are categorized as the highest SDC (SDC-5) in ASCE/SEI 43-19. The evaluation of seismic hazards under subpart D would need to be sufficient to inform a site-specific design (
                        *e.g.,*
                         a CP or custom COL) or confirm the use of a standard design for a commercial nuclear plant under § 53.480 and other sections of subpart C. A risk-informed approach could use several design-basis ground motions (DBGMs) to assess SSCs in various SDCs (
                        *i.e.,*
                         one DBGM per SDC). Section 53.510(d) would state that geologic and seismic siting factors must also include related hazards such as seismically induced flooding and volcanic activity that may affect the design and operation of a proposed commercial nuclear plant for the proposed site. Section 53.520 would require applicants to identify and assess site characteristics related to topics which might include meteorology, geology, hydrology, or other areas in the design and analyses required under subpart C. Proposed section 53.530 would set requirements for population-related considerations and maintain requirements and definitions similar to those currently in part 100 for an exclusion area, low population zone, 
                        ( print page 86931)
                        and population center distance. The NRC recognizes that some applicants may propose to essentially collapse the exclusion area and low population zone to the site boundary. This approach would rest on a demonstration that the calculated consequences of DBAs remain below the proposed dose guidelines used in § 53.210, which are the same as those in the existing regulations in parts 50, 52, and 100. The proposed definitions in § 53.020 would allow such configurations, assuming they were justified by the design and analyses from subpart C. This approach should provide flexibility to justify alternative exclusion areas and low population zones without foreclosing the option for an applicant to define more conventional exclusion areas and low population zones outside of a defined site boundary. The NRC's long-standing preference for siting reactors in areas of low population density would be maintained in part 53 by using the current language from part 100 in proposed § 53.530(c). The NRC revised guidance related to population densities surrounding a commercial nuclear plant in Revision 4 to RG 4.7, “General Site Suitability Criteria for Nuclear Power Stations” to reflect Commission direction in SRM-SECY-20-0045, “Population Related Siting Considerations for Advanced Reactors.” Site-related requirements in part 20 (restricted area) and part 73 (protected and owner-controlled areas) would remain applicable to commercial nuclear plants licensed under part 53. Proposed section 53.540 would require that site characteristics be appropriately considered in other activities such as the design and analysis performed under proposed subpart D and the emergency planning and security programs under proposed subpart F. Subpart E—Construction and Manufacturing Requirements The proposed part 53 language would establish construction and manufacturing requirements in subpart E. The proposed language for construction-related activities would largely reflect current requirements in part 50 without any fundamental changes. Limited changes would be made in several places, as described in the following paragraphs, to be technology-neutral and for consistency with the organization and language of part 53. The proposed language for requirements for manufacturing activities would largely mirror those for construction-related activities. However, the proposed manufacturing requirements have been updated from the current requirements in subpart F of part 52 to better accommodate the possible factory fabrication of manufactured reactors. The manufacturing of specific components outside the scope of an ML would not be addressed by these proposed subparts. Section 53.600 would establish the overall construction and manufacturing requirements for CPs, OLs, COLs, MLs, and limited work authorizations (LWAs). This section would connect the construction and manufacturing requirements to the safety criteria, quality assurance requirements, and other requirements located in other subparts. These requirements would require that construction and manufacturing activities be managed and conducted such that when combined with associated design features and programmatic controls, the constructed plant would satisfy the relevant requirements in subpart B. Section 53.605 would establish requirements for the reporting of defects and instances of noncompliance during construction. This section would provide equivalent requirements to those in § 50.55(e). Section 53.610(a) would establish the requirement to have in place a well-defined command and control structure to manage construction activities. The requirements would generally reflect current requirements, with an emphasis on the quality assurance programs for complying with the requirements in appendix B to part 50. The proposed § 53.610(a)(6) would require programmatic controls for implementing special treatment for NSRSS SSCs to align with requirements in other subparts in part 53. The section would also refer to other NRC regulations to address matters such as requirements to have a FFD program, a radiation protection program if radioactive materials are brought onto the site, and security programs to protect sensitive information and protect against cyber threats. Section 53.610(b) would provide requirements governing construction activities, including the equivalent of the requirement in § 50.10(e) that prohibits starting construction until the NRC has authorized the activities by issuing a CP, COL, ESP, or LWA. Section 53.610(b)(1)(iii) would require procedures to be in place prior to beginning construction to ensure that construction-related activities do not undermine important features such as slope stability and that construction-related activities such as backfilling of excavated portions of the site appropriately address potential pre-construction activities such as the emplacement of retaining walls or drainage systems. Other requirements in these paragraphs would be equivalent to requirements in parts 50 and 52 with appropriate references to other parts for items such as possession of byproduct material or SNM, protecting operating units from construction activities for commercial nuclear plants with multiple reactor units, and having a redress plan in case LWA activities are terminated. Section 53.610(c) would address inspection and acceptance activities by including requirements in part 53 equivalent to specific quality assurance criteria in appendix B to part 50 and inspections, tests, analyses, and acceptance criteria (ITAAC) in part 52 for COLs. Section 53.620(a) would include proposed requirements covering the activities performed under an ML issued under part 53. Provisions related to MLs were first adopted by the NRC in 1973 through the addition of appendix M to part 50. The regulation supported the manufacture of a nuclear power reactor to be incorporated into a commercial nuclear plant under a CP and operated under an OL at a different location from the place of manufacture.[1]
                        
                         The regulations and processes for MLs were changed substantially in the part 52 rulemaking in 2007 (72 FR 49352). The most important shift in the ML concept in that rulemaking was that a final reactor design, which would be equivalent to that required for a standard DC under part 52 or an OL under part 50, must be submitted and approved before issuance of an ML. The rationale for that change was that approval of a final design ensures early consideration and resolution of technical matters before there is any substantial commitment of resources associated with the actual manufacture of the reactor, which greatly enhances regulatory stability and predictability. The proposed part 53 sections in subpart E for manufacturing and in subpart H for licensing matters would maintain requirements equivalent to those in part 52 for MLs. The NRC approval of a standard design and related manufacturing processes, coupled with a stable workforce and established procedures, has the potential for maintaining and even improving the quality and consistency of manufacturing, as compared to the traditional method of constructing 
                        ( print page 86932)
                        reactors onsite by a variety of contractors and subcontractors. Subpart E would include requirements that would apply to portions of a manufactured reactor in recognition that some activities covered by an ML may occur at different fabrication facilities. As with the preceding sections on construction, § 53.620 would establish the requirements to have in place programs, procedures, and a well-defined command and control structure to manage manufacturing-related activities. Section 53.620(b) in subpart E would propose requirements for executing the manufacturing activities following receipt of an ML under part 53. Information about the design and manufacturing processes should be provided by the applicant. The importance of the ML is reflected in several of the proposed requirements in § 53.620(b) that would refer to complying with the ML, including conducting manufacturing processes within facilities for which the license holder can control activities. The essential role of post-manufacturing inspections would also be incorporated into this proposed section by requiring the holder of the ML to perform inspections and have acceptance processes for manufactured reactors or portions of a manufactured reactor. Section 53.620(c) would provide proposed requirements for the control of radioactive materials if the holder of an ML plans to possess and use source, byproduct, or SNM as part of the manufacturing process. By and large, the proposed subpart E would refer to NRC regulations in 10 CFR part 30, “Rules of General Applicability to Domestic Licensing of Byproduct Material,” 10 CFR part 40, “Domestic Licensing of Source Material,” and part 70 for the requirements on controlling radioactive materials. Several specific requirements to address the potential hazards of radioactive materials are proposed in areas such as having a fire protection program, an emergency plan, training programs, and procedures to minimize contamination. The most significant change proposed for MLs in part 53 as compared to MLs under part 52 relates to § 53.620(d) in subpart E and the associated licensing provisions in subpart H. These provisions would allow and establish requirements for the loading of fuel into a manufactured reactor at the manufacturing site for subsequent transport to a commercial nuclear facility that will operate pursuant to a COL. The first requirement in the proposed § 53.620(d) would establish limitations on when a license under part 70 would authorize the loading of fuel into a reactor manufactured under an ML. The proposed regulation would require the manufactured reactor to include at least two independent physical mechanisms that will each prevent criticality should conditions most favorable to critical operation be introduced (
                        *e.g.,*
                         optimum neutron moderation and reflection). This requirement would contribute to the NRC's longstanding practice of requiring defense in depth for preventing accidents in any facility dealing with SNM, including requirements in § 70.64 for certain part 70 licensees to adhere to the “double contingency principle.” The requirements to have in place mechanisms to prevent criticality could likewise support meeting other provisions in subpart H to part 70, such as those related to having a safety program and integrated safety assessment. The mechanisms to preclude criticality in the proposed requirements would reasonably ensure that a manufactured reactor would not become critical assuming optimum neutron moderation, and optimum neutron reflection conditions. With the proposed requirements for mechanisms to prevent criticality and all criticality safety controls required by 10 CFR part 70 in place, the presence of fuel in the manufactured reactor would not create a nuclear hazard different than the hazard from the presence of the same fuel in a storage location or container licensed under 10 CFR part 70. Collectively, the proposed measures would reasonably ensure that the manufactured reactor would not be capable of operations, thereby obviating the need for a COL under §§ 53.1416 and 53.1440 to authorize fuel loading. Additionally, this approach would focus the ML application and its review on the design, manufacture, and deployment of the manufactured reactor. The activities involving SNM within the manufacturing facility, including the loading of fuel, would be regulated primarily under the part 70 license. The reference to the requirements in subpart H of part 70 in section 53.620(d) assures that the activities involving the receipt, storage, and loading of a variety of possible fuel forms and enrichments at the manufacturing facility will be analyzed in a systematic manner and appropriate protection will be provided against equipment malfunctions, human errors, external hazards, and other adverse conditions. The regulations in part 51 provide a flexible approach for environmental review to address the range of regulated activities under part 70. The flexibility in part 51 will enable the NRC to determine the appropriate type of environmental review based on the circumstances associated with the loading of fuel into a specific manufactured reactor. The proposed § 53.620(d) cites the requirements in parts 70, 71, and 73 to ensure important features and programs are in place prior to the receipt of SNM. The features and programs required to be in place prior to receipt of SNM include (1) radiation monitoring instrumentation and alarms; (2) measures to detect potential criticality accidents; (3) appropriate procedures, equipment, and personnel qualified for the fuel loading; (4) programs for physical security and cybersecurity; and (5) material control and accounting (MC&A) programs. Section 53.620(d)(2)(i) proposes requirements to address security programs for any ML authorizing possession of a manufactured reactor into which fuel has been loaded at the manufacturing facility. Currently, for category II SNM, security measures may be required in addition to requirements included in § 73.67, “Licensee fixed site and in-transit requirements for the physical protection of special nuclear material of moderate and low strategic significance,” on a case-by-case basis. Including appropriate security measures in the proposed part 53 regulations will provide additional openness and transparency for applicants applying for an ML who seek to load fuel into manufactured reactors at a manufacturing site. Currently, § 73.67 only requires a security plan for licensees who possess, use, transport, or deliver to a carrier for transport SNM of moderate strategic significance, or 10 kg or more of SNM of low strategic significance. However, the proposed physical security program for fueled manufactured reactors would require a security plan for any ML authorizing possession of a manufactured reactor into which fuel has been loaded at the manufacturing facility, regardless of fuel type, enrichment, and quantity. This is consistent with other controls for MLs, including reactivity and criticality controls. The proposed requirements would also require a holder of an ML and part 70 license to address cybersecurity to ensure a cyberattack would not adversely impact the functions performed by digital assets used by the licensee for physical security, radiation monitoring, or criticality prevention. The proposed regulations in part 53 covering the activities related to the storage, movement, and loading of fresh 
                        ( print page 86933)
                        fuel into a manufactured reactor in the manufacturing facility would likewise refer to the applicable regulations in part 70. The proposed § 53.620(d) would also require the loading or unloading of unirradiated fuel into or from a manufactured reactor and any changes to the configuration of reactivity-related systems to be performed by a certified fuel handler meeting the requirements in subpart F. The NRC is aware of proposals to introduce reprocessing of existing or future spent nuclear fuel into the fuel cycle for some potential commercial nuclear plants. This proposed rule does not address the loading of spent nuclear fuel or fuel resulting from reprocessing of spent nuclear fuel into a manufactured reactor. Section 53.620(e) would limit the transport and delivery of a manufactured reactor or portions of a manufactured reactor only to a site for which the Commission has issued a COL authorizing the construction of a commercial nuclear plant using a manufactured reactor under the specific ML. This proposed requirement is similar to the limitations in § 52.153, with the difference being that part 53 would allow the installation of a manufactured reactor at the site of a COL but would not include provisions for installation at a site under a CP. The possible combination of a manufactured reactor and the licensing option of CP and OL seems unlikely and would require the introduction of ITAAC into the licensing provisions for a CP and OL. An additional proposed paragraph in § 53.620(e) would provide requirements for protecting fueled manufactured reactors during transport to the site of the commercial nuclear plant by referencing the transportation and security requirements in 10 CFR part 71, “Packaging and Transportation of Radioactive Material,” and part 73. Section 53.620(f) would include proposed requirements for the acceptance and installation of a manufactured reactor at the site of a commercial nuclear plant. The proposed requirements would reference the construction requirements in § 53.610 to govern the integration of the manufactured reactor into the construction of a commercial nuclear plant. Other proposed requirements in the section would address required receipt inspections and verification that interface requirements between the manufactured reactor and the balance of the commercial nuclear plant have been met. Subpart F—Requirements for Operation Proposed subpart F would provide the requirements for the operations phase of a commercial nuclear plant to ensure that the safety criteria in subpart B are satisfied throughout the plant's lifetime and during all modes of normal operation and unplanned events. Section 53.700 would provide the overall objectives and general organization of subpart F, which would be to establish requirements during operations for: (1) plant SSCs; (2) plant personnel; and (3) plant programs. Proposed § 53.710 would provide the requirements for maintaining capabilities, availability, and reliability of SSCs to demonstrate compliance with the safety criteria and design requirements for unplanned events that are described in proposed subparts B and C. The basic structure of this proposed section would be that controls for SR SSCs are provided by TS and controls for NSRSS SSCs are required to be addressed with licensee-controlled documents and procedures. The general content and control of TS under the proposed part 53 would be similar to the requirements in part 50. The proposed requirements for TS would include limits on the inventories of radioactive materials, plant operating limits, and specific requirements for each SR SSC, including limiting conditions for operation (LCO) and required surveillances. The proposed requirements for TS would also include a section on important design elements, which is similar to design features in § 50.36, and a section for administrative controls. A provision addressing the development and submittal of TS to address decommissioning activities would also be included in the proposed subpart G. The proposed requirements for TS under part 53 would not carry over safety limits or associated limiting safety system settings from § 50.36, which contains TS requirements for operating reactors under parts 50 and 52. As discussed in SECY-18-0096, systematic assessments and more mechanistic approaches to evaluating source terms support an alternative approach to establishing barrier-based safety limits. An example provided in that paper is a comparison of: (1) the traditional specified acceptable fuel design limits (SAFDL) that support protecting a specific barrier from potential failure mechanisms (
                        *e.g.,*
                         departure from nucleate boiling to protect fuel cladding); and (2) the specified acceptable system radionuclide release design limit (SARRDL) concept, which limits the possible increase in circulating radionuclide inventory during normal operations or an AOO as part of an integrated or “functional containment” approach. Additional discussion of the use of SARRDL in the design and licensing of advanced reactors is provided in RG 1.232. The SARRDL could be addressed as an operating limit within this proposed construct of requirements for TS. In cases, such as LWRs, where a SAFDL approach might be used as part of a mechanistic approach to meeting the design and analysis requirements in subpart C, the associated functional design criteria proposed in § 53.410 and TS under the proposed § 53.710(a) would define similar requirements as those provided by the safety limit and limiting safety system setting requirements in § 50.36. The proposed requirements for TS under part 53 would not include specific criteria for identifying when LCOs must be established (
                        *i.e.,*
                         would not include an equivalent to § 50.36(c)(2)(ii)). Instead, consistent with subparts B and C, the TS requirements in subpart F of part 53 would define TS LCOs as providing limits on SR SSCs. The SR SSCs protect against DBAs to demonstrate compliance with the safety criteria in the proposed § 53.210. In the proposed construct for part 53, risk-significant SSCs would be addressed through a combination of TS for the SR SSCs and establishment and monitoring of performance standards for NSRSS SSCs. In addition to addressing TS for SR SSCs, proposed § 53.710 would require appropriate controls be developed and implemented for NSRSS SSCs. Examples include appropriate surveillances and controls established through reliability assurance programs. Configuration management and other special treatments would provide that the capabilities, availabilities, and reliabilities of NSRSS SSCs are maintained consistent with the underlying risk assessments while providing flexibility to licensees through maintaining the management functions within licensee-controlled programs. Controls on NSRSS SSCs are appropriate as part of the overall performance-based approach within proposed part 53. Special treatments beyond those defined for their SR functions may also be warranted for SR SSCs to reflect their role in meeting the safety criteria in § 53.220 and the evaluation criteria in § 53.450(e). The performance objectives for NSRSS SSCs would reflect that the comprehensive risk metrics and related risk performance objectives established under § 53.220 may involve assessing 
                        ( print page 86934)
                        and averaging the risks over a defined period (
                        *e.g.,*
                         plant year) and would not constitute a real-time requirement that must be continuously demonstrated by the licensee. The controls under § 53.710(b) justify proposed changes in part 53 from the traditional or deterministic approaches in parts 50 and 52 in areas such as replacing the single-failure criterion with a probabilistic reliability criterion (see SRM-SECY-03-0047, “Policy Issues Related to Licensing Non-Light-Water Reactor Designs,” dated June 26, 2003). This approach could also support the incorporation of risk insights and analytical margins to gain operational flexibilities in areas such as siting and staffing requirements described in subsequent sections of proposed subpart F. Proposed § 53.715 would provide the requirements for developing and implementing a program to do the following: (1) control maintenance activities; (2) take appropriate corrective action when performance issues are identified; (3) conduct routine evaluations of effectiveness; and (4) assess and manage risks resulting from maintenance activities. These proposed requirements are similar to those included in § 50.65 (maintenance rule), including the need to assess and manage the increase in risk that may result from the proposed maintenance activities. While, for the maintenance rule, specific criteria must be developed to capture both SR and non-SR but otherwise important SSCs, the proposed § 53.715 would cover SR SSCs and NSRSS consistent with other subparts in part 53. Proposed § 53.720 would provide the requirements for responding to a seismic event during the operating phase of the life cycle of a commercial nuclear plant and would be equivalent to the requirements in paragraph IV(a)(3) of appendix S, “Earthquake Engineering Criteria for Nuclear Power Plants,” to part 50. The proposed part 53 would include provisions to address staffing, training, personnel qualifications, and human factors engineering (HFE) in a manner that is risk informed, technology inclusive, performance based, and flexible in nature. During the development of part 53, the staff prepared a draft white paper on “Risk Informed and Performance Based Human-System Considerations for Advanced Reactors,” to support interactions with stakeholders and the ACRS. Key considerations include the recognition that staffing, operator qualifications, and HFE are interconnected areas that must be approached in an integrated manner and, furthermore, that safety functions, including the means by which they are fulfilled, provide an effective method for informing technology-inclusive requirements. The requirements associated with this approach would be in §§ 53.725 through 53.830. Section 53.725 discusses applicability and defines specific terms. Some definitions draw from those in § 55.4. Several new definitions would be introduced for use within the context of subpart F. These new definitions would be the following: “
                        *Automation,*
                        ” “
                        *Auxiliary operator,*
                        ” “
                        *Generally licensed reactor operator,*
                        ” “
                        *Interaction-dependent-mitigation facility,*
                        ” “
                        *Load following,*
                        ” “
                        *Self-reliant-mitigation facility.*
                        ” Sections 53.725 through 53.830 would be divided into four portions that would cover general operational requirements, operator and senior operator licensing requirements, generally licensed reactor operator (GLRO) requirements, and general training requirements for plant staff. The NRC intends to provide guidance addressing the review of operator staffing plans; the review of operator, senior operator, and GLRO examination programs; and the implementation of scalable HFE reviews. Licensees would be required to use GLROs upon demonstrating compliance with the criteria in § 53.800. Certain routine communications are necessary to facilitate the operator licensing process. The NRC is proposing to adapt the requirements of §§ 55.5 and 50.74 to § 53.726 to accomplish this. Specific information must be collected in order to facilitate the initial issuance of operator licenses, as well as to allow for license renewals and required updates thereafter. Such information collection activities must also be approved by the OMB. The NRC is proposing to adapt the requirements of § 55.8, to include any needed updates in OMB approval information, to § 53.120 to accomplish this. The information used within the regulatory processes of the NRC must be free from omissions and inaccuracies to facilitate effective regulation. Consistent with this, the NRC is proposing to adapt the requirements of § 55.9 to § 53.728 to require the completeness and accuracy of material information provided by individual applicants and license holders. Section 53.730 would provide performance-based and technology-inclusive requirements for assessing the role of personnel in facility safety, applying human-system considerations within facility design, and incorporating operational approaches that are consistent with design-specific safety considerations. Most of these requirements would be adapted from portions of §§ 50.34(f) and 50.54 and 10 CFR part 55, “Operators' Licenses,” with considerable modification in order to reflect the introduction of new technologies and possible changes in the roles of personnel in preventing and mitigating events. The NRC is proposing that these technical requirements would, together, serve as a component of the required content of applications for OLs and COLs under part 53. Additionally, the NRC proposes that the specific technical requirements associated with HFE, human-system interface design, concept of operations, functional requirements analysis, and function allocation would serve as a component of the required content of applications for standard DCs, standard design approvals, MLs, and CPs, as well. Human factors engineering is essential to facilitate the role of personnel in facility safety in a manner that is both effective and reliable. The NRC proposes to adapt § 53.730(a) from the HFE design requirements of § 50.34(f)(2)(iii). A key difference would be that the requirement would now be focused on settings where personnel fulfill their safety or emergency response roles wherever they may occur. The NRC additionally proposes to include within the scope of this requirement activities for assuring the continued availability of plant equipment that is needed for safety, and envisions that this may encompass relevant maintenance, inspections, and testing as well. The NRC intends that this requirement would be associated with staff guidance for conducting scalable reviews of HFE that is planned to accompany part 53. Human-system interfaces provide vital information to operators across a spectrum of operating conditions that can range from normal operations through severe accident conditions. The specific types of information that must be available to support operations staff during such conditions include, in part, those associated with safety function parameters, safety system status, possible core damage states, barrier integrity, and radioactive leakage. Due to the importance of such information, the NRC proposes under § 53.730(b) to require such human-system interface design features for all facilities, irrespective of other flexibilities proposed under part 53. Therefore, the NRC proposes to adapt specific post-Three Mile Island requirements of § 50.34(f) in a technology-inclusive manner as detailed in the following:
                        ( print page 86935) Paragraph (b)(1) would be adapted from § 50.34(f)(2)(iv).
                    Paragraph (b)(2) would be adapted from § 50.34(f)(2)(v).
                    Paragraph (b)(3) would be adapted from § 50.34(f)(2)(xi), 50.34(f)(2)(xii), and 50.34(f)(2)(xxi).
                    Paragraph (b)(4) would be adapted from § 50.34(f)(2)(xvii), 50.34(f)(2)(xviii), 50.34(f)(2)(xix), and 50.34(f)(2)(xxiv).
                    Paragraph (b)(5) would be adapted from § 50.34(f)(2)(xxvi).
                    Paragraph (b)(6) would be adapted from § 50.34(f)(2)(xxvii). In addition to the requirements of § 53.730(b)(1) through (6), a further set of human-system interface design requirements applicable only to those facilities that will be staffed by GLROs would be provided under § 53.730(b)(7). This prescriptive set of design requirements for those facilities which demonstrate compliance with the criteria of § 53.800 would recognize that the application of HFE under § 53.730(a) is anticipated to be significantly reduced at such facilities in the absence of an expected operator role for the fulfillment of safety functions. However, it should be noted that the capability for an immediately initiated, manual reactor shutdown would be conservatively mandated irrespective of any other design considerations. The NRC proposes § 53.730(c) to require the submittal of a concept of operations that is of sufficient scope and detail to appropriately inform the staff. The development of a concept of operations can facilitate a clear understanding on the part of the NRC for potential novel operating concepts. Additionally, such information is likely to reduce the degree of resources and interactions needed for the NRC to obtain the understanding necessary to enable flexible requirements in areas such as staffing, operator qualifications, and HFE. The NRC proposes § 53.730(d) to require the submittal of both a Functional Requirements Analysis and a Function Allocation. The identification of design-specific safety functions and how they are fulfilled serves as a primary means for achieving technology-inclusive requirements within areas such as staffing, operator qualifications, and HFE. The Functional Requirements Analysis and Function Allocation processes (which are both HFE methods derived from systems engineering principles), provide an effective means to identify both how safety functions will be satisfied and how to characterize any associated operator role in doing so. A Functional Requirements Analysis shows what features, systems, and human actions are relied upon to demonstrate safety (
                        *i.e.,*
                         fulfill safety functions). A Function Allocation then describes how safety functions are assigned to both personnel and automatic systems. However, an important adaptation of the Function Allocation for use under the proposed rule would be the further need to not only describe allocations of safety functions to human action and automation, but also to identify allocations made to active safety features, passive safety features, or inherent safety characteristics as well. Operating experience provides an important source of information by which to inform various aspects of facility design and operations. Accordingly, the NRC proposes in § 53.730(e) to adapt the requirements of § 50.34(f)(3)(i) for requiring an operating experience program. New technologies may involve concepts of operations that are more conducive to customizable licensed operator staffing requirements than the prescriptive requirements of § 50.54(m). Analyses and assessments that are based on HFE principles provide a performance-based means of determining licensed operator and senior operator staffing needed to support safe operations. In contrast, for those facilities required to be staffed by GLROs, the NRC anticipates that the operator staffing plans will reflect a simpler approach of showing that a continuity of responsibility will be maintained for facility operations throughout the operating phase, with at least one GLRO providing continuous oversight and remaining immediately available when any units are fueled. Additionally, a revised approach to the traditional position of the shift technical advisor that focuses on the availability of engineering expertise as a means of addressing uncertainties and abnormal circumstances is more suitable within the context of part 53 and is intended to be applicable to all facilities, irrespective of other design and staffing considerations. Consistent with this approach, the NRC proposes under § 53.730(f) to require the submittal of a staffing plan that details operations staffing, how engineering expertise will be provided, and what staffing will be available to provide other needed support functions. The NRC intends that this requirement would be associated with staff guidance for reviewing operations staffing plans that is planned to accompany part 53 and that, following NRC approval of the OL or COL, the staffing plan would become a condition of the facility license. The NRC intends that, at a minimum, the approved licensed operator and senior operator (or, if applicable, GLRO) staffing, positions, and personnel locations will be incorporated into corresponding requirements within the facility TS and that a license amendment would thus be required for any subsequent changes. Operator training and qualification programs provide an essential component of supporting human performance in implementing tasks with safety implications. Such programs must include components that cover the stages of initial training, examination, and continuing training. Additionally, recognizing the potential for varying concepts of operations to affect traditional, prescriptive approaches to operator proficiency, the NRC proposes under part 53 to allow facilities to develop operator proficiency programs based on facility-specific considerations. Therefore, the NRC proposes in § 53.730(g)(1) to require approval as part of its approval of the OL or COL, of the programs that will be used for the initial training, initial examination, requalification training and examination, and proficiency of both licensed operators and senior operators. In a corresponding manner, the NRC proposes in § 53.730(g)(2) to require approval of the programs that will be used for the GLRO equivalents of each of these programs for facilities with such staffing. The NRC intends that examination program requirements would be associated with staff guidance for the review of tailored examination processes that are planned to accompany part 53. Following the completion of an initial training program, continuing training programs provide an important means of sustaining the knowledge and abilities of individuals. The NRC is proposing to adapt the requirements of § 50.54(i-1) in § 53.730(g)(3) to require that operator continuing training programs be in effect to support operator performance. Under part 53, the NRC proposes to require these programs to be in effect concurrent with when the initial operator examinations first commence, in effect putting the programs in place only when they are needed. This represents a modification of the comparable requirement of § 50.54(i-1), which links the commencement of these programs to a timeline driven by the licensing of the facility. The authorization to manipulate controls of the facility that directly affect reactivity or power level is restricted to individuals who are either licensed operators, licensed senior operators, or GLROs. However, for practical purposes, situations in which 
                        ( print page 86936)
                        an individual is participating in an approved training program or reestablishing proficiency may also call for them to operate the controls of the facility under the cognizance of a licensed individual. The NRC is proposing to adapt the requirements of § 55.13 in § 53.735 to accomplish this, with a notable difference being the incorporation of GLROs. Section 53.740 would provide requirements for OL and COL holders under part 53. Portions of § 53.740 would be adapted from the conditions of § 50.54. In general, the conditions for operations staffing under part 53 would reflect considerations for potential technological differences and varying concepts of operation that are expected among part 53 facility licensees. Additionally, certain requirements would be specific to the operating phase while others would remain in effect following the permanent cessation of facility operations during the decommissioning phase. All commercial nuclear plants licensed under part 53 would require some form of licensed operator staffing, whether it be by specifically or generally licensed operators. Consistent with this, the NRC is proposing under § 53.740(a) to require facility licensees to demonstrate compliance with the programmatic requirements for either specifically licensed operators and senior operators or for GLROs, as applicable to the facility. The NRC recognizes that technology-inclusive facility staffing will need to account for a potentially wide range of concepts of operations; for this reason, flexible and performance-based approaches for establishing required facility staffing are appropriate. However, once the appropriate facility staffing has been determined and approved by the NRC, such staffing must be maintained to ensure that the appropriately qualified individuals will be available when needed to support the safe operation of the facility. Therefore, the NRC is proposing under § 53.740(b) to require that the staffing described within the approved facility staffing plan be maintained as a condition of the facility license as opposed to prescriptive staffing requirements like those of § 50.54(k) and (m). Because operation of facility controls directly affects reactivity or power level, only those individuals who possess appropriate levels of qualification and authorization are permitted to operate those controls. The NRC is proposing to adapt the requirements of § 50.54(i) in § 53.740(c) to require that only specifically licensed operators and senior operators or, alternatively, GLROs, may operate facility controls, with allowance for specified exceptions for the purposes of operator training or proficiency. Senior operators, by virtue of their license level, are qualified and authorized both to perform certain important responsibilities and to direct the licensed activities of licensed operators. Therefore, facilities that are required to be staffed by specifically licensed operators must also include senior operators within their staffing. In contrast, facilities staffed with GLROs only have a single license level available and, therefore, there is no equivalent provision for such facilities. The NRC is proposing to adapt the requirements of § 50.54(l) in § 53.740(d) to require the licensing and designation of senior operators at facilities staffed by specifically licensed operators. In contrast with control manipulations that directly affect reactor power and reactivity (
                        *e.g.,*
                         control rod movement, control drum rotation, recirculation pump speed adjustment, reactor coolant system boration or dilution, etc.) and are therefore restricted to performance only by licensed operators, other types of plant operations that may result in reactor power and reactivity changes via means that are indirect in nature (
                        *e.g.,*
                         electrical generation changes, turbine bypass valve operation, steam usage by process heat applications, etc.) may be implemented by non-licensed personnel. However, due to the potential influence of such operations on reactor power and reactivity, the continuous oversight of reactor parameters by a licensed operator is necessary during these operations. The NRC is therefore proposing to adapt the requirements of § 50.54(j) in § 53.740(e) to require appropriate oversight of operations, other than those associated with the controls themselves, that may affect reactivity or power level. Load following where plant output automatically changes in response to externally originated instructions or signals is not permitted under the existing regulations of § 50.54. However, new technological considerations and concepts of operation may justify such an operational approach under appropriate circumstances. The NRC recognizes that, beyond electrical power generation, load following may also affect other applications of plant output, such as hydrogen production, desalination, or district heating. For load following to be permissible, measures must be in place to provide assurance that plant output considerations are not permitted to lead to challenges to safe reactor operations. These measures may consist of automated control systems, automatic protective features, or the continuous oversight and immediate intervention capability of an appropriately qualified and authorized individual. Section 53.740(f) would allow for load following, provided that appropriate measures are in place. In considering the acceptability of the measures associated with load following, the NRC expects that any automatic protection relied upon would be separate from that credited for reactor protection purposes and would employ setpoints that are set so as to prevent actuation of the reactor protection system while accomplishing its functions to the extent practical. Core alterations such as refueling are associated with specific considerations that warrant limiting the oversight of such operations to appropriately qualified and authorized individuals. Unlike other types of fuel handling operations, core alterations occur within the confines of a reactor vessel that is specifically designed to support and sustain nuclear criticality, thereby justifying the imposition of higher qualification levels within such contexts. The NRC is proposing to adapt the requirements of § 50.54(m)(2)(iv) in § 53.740(g) to require the supervision of core alterations by either a specifically licensed senior operator, a specifically licensed senior operator whose license is limited to fuel handling, or by a GLRO, as applicable to the facility. Because certain commercial reactor designs may be capable of refueling while at power and, in any event, overall facility oversight would already be required by either a specifically licensed senior operator or by a GLRO, the NRC proposes to omit this requirement as redundant during periods where core alterations occur while the plant is operating. It is impossible to predict every possible scenario that a commercial nuclear plant might potentially encounter. Therefore, it is prudent to grant the authority for appropriately qualified individuals to depart from facility license conditions when emergency circumstances dictate that doing so is in the interest of public health and safety. The NRC is proposing to adapt the requirements of § 50.54(x) and (y) in § 53.740(h) to permit specific individuals to authorize departures from facility license conditions or TSs when emergency conditions warrant doing so for the protection of the public health and safety. Recognizing that certain facilities licensed under part 53 may be staffed by GLROs in lieu of specifically licensed senior operators, the NRC proposes to extend this authority to 
                        ( print page 86937)
                        GLROs. While it is not anticipated that GLROs will have a role in the fulfillment of safety functions at self-reliant-mitigation facilities and, furthermore, that operators at such facilities would not be in a position by which to significantly influence radiological safety outcomes, the very nature of the § 50.54(x) and (y) and the proposed § 53.740(h) provisions concern situations that are unanticipated and, therefore, unforeseeable. Thus, it is appropriate to grant GLROs a comparable authority to that of senior licensed operators and certified fuel handlers as it relates to invoking this provision under emergency conditions as a means of accounting for such possibilities. Due to the unique authorities and responsibilities of both specifically and generally licensed reactor operators, it is essential that any individual fulfilling such a role demonstrate compliance with the regulatory requirements for operator licensing. Section 107 of the Act authorizes the Commission to prescribe conditions for the licensing of operators and to issue licenses consistent with those conditions. The NRC is proposing to adapt the requirements of § 55.3 in § 53.745 to require that any person performing the function of an operator, senior operator, or GLRO must be authorized by a license issued by the Commission. The NRC proposes to license individuals as operators under both specific and general licensing frameworks. Specific licenses would be for licensed operators (
                        *i.e.,*
                         reactor operators) and senior operators (
                        *i.e.,*
                         senior reactor operators) and would be issued to a named person upon approval by the Commission of an application for that named person. In contrast, GLROs would perform duties under the provisions of a general license that would be effective without the filing of an application with the Commission or the issuance of licensing documents to a particular person. The NRC proposes requirements for the use of a specific licensing process for licensed operators and senior operators under §§ 53.760 through 53.795, with § 53.760 addressing applicability. Medical fitness is an important component of the overall process of specifically licensing operators because it provides assurance that operators will be able to carry out important duties without being precluded from doing so by health-related issues. Medical fitness also provides assurance that such issues will not adversely affect the performance of assigned job duties or cause operational errors that endanger public health and safety. In addition to a requirement for medical fitness, a medical examination by a physician to confirm compliance with this requirement is necessary. The NRC is proposing to adapt the requirements of §§ 55.21, 55.23, and 55.27 under § 53.765 to require medical fitness, examinations by physicians, and medical certification for specifically licensed operators and senior operators. In recognition of the fact that GLROs are not expected to have a role in the fulfillment of safety functions at the facilities at which they are licensed, the NRC proposes to not extend a comparable medical requirement to GLROs. The NRC is also proposing to adapt the requirements of §§ 55.25 and 50.74(c) in § 53.770 to require that timely notifications be made to the NRC if a specifically licensed operator or senior operator develops a permanent physical or mental condition that adversely affects the performance of assigned operator job duties or could cause operational errors endangering public health and safety. Notwithstanding this requirement related to permanent medical conditions, the NRC continues to recognize that it is appropriate for facility licenses to impose administrative restrictions and conditions upon specifically licensed operators and senior operators in response to temporary medical conditions. The process of specifically licensing individuals as licensed operators or senior operators requires the submittal of applications to the NRC for review. These applications must detail certain elements associated with licensing, including the demonstration of compliance with examination, experience, and medical requirements. The NRC is proposing to adapt the requirements of §§ 55.31 through 55.35 in § 53.775 to include requirements for the applications associated with the specific licensing of licensed operators and senior operators at commercial nuclear plants licensed under part 53. In contrast with the part 55 requirements, the NRC proposes to provide additional flexibility by locating certain details associated with the preparation and submittal of these applications within guidance in lieu of placement within this proposed rule itself. The NRC proposes overall programmatic requirements for specifically licensed operator and senior operator training, examination, and proficiency in § 53.780. In general, the proposed requirements are adapted from those in part 55, with several additional flexibilities being incorporated to better account for potential variations in reactor technologies and concepts of operations. The requirements proposed in § 53.780 cover, in part, the initial training, initial examination, requalification training, requalification examination, and proficiency of specifically licensed operators and senior operators. The initial training process provides individuals with the knowledge and abilities needed to subsequently fulfill assigned duties as licensed operators or senior operators in a safe and reliable manner. The use of a systems approach to training (SAT) ensures that the training program is based upon job requirements in a manner that can be adapted to account for differences in plant technology, concepts of operations, and operator roles in the fulfillment of design-specific safety functions. The NRC is proposing under § 53.780(a) to require facility licensees to implement a SAT-based training program for the initial training of licensed operator and senior operator applicants. The program must be adequate to ensure that applicants will be capable of performing the duties necessary both to protect public health and safety and to maintain plant safety functions. The NRC further proposes that such programs be subject to NRC approval and subsequent change control processes of an appropriate nature. Examinations provide a means of assessing that individuals have achieved a degree of knowledge and ability that is sufficient to carry out assigned duties as licensed operators or senior operators in a manner that is safe and reliable. The NRC is proposing to adapt the requirements of §§ 55.40, 55.41, 55.43, and 55.45 in § 53.780(b) to require that facilities establish and implement an initial examination program. However, a key difference from the comparable requirements of part 55 would be that facilities have the flexibility to propose, subject to NRC approval, the examination methods and criteria to be used in assessing satisfactory applicant performance. Such examination programs (including those used within the scope of requalification training) would need to provide for acceptable levels of both test validity and test reliability in order to be considered acceptable. The NRC intends that staff guidance would be available to facilitate the review of licensing examination programs that are proposed by facility licensees and that, following NRC approval, initial examination programs would be subject to an appropriate change control process. Furthermore, the NRC proposes that holders of licenses to operate commercial nuclear 
                        ( print page 86938)
                        plants under part 53 be provided the alternative of administering their own approved licensing examinations. The NRC would continue to exercise appropriate oversight of the program, make operator licensing decisions based upon the examination results, and reserve the right to administer the examinations in lieu of permitting the facility to do so. However, irrespective of the provided flexibilities in examination format and structure, at a minimum, topics from the following general categories of knowledge and abilities should be sampled in such examinations: Reactor Theory, Thermodynamics, and Chemical Interactions
                    Plant Systems and Components
                    Reactivity Management and Manipulations
                    Radiation Control and Safety
                    Emergency, Abnormal, and Normal Operations
                    Administrative Requirements and Conditions of the Facility License Requalification training programs provide for the continuing training and examination of specifically licensed operators and senior operators to ensure that they maintain the knowledge and abilities needed to support the safe and reliable performance of job duties following the completion of an initial training and examination program. The NRC is proposing to adapt the requirements of § 55.59 in § 53.780(c) to require that facilities implement both a SAT-based requalification training program and a biennial requalification examination program. However, a notable difference from the biennial requalification examinations required under part 55 would be that distinct annual operating test and biennial written examination components would not be mandated, with the facility licensee instead proposing the examination methods and criteria to be used in assessing satisfactory performance. The NRC intends that guidance would be available to facilitate the review of the requalification examination programs that are proposed by facility licensees and that, following NRC approval, requalification examination programs would be subject to an appropriate change control process. For examinations to provide for valid assessments of the knowledge and abilities of individuals, the examinations must remain free from compromises that could affect their underlying integrity. The NRC is proposing to adapt the requirements of § 55.49 in § 53.780(d) to require that examinations and related activities remain free from any compromise that might affect the integrity of the examination process. Simulators provide a valuable means of training and evaluating plant operators, and the NRC is specifically authorized under the Nuclear Waste Policy Act of 1982, as amended (NWPA), section 306 (42 U.S.C. 10226) to establish regulations for the use of simulators within such context. The NRC is proposing to adapt the requirements of § 55.46 in § 53.780(e) to address the use of simulation facilities for training, examinations, and applicant experience requirements, as well as to address the maintenance of simulator fidelity. However, the proposed requirements of part 53 would not mandate that full scope, plant-referenced simulators be used and would allow the use of alternative simulation facilities consisting of, for example, partial scope simulators or the plant itself, provided that all associated requirements can be demonstrated to be met using alternative approaches and methods. Additionally, in allowing for the possibility that an applicant or licensee might demonstrate compliance with training, examination, or experience requirements using the plant itself, the NRC is not allowing the initiation of transients on the actual plant. Consistent with this, aside from controlled reactivity manipulations that are conducted for the purposes of demonstrating compliance with experience requirements, actual plant components may not be operated for these purposes. Rather, the NRC perspective is that the use of the plant for training and examination purposes should be restricted to techniques such as walkthroughs, job performance measures, simulated tasks, use of augmented reality technology, and similar approaches that provide training and examination value while avoiding the operation of actual plant components. There may be situations in which applicants for operator or senior operator licenses have previous training and experience that justifies waiving some, or all, of the initial examination requirements. The NRC is proposing to adapt the requirements of § 55.47 in § 53.780(f) to allow for consideration of requests for waivers of examinations requirements. In contrast with the part 55 requirements, the NRC proposes to locate certain details associated with such waiver requests within guidance documentation in lieu of placement within the rule itself. For licensed operators and senior operators to perform their assigned duties safely and reliably, it is essential that they perform those duties frequently enough so as to maintain a sufficient degree of proficiency. The NRC is proposing to adapt the requirements of § 55.53(e) and (f) in § 53.780(g) to require that specifically licensed operators and senior operators maintain proficiency and, if proficiency is not maintained, regain proficiency prior to resuming licensed duties. However, in recognition of the fact that varying concepts of operations are possible for advanced reactor facilities, the NRC is proposing, in contrast with the requirements of part 55, to allow facility licensees to establish their own programs for operator proficiency, subject to NRC approval. As the holders of specific licenses, licensed operators and senior operators must be subject to license conditions on an individual basis to ensure that the basis upon which the licenses were issued remains valid. The NRC is proposing to adapt the requirements of § 55.53 in § 53.785 to require appropriate conditions of licenses for specifically licensed operators and senior operators. However, in contrast with the requirements of § 55.53(e) and (f), the NRC is proposing to allow certain aspects of operator proficiency to be addressed by an NRC-approved facility proficiency program. Licenses for specifically licensed operators and senior operators are issued by the NRC and must remain subject to modification or revocation. The NRC is proposing to adapt the requirements of §§ 55.51 and 55.61 in § 53.790 to address the issuance, modification, and revocation of licenses issued to specifically licensed operators and senior operators. The licenses issued to specifically licensed operators and senior operators are valid for a period of six years, after which they expire, unless otherwise renewed. The NRC is proposing to adapt the requirements of §§ 55.55 and 55.57 in § 53.795 to address the expiration and renewal of licenses issued to specifically licensed operators and senior operators. In developing this proposed rule, the NRC has discussed with stakeholders the considerations that might justify the omission of the specifically licensed operators and senior operators. However, even for an inherently safe reactor with autonomous operation features, certain important administrative functions (
                        *e.g.,*
                         compliance with TS, operability determinations, NRC notifications, emergency declarations, risk assessment, maintenance oversight, and radiological release limit compliance) would still need to be accomplished by 
                        ( print page 86939)
                        appropriately qualified and authorized individuals. Additionally, the NRC recognized that manual manipulations of facility reactivity controls must only be performed by individuals who have been appropriately licensed by the Commission. The NRC therefore proposes under § 53.800 to establish a new class of facility (defined as a self-reliant-mitigation facility), according to the criteria contained in § 53.800 for part 53. These facilities would employ GLROs rather than specifically licensed operators and senior operators. The GLRO regulations offer enhanced flexibilities and targeted relaxations in a manner that is commensurate with the modified role of such operators to ensure the safe operation of the associated facilities. In contrast, those facilities not meeting the criteria of § 53.800 would instead be considered interaction-dependent-mitigation facilities and would require staffing by specifically licensed operators and senior operators. The terminology used to designate these facility types reflects differences in how operators are anticipated to need to interact with their plant systems in mitigating events and achieving safe outcomes; such systems may either need operators to interact with them in some manner (
                        *i.e.,*
                         be interaction-dependent) or may instead be able to rely fully upon their own capabilities independent of operator interaction (
                        *i.e.,*
                         be self-reliant). Generally licensed reactor operators would differ from specifically licensed operators because the latter would be directly and independently evaluated by the NRC as part of their licensing process. This direct and independent evaluation remains appropriate when operators may reasonably be expected to exert a significant influence on public health and safety outcomes. Therefore, a key determinant as to whether generally licensed reactor operators can be utilized in facility staffing is the assessment of the operator's role in maintaining and fulfilling safety functions at the facility, such as through the performance of credited actions for the mitigation of plant events. The criteria proposed in § 53.800 would designate self-reliant-mitigation facilities. These criteria are derived from the following set of considerations: no human action needed to satisfy radiological consequence criteria;
                    no human action needed to address LBEs;
                    safety functions not allocated to human action;
                    reliance upon robust and highly reliable safety features; and
                    adequate defense in depth achieved without reliance on human action. It should be noted that those facilities not meeting the criteria proposed in § 53.800 would instead be classified as interaction-dependent-mitigation facilities and would require staffing by specifically licensed operators and senior operators instead. Generally licensed reactor operators would perform duties under the provisions of a general license that would be effective without the filing of an application with the Commission or the issuance of licensing documents to a particular person. The NRC proposes requirements for the general licensing process for GLROs under §§ 53.805 through 53.820. The requirements for GLROs would parallel those for senior operators in regard to their comparable administrative responsibilities. Nonetheless, the requirements for GLROs would be relaxed and incorporate greater flexibilities compared to the requirements for specifically licensed operators in a manner that is consistent with the GLRO's role in safety at self-reliant-mitigation facilities. In order to use GLROs in lieu of specifically licensed operators and senior operators, a OL/COL applicant would need to demonstrate that its proposed facility is a self-reliant-mitigation facility, 
                        *i.e.,*
                         that it will comply with the following requirements on an ongoing basis: maintaining GLRO qualifications for the performance of important functions and tasks; incorporating relevant programmatic controls into TS; administering the related programs for training, examination, and proficiency; and ensuring that the relevant provisions of parts 26 and 73 are met. Additionally, to provide for an accurate accounting of what individuals are licensed under the general license, facility licensees would be required to report the identities of all generally licensed reactor operators to the NRC on an annual basis. Furthermore, a facility licensee must ensure that the facility design and performance continue to meet the technological criteria to be classified as a self-reliant-mitigation facility (
                        *i.e.,*
                         the criteria of § 53.800) on a continual basis during the operating phase, as the relaxations afforded to such facilities in the areas of operator licensing, staffing, and HFE would be predicated on this assumption. The NRC therefore proposes under § 53.805 to establish requirements for facility licensees that address issues such as these. Finally, the failure of a self-reliant-mitigation facility to subsequently meet the criteria of § 53.800 after the issuance of an OL or COL would constitute a reportable event (
                        *i.e.,*
                         an unanalyzed condition that significantly degrades plant safety) under the provisions of § 53.1630. The NRC proposes the general license for GLROs under § 53.810. GLROs would be licensed as a class of individuals under the provision of § 53.810(a) and would be subject to the conditions specified in § 53.810(b) through (g). Portions of these conditions are adapted from § 55.53 and from those conditions currently included in the licenses issued to specifically licensed operators and senior operators. The NRC would retain the ability to suspend or prohibit individuals from operating under the general license should such action be warranted. The NRC proposes overall programmatic requirements for GLRO training, examination, and proficiency under § 53.815. In general, these proposed requirements are adapted from those of part 55 and parallel those also proposed for specifically licensed senior operators in § 53.780. These requirements include increased flexibilities and several targeted relaxations that reflect the limited role of GLROs in facility safety. The requirements proposed under § 53.815 cover, in part, the initial training, initial examination, continuing training, requalification examination, and proficiency of GLROs. Section 53.805 would require the facility licensee to develop, implement, and maintain these programs. Section 53.810, in turn, would prescribe that the requirements of § 53.805 would need to be met as a requirement of the general license. The implication of this structure is that the facility licensee would need to implement these programs for training, examination, and proficiency, and GLROs would need to participate in these programs to demonstrate compliance with the requirements of the general license. The initial training process provides GLROs with the knowledge and abilities needed to fulfill assigned duties as GLROs. The use of a SAT serves to ensure that the training program is based upon job requirements in a manner that can be adapted to account for differences in plant technology and concepts of operations. The NRC is proposing under § 53.815(b) to require facility licensees to implement a SAT-based training program for the initial training of GLROs that is adequate to ensure that they have the necessary knowledge, skills, and abilities to perform their duties. The NRC further proposes that such programs would be subject to NRC approval, oversight, and appropriate change control processes. The training program must ensure that 
                        ( print page 86940)
                        GLROs maintain the necessary knowledge, skills, and abilities. Examinations provide a means of assessing that individuals have achieved a degree of knowledge and ability that will be sufficient to enable them to carry out assigned duties as GLROs in a manner that is both safe and reliable. The NRC proposes to adapt the requirements of §§ 55.40, 55.41, 55.43, and 55.45 in § 53.815(b) to require that facility licensees establish and implement an initial examination program. A key difference from the comparable requirements of part 55 would be that facility licensees would be afforded the flexibility to propose, subject to NRC approval, the examination methods and criteria to be used in assessing satisfactory individual performance. Such examination programs (including those used within the scope of continuing training) would need to provide for acceptable levels of both test validity and test reliability in order to be considered acceptable. The NRC intends that staff guidance would be available to facilitate the review of initial examination programs that are proposed by facility licensees and that approved initial examination programs would be subject to an appropriate change control process. In contrast with both the requirements of part 55 and the proposed requirements of § 53.780, the NRC does not intend to administer or evaluate these initial examinations. However, the examination processes themselves will continue to be subject to ongoing NRC oversight. Irrespective of the provided flexibilities in examination format and structure, topics from the following general categories of knowledge and abilities should be sampled in such examinations: Reactor Theory, Thermodynamics, and Chemical Interactions
                    Plant Systems and Components
                    Reactivity Management and Manipulations
                    Radiation Control and Safety
                    Emergency, Abnormal, and Normal Operations
                    Administrative Requirements and Conditions of the Facility License Continuing training programs provide the ongoing training and examination of GLROs to ensure that they maintain the knowledge and abilities needed to support the safe and reliable performance of job duties following the completion of an initial training and examination program. The NRC is proposing to adapt the requirements of § 55.59 in § 53.815(b) to require that facility licensees implement both a SAT-based continuing training program and a requalification examination program. However, a notable difference from the examinations required under part 55 would be that distinct annual operating test and biennial written examination components would not be mandated. The facility licensee would instead propose examination methods and criteria to be used in assessing satisfactory performance. Furthermore, unlike the comparable requirements of part 55 and those proposed for specifically licensed operators and senior operators, a biennial periodicity for requalification examinations would not be prescribed. However, adequate justification for the proposed periodicity of requalification examinations would be required. The NRC intends that staff guidance would be available to facilitate the review of the requalification examination programs that are proposed by facility licensees. Approved requalification examination programs would be subject to an appropriate change control process. For examinations to provide for valid assessments of the knowledge and abilities of individuals, the examinations must remain free from compromises that could affect their underlying integrity. The NRC is proposing to adapt the requirements of § 55.49 in § 53.815(d) to require that examinations and related activities remain free from any compromise that might affect the integrity of the examination process. Simulators provide a valuable means of training and evaluating plant operators and the NRC is specifically authorized under the NWPA, section 306 (42 U.S.C. 10226) to establish regulations for the use of simulators within such context. The NRC is proposing to adapt the requirements of § 55.46 in § 53.815(e) to address the use of simulation facilities for training and examinations, and experience requirements, as well as to address the maintenance of simulator fidelity. The use of full scope, plant-referenced simulators would not be mandated. The potential use of alternative simulation facilities consisting of, for example, partial scope simulators or the plant itself, would be allowed provided that all associated requirements could be demonstrated to be met using alternative approaches and methods. Additionally, in allowing for the possibility that an applicant or licensee might demonstrate compliance with training and examination requirements using the plant itself, the NRC is not allowing the initiation of transients on the actual plant. Consistent with this, aside from controlled reactivity manipulations that are conducted for the purposes of demonstrating compliance with experience requirements, actual plant components may not be operated for these purposes. Rather, the use of the plant for training and examination purposes should be restricted to techniques such as walkthroughs, job performance measures, simulated tasks, use of augmented reality technology, and similar approaches that provide training and examination value while avoiding the operation of actual plant components. There may be situations in which GLROs have previous training and experience that justifies waiving some, or all, of the initial examination. Therefore, the NRC is proposing under § 53.815(f) to allow facility licensees to waive some, or all, portions of initial examinations provided that such waivers are consistent with a program that has been approved by the NRC. For GLROs to safely and reliably perform their assigned duties, it is essential that they perform those duties frequently enough so as to maintain a sufficient degree of proficiency. However, the NRC recognizes that facilities that utilize GLROs may have concepts of operation that warrant unique proficiency considerations. Therefore, the NRC is proposing in § 53.815(g) to require that facility licensees develop, implement, and maintain programs to maintain and reestablish, if needed, the proficiency of GLROs. This could occur, for example, if an individual's extended absence from watch standing has rendered proficiency requirements unmet. The general license should remain in effect for an individual only while that individual remains employed in a position that may call for the individual to manipulate the reactivity controls of the facility. The NRC proposes under § 53.820 to require that the general license would cease to be applicable on an individual basis when an individual's employment status becomes such that this is no longer the case. However, the NRC recognizes that for some types of self-reliant-mitigation facilities, very long periods may elapse between circumstances that necessitate manual manipulation of reactivity controls. Therefore, the general license remains in effect for an individual as long as the individual's current position could potentially require that individual to manipulate reactivity controls at some point within the course of the individual's assigned job duties. The NWPA, section 306 (42 U.S.C. 10226) authorizes and directs the NRC to, in part, issue regulations and guidance that address the training and 
                        ( print page 86941)
                        qualifications of civilian nuclear power plant operators, supervisors, technicians, and other appropriate operating personnel. The NRC implements this in part 50 through the requirements of § 50.120, “Training and qualification of nuclear power plant personnel.” The NRC is proposing under § 53.830 to adapt, with modifications, the requirements of § 50.120 for use in part 53 to provide more flexible personnel training and qualification requirements than those in § 50.120 and better reflect diverse concepts of operations. The NRC recognizes that the categories of nuclear power plant personnel in § 50.120 may not be needed for the diverse concepts of operations, staffing models, and non-traditional personnel roles and responsibilities anticipated under proposed part 53; conversely, and for the same reasons, additional categories of plant personnel may need to be covered by part 53. The NRC also recognizes that the timeframe prescribed in § 50.120 for the establishment of training programs may not be aligned with the schedules associated with the startup of certain types of commercial nuclear plant facilities. However, the NRC also recognizes that the SAT-based training required under § 50.120 remains an appropriate means by which training programs should continue to be developed and implemented. Therefore, the approach taken by the NRC in addressing the training of certain plant staff under the proposed part 53 reflects greater flexibilities in personnel categories and programmatic timeframes, while still retaining the requirement that such training programs be based on SAT. The NRC is proposing under § 53.830 to require SAT-based training programs with the timeframe for when such programs are required being based upon when the associated personnel are needed to support facility-specific needs. The training programs would cover the training and qualification of plant personnel in the general categories of supervisors, technicians, and other appropriate operating personnel. The licensee would not be required to seek NRC approval of a training program prior to usage. However, the licensee is required to accommodate NRC inspection of the training program. The NRC intends to develop guidance to facilitate the inspection of these training programs but does not intend for such guidance to preclude the potential for the training programs to be maintained by a separate, NRC-approved accreditation process. The proposed § 53.845 would require programs to be developed, implemented, and maintained to help ensure that design features and human actions have the capabilities and reliabilities necessary to demonstrate compliance with the safety criteria in subpart B throughout the operating life of each commercial nuclear plant. The proposed programmatic requirements in subpart F would also address areas such as radiation protection needed to control routine effluents during normal operations. The proposed §§ 53.850 through 53.910 would require programs to support specific activities needed to ensure the prevention or mitigation of unplanned events or to support normal operations for any reactor design. However, each holder of an OL or COL would be required to assess whether additional programs are needed for the specific reactor design and location of the commercial nuclear plant. Licensees would be able to combine, separate, and otherwise organize programs and related documents as appropriate for the technologies and organizations associated with the commercial nuclear plant. Proposed § 53.850 would require a radiation protection program associated with the requirements in subparts B and C for public doses resulting from normal operations and the protection of plant workers. The proposed requirements related to doses from normal operations, including routine effluents, would be similar to those specified in § 50.36a, “Technical specifications on effluents from nuclear power reactors,” and related requirements in standard TS for offsite dose calculation manuals. While the proposed section would include requirements that are technically and programmatically similar to part 50, proposed § 53.850 would not include a requirement for effluent-related TS as is required in § 50.36a. A proposed requirement similar to that found in the administrative controls section of TS for operating reactors licensed under parts 50 and 52 would be included for programmatic controls of solid wastes to complement the design requirements in proposed § 53.425. Proposed § 53.855 would require an emergency response plan that demonstrates compliance with the requirements in appendix E to part 50 and § 50.47(b) or § 50.160. The regulations in § 50.47 stating that the NRC will not issue certain licenses unless it finds that there is reasonable assurance that adequate protective measures can and will be taken to protect public health and safety in the event of a radiological emergency apply equally to applications under part 53 complying with the applicable standards set forth in either § 50.160 or the requirements in appendix E to part 50 and § 50.47(b). In its 2008 Advanced Reactor Policy Statement, the Commission stated their expectation that “the safety features of advanced reactor designs will be complemented by the operational program for Emergency Planning (EP). This EP operational program, in turn, must be demonstrated by inspections, tests, analyses, and acceptance criteria to ensure effective implementation of established measures.” Consistent with this policy statement, emergency plans and emergency planning zones are not safety features in the design. In SECY-97-020, “Results of Evaluation of Emergency Planning for Evolutionary and Advanced Reactors,” dated January 27, 1997, the staff indicated that the rationale upon which EP for current reactor designs is based, that is, potential consequences from a spectrum of accidents, is appropriate for use as the basis for EP for evolutionary and passive advanced LWR designs and is consistent with the Commission's defense-in-depth safety philosophy. Also, in its Safety Goals Policy Statement the Commission stated that: “A defense-in-depth approach has been mandated in order to prevent accidents from happening and to mitigate their consequences. Siting in less populated areas is emphasized. Furthermore, emergency response capabilities are mandated to provide additional defense-in-depth protection to the surrounding population.” Consistent with this policy statement, proposed § 53.855 contributes an additional independent layer of defense in depth for commercial nuclear plants. Therefore, the emergency plans and emergency planning zones under proposed § 53.855 are not used to demonstrate compliance with subpart B and subpart C of this part. Rather, compliance with the requirements in proposed § 53.855 would provide reasonable assurance that adequate protective measures can and will be taken to protect public health and safety in the event of a radiological emergency. Proposed § 53.860 would identify the applicable regulations for part 53 applicants related to the programs for physical security, cybersecurity, FFD, AA, and information security. These programs are discussed in more detail in section V, “Changes to Other Parts of 10 CFR,” of this document. Proposed § 53.860(a) would establish the physical protection program and present a graded approach to physical protection requirements. If a licensee can meet the proposed criterion in 
                        ( print page 86942)
                        § 53.860(a)(2)(i), then the requirement to protect against the design-basis threat (DBT) of radiological sabotage would not be applicable. The criterion in § 53.860(a)(2)(i) would require a licensee to show that potential consequences resulting from a DBT initiated event would result in offsite doses below the values in § 53.210 even if licensee mitigation and recovery actions, including any operator action, are unavailable or ineffective. Where the criterion is met, the resulting physical protection requirements would be those for protection of SNM and Category 1 and Category 2 radioactive material, if applicable. This proposal would apply a new regulatory approach for certain commercial nuclear plants in which the DBT of radiological sabotage would not be applicable. For those licensees able to meet the criterion in § 53.860(a)(2), the NRC would not conduct Force-On-Force (FOF) exercise inspections. Section 170D.a of the Act permits the Commission to determine which licensed facilities are part of a class of licensed facilities where NRC-conducted FOF exercises are appropriate to assess the ability of a private security force of a licensed facility to defend against any applicable DBT. For the class of licensees that meet the criterion of § 53.860(a)(2), it would not be appropriate to conduct FOF exercises to evaluate performance at commercial nuclear plants where the DBT of radiological sabotage is not applicable and the facility poses a lower risk to public health and safety from potential radiation exposure. These facilities would still have tailored security requirements and oversight consistent with their relatively low risk. For those licensees not able to meet the criterion in § 53.860(a)(2), proposed § 53.860(a) would permit the licensee to choose one of two paths to provide physical protection: (1) the current set of requirements in § 73.55, which would include any changes resulting from the ongoing proposed rulemaking on Alternative Physical Security Requirements for Advanced Reactors [2]
                        
                         that provides pre-determined physical security alternatives; or (2) the performance-based requirements in proposed § 73.100. In either case, the licensee would be subject to NRC-conducted FOF inspections. Proposed § 53.860(b) would require licensees to establish, implement, and maintain an FFD program under part 26. Section 53.860(c) would require licensees to establish, implement, and maintain an AA program in accordance with either § 73.56 or proposed § 73.120, as appropriate. Section 53.860(d) would require licensees to establish, implement, and maintain a cybersecurity program in accordance with either § 73.54 or proposed § 73.110. Section 53.860(e) would require licensees to establish, implement, and maintain an information protection system that complies with the requirements of §§ 73.21, 73.22, and 73.23, as applicable. Proposed § 53.865 would establish requirements for quality assurance and refer to appendix B of part 50 for the part 53 requirements for SR design features. Proposed requirements related to evaluating and reporting changes to the quality assurance program would be included in proposed subpart I and would be equivalent to those found in § 50.54. The proposed § 53.870 would require licensees to actively assess possible degradation of SSCs from the effects of aging, fatigue, and environmental conditions. The proposed inclusion of requirements related to designing and monitoring for possible degradation mechanisms reflects important lessons learned from the history of LWRs and the likely introduction of new design features and materials in future commercial nuclear plants. The allowable combinations of design features, operating experience, testing, and monitoring during operations would support performance-based approaches to the initial licensing of new technologies. The proposed performance-based approach to integrity assessment programs would also allow for the subsequent consideration of operating experience and appropriate corrective actions or allowable relaxations for ensuring that design features comply with the proposed functional design criteria of §§ 53.410 and 53.420. The proposed program would be based upon a comprehensive and integrated evaluation of the aging and other degradation mechanisms applicable to the design; identification of the affected SSCs; the allowances provided in the design of the SSCs for degradation; and schedules and procedures for determining if and at what rate degradation is occurring, as well as its cause. Risk insights could be used to prioritize the monitoring, evaluation, and management of degradation based upon the importance of the SSC to safety and the time frame for when the effects of degradation could be of concern. Proposed § 53.875 would establish requirements for a fire protection program supporting operations similar to § 50.48. The proposed fire protection program during operations would work in concert with specific fire protection requirements proposed in subpart C for design and analyses and in proposed subpart E for construction and manufacturing. Proposed § 53.880 would establish requirements for an inservice inspection (ISI) and inservice testing (IST) program, which are historically important activities conducted in accordance with ASME codes and regulations in § 50.55a. While the proposed part 53 would not incorporate specific consensus codes and standards into the regulations, § 53.880 allows for the use of generally accepted codes and standards. The proposed requirement for an ISI and IST program would reinforce the need to develop monitoring programs to be conducted during a plant's operations phase to complement the design process and address inherent uncertainties. The NRC encourages the continued use of consensus codes and standards supporting design, testing, and inspections to support integrated and performance-based approaches in demonstrating compliance with the proposed requirements in part 53. Proposed § 53.910 would establish requirements for developing, implementing, and maintaining procedures (
                        *e.g.,*
                         operations and emergency operating procedures) and guidelines (
                        *e.g.,*
                         accident management guidelines). The programmatic requirements for many of the procedures listed in this proposed section would be similar to the requirements found in the administrative controls section of TS for plants licensed under parts 50 and 52. The proposed inclusion, where appropriate, of accident management guidelines in these requirements is intended to ensure that an integrated set of procedures and guidelines would be established by licensees to ensure command and control across the spectrum of possible event sequences. The proposed required procedures would also include those needed to complement the design requirements in proposed § 53.440(m) related to criticality alarms and the equivalent of the procedures required in § 50.54(hh) to address notifications of potential aircraft threats. Subpart G—Decommissioning Requirements The proposed subpart G would provide the regulatory requirements for the decommissioning phase of the life cycle of a commercial nuclear plant. 
                        ( print page 86943)
                        The requirements being proposed in subpart G for the decommissioning of a commercial nuclear plant are adapted from the current regulations in § 50.75, “Reporting and recordkeeping for decommissioning planning,” § 50.82, “Termination of license,” and § 50.83, “Release of part of a power reactor facility or site for unrestricted use.” Although the requirements from those sections of part 50 have been copied into proposed subpart G with relatively few changes, the requirements are reorganized to fit within the part 53 structure. The few changes made were primarily to make the proposed requirements more technology inclusive by adding alternatives within sections, whereas some requirements in part 50 were developed specifically for LWRs. As an example, § 50.75 provides minimum amounts of decommissioning funds required to demonstrate reasonable assurance of funds for decommissioning LWRs. Such generic amounts have not been developed for all reactor technologies that may be licensed under part 53. Therefore, the Commission proposes in § 53.1020, “Cost estimates for decommissioning,” that site-specific cost estimates for decommissioning must be developed considering costs in such areas as engineering, labor, and waste disposal. The derivation of the generic cost estimates for LWRs in § 50.75 is provided in NUREG/CR-5884, “Revised Analyses of Decommissioning for the Reference Pressurized Water Reactor Power Station,” and NUREG/CR-6187, “Revised Analyses of Decommissioning for the Reference Boiling Water Reactor Power Station.” Similar to part 50, a provision for an annual adjustment of decommissioning cost estimates would be included in proposed § 53.1030. The NRC is currently pursuing another rulemaking, “Regulatory Improvements for Production and Utilization Facilities Transitioning to Decommissioning,” which was published as a proposed rule for public comment on March 3, 2022 (87 FR 12254). As these rulemakings progress, the NRC will consider revisions to part 53 to align the two rulemaking efforts. For example, the proposed § 53.1075 could be expanded to include or reference requirements for decommissioning in areas such as EP and security in addition to the proposed decommissioning fire protection plans that would provide an equivalent to § 50.48(f). Subpart H—Licenses, Certifications, and Approvals Proposed subpart H would provide requirements related to applications under part 53 for NRC licenses, certifications, or approvals for commercial nuclear plants. Proposed subpart H would specify requirements applicable to all part 53 applications as well as requirements specific to part 53 applications for LWAs, ESPs, standard design approvals, standard DCs, MLs, CPs, OLs, and COLs. Proposed subpart H would be equivalent to and include all existing licensing, certification, and approval processes currently covered under parts 50 and 52, with the exception of the process for early review of site suitability issues. Interactions with external stakeholders during the development of the proposed rule did not identify significant interest in or need for including the process for early review of site suitability issues in part 53. Much of the proposed subpart H regulatory text is identical to the corresponding language in parts 50 and 52, with minor changes to account for cross references in part 53, to make language technology neutral, or to reflect the unique analytical approach in part 53. In these instances, this preamble discussion will describe the language as “equivalent” to the existing corresponding requirement in part 50 or part 52 and will describe any deviations, where applicable. Because part 53 carries over the majority of the licensing options from parts 50 and 52, there are several sections in proposed subpart H that are similar to existing regulations in parts 50 and 52. Proposed § 53.1100 would address filing of applications for licenses, certifications, or approvals under oath or affirmation and is equivalent to § 50.30. The proposed § 53.1100 does not include the current requirement in § 50.30(a)(2) that the applicant maintain the capability to generate additional copies, because it is unnecessary in the age of electronic submissions. In addition, the existing requirement on applications for OLs in § 50.30(d) is included in proposed § 53.1124(g)(2), “Relationship between sections,” covering OLs, rather than in proposed § 53.1100. Proposed § 53.1101 would lay out activities requiring an NRC license and is equivalent to § 50.10(b). Proposed § 53.1103 would address combining applications and is equivalent to §§ 50.31, 50.52, and 52.8. Proposed § 53.1103(b) would continue the Commission's practice of combining multiple authorizations for a facility under parts 30, 40, 50, 52, and 70 into one license based on the Commission's authority under Section 161h. of the Act to combine NRC licenses. Proposed § 53.1106 would address elimination of repetition and is equivalent to § 50.32. Proposed § 53.1109 would provide general information requirements for the content of applications submitted to the NRC under part 53 and is equivalent to § 50.33, with the exception of § 50.33(f) on financial qualifications, which is covered in proposed subpart J, and § 50.33(h) on earliest and latest dates for completion of construction, which is covered in § 53.1306 of this subpart. Each application would need to include information to address the items in proposed § 53.1109 as cited in the appropriate section of this subpart for the application type. One change from current requirements can be found in proposed § 53.1109(i), which is not limited to electricity generation as it is currently in part 50. Some prospective NRC applicants are considering development of nuclear plants for other commercial ventures, such as process heat generation or hydrogen production. In addition, § 53.1109(j), which requires applications containing classified information to separate that information from the unclassified information in the application, refers to “Restricted Data or classified National Security Information” instead of the term used in the corresponding provision in § 50.33(j), “Restricted Data or other defense information.” This change was made to use the defined term in part 95 rather than “defense information” as used in § 50.33(j). The usage in § 50.33(j) dates back to the Atomic Energy Commission amendment of that section on January 19, 1956 (21 FR 355, 357) and was not changed with the issuance of part 95 (45 FR 14476; March 5, 1980) after the establishment of the NRC and the 1975 reissuance of the former Atomic Energy Commission regulations. The revised terminology also aligns with its usage in § 53.1115. Proposed § 53.1112 would address environmental conditions and is equivalent to § 50.36b. Proposed § 53.1115 would address requirements for agreements limiting access to classified information and is equivalent to § 50.37. Proposed § 53.1118 would address ineligibility of certain applicants and is equivalent to § 50.38. Proposed § 53.1120 would address exceptions and exemptions from licensing requirements for Department of Defense and DOE facilities and is equivalent to § 50.11. Proposed § 53.1121 would address public inspection of applications and is equivalent to § 50.39. Proposed § 53.1124 would address the relationship between the various licenses, certifications, and approvals 
                        ( print page 86944)
                        provided in this subpart, and the requirements are equivalent to a number of similar provisions in parts 50 and 52 including §§ 50.10, 52.13, 52.43, 52.73, 52.133, and 52.153. New provisions are provided in § 53.1124(c) and (d), that would allow an application for either a standard design approval or a standard DC under part 53 to reference applicable licensing-basis information that supported issuance of an OL or COL under part 53. These provisions are being proposed to offer additional flexibility beyond what is currently allowed under parts 50 or 52 for an applicant who may wish to license a first-of-a-kind reactor for operation prior to seeking generic approval or certification of the standard design. Proposed § 53.1124(e) would address the limitations that a manufactured reactor may only be transported to a site with a COL and is equivalent to § 52.153. Proposed § 53.1130 would address LWAs and is equivalent to § 50.10. Proposed §§ 53.1140 through 53.1188 would govern the content of ESP applications. Proposed § 53.1140 is equivalent to § 52.12. Proposed § 53.1143 would address filing of applications and is equivalent to § 52.15. Proposed § 53.1144 would address general information requirements for the content of applications and is equivalent to § 52.16. Proposed § 53.1146 would specify requirements for the technical contents of applications and is equivalent to § 52.17. Proposed § 53.1146(b)(2) provides applicants for ESPs a regulatory option to propose major features of the emergency plans or complete and integrated emergency plans in accordance with either the requirements in § 50.160 of this chapter, or the requirements in appendix E to part 50 of this chapter and § 50.47(b) of this chapter, as applicable. Proposed § 53.1149 would address standards for review of ESP applications and administrative review of applications, including hearings, and is equivalent to §§ 52.18 and 52.21. Proposed § 53.1155 would address referral to the ACRS and is equivalent to § 52.23. Proposed § 53.1158 would address issuance of ESPs and is equivalent to § 52.24. Proposed § 53.1161 would address the extent of activities permitted and is equivalent to § 52.25. Proposed § 53.1164 would address the duration of an ESP and is equivalent to § 52.26. Proposed § 53.1167 would address provisions for requesting a LWA after issuance of an ESP and is equivalent to § 52.27. Proposed § 53.1170 would address transfers of ESPs and is equivalent to § 52.28. Proposed § 53.1173 would address applications for ESP renewals and is equivalent to § 52.29. Proposed § 53.1176 would address criteria for renewal of an ESP and is equivalent to § 52.31. Proposed § 53.1179 would address the duration of an ESP renewal and is equivalent to § 52.33. Proposed § 53.1182 would address the use of a site for purposes other than those described in the permit and is equivalent to § 52.35. Proposed § 53.1188 would address finality of ESP determinations and is equivalent to § 52.39. Proposed §§ 53.1200 through 53.1221 would govern the contents of standard design approval applications. Proposed § 53.1200 is equivalent to § 52.131. Proposed § 53.1203 would address filing of applications and is equivalent to § 52.135. Proposed § 53.1206 would address general information requirements for the content of applications and is equivalent to § 52.136. Proposed § 53.1209 would address requirements for the technical content of applications and is largely equivalent to § 52.137. In proposed § 53.1209(a), the NRC proposes text that expands the discussion of “major portion” standard design approvals. Additional discussion regarding standard design approvals for a major portion of a standard design can be found in the NRC's “A Regulatory Review Roadmap for Non-Light Water Reactors,” which considers the Nuclear Innovation Alliance report “Clarifying `Major Portions' of a Reactor Design in Support of a Standard Design Approval.” Proposed § 53.1209(b) outlines the required content of the Final Safety Analysis Report (FSAR). Proposed requirements in § 53.1209(b)(2) for portions of the application addressing design information state that the application must include design information equivalent to that required for a standard DC. This reference to the pertinent DC requirements (specifically, those in proposed § 53.1239(a)(2) through (27)) is an efficiency that would prevent the need to repeat many of the same requirements for the content of a standard design approval application. Proposed § 53.1210 would address requirements for the content of a standard design approval application other than the FSAR. Proposed § 53.1210(a) would require the inclusion of a description of availability controls that are not included in the FSAR. Proposed § 53.1212 would address standards for review of applications and is equivalent to § 52.139. Proposed § 53.1215 would address referral to the ACRS and is equivalent to § 52.141. Proposed § 53.1218 would address staff approval of designs and duration of design approvals and is equivalent to §§ 52.143 and 52.147. Proposed § 53.1221 would address finality of standard design approvals and information requests and is equivalent to § 52.145 with the exception that it extends such finality to a standard approval referenced in a DC application. Standard design approvals issued to date under part 52 have been issued during the NRC's review of the standard DC application and have relied on the same application content. However, a future scenario could arise where the DC application is not submitted until after a design approval has been granted. The NRC would apply the same finality provisions in this situation as in the situation where a standard design approval is referenced in a COL application. There is no equivalent to proposed § 53.1221(d) in part 52 for standard design approvals. This provision would state that the Commission will require, before granting a CP, COL, OL, or ML which references a standard design approval, that engineering documents be completed and available for audit. A similar provision is included in part 52 in relation to a standard DC; and the NRC would require that design and analysis information needed for the Commission to make its safety determination be complete and available for any application the NRC is reviewing. Making this explicit provides increased clarity to future standard design approval applicants under part 53. Proposed §§ 53.1230 through 53.1263 would address standard DCs. Proposed § 53.1230 would address general provisions for standard DCs and is equivalent to § 52.41. Proposed § 53.1233 would address filing of applications and is equivalent to § 52.45. Proposed § 53.1236 would address general information requirements for the content of applications and is equivalent to § 52.46. Proposed § 53.1239 would address requirements for the technical content of applications and is equivalent to § 52.47(a). The requirements in proposed § 53.1239 have been modified from the analogous requirements in § 52.47(a) to align with the technical requirements in proposed part 53. Proposed § 53.1241 would address requirements for the content of a standard DC application other than the FSAR and is equivalent to § 52.47(b) and (d).
                        ( print page 86945) Proposed § 53.1242 would address review of applications and is equivalent to §§ 52.48 and 52.51. Proposed § 53.1242(c) would include a provision that would allow a DC applicant to reference applicable licensing-basis information for an OL or COL issued under part 53. As explained previously, this provision is being proposed to explicitly allow flexibility for an applicant who may wish to license a first-of-a-kind reactor for operation prior to seeking certification of the generic reactor design. For NRC findings on a reactor design in an OL or COL proceeding, this proposal would provide finality in a subsequent DC application that references information on the OL or COL proceeding's docket. This finality accorded to the OL or COL findings would bind the NRC staff and the ACRS but would not bind members of the public or the Commission. (To the extent an Atomic Safety and Licensing Board (ASLB) might have a role in a DC rulemaking, the OL or COL findings would not bind the ASLB either.) Specifically, members of the public would have the opportunity to comment on a proposed DC rule under well-established NRC practice. The rationale for binding the NRC staff and ACRS is similar to the rationale for a COL applicant referencing a standard design approval under part 52. Proposed § 53.1245 would address referral to the ACRS and is equivalent to § 52.53. Proposed § 53.1248 would address issuance of standard DCs and is equivalent to § 52.54. Proposed § 53.1251 would address duration of certifications and is equivalent to § 52.55(c). Proposed § 53.1254 would address application for renewal and is equivalent to § 52.57. Proposed § 53.1257 would address criteria for renewal and is equivalent to § 52.59. Proposed § 53.1260 would address duration of renewals and is equivalent to § 52.61. Proposed § 53.1263 would address finality of standard DCs and is equivalent to § 52.63. Proposed §§ 53.1270 through 53.1291 would address MLs covering manufacturing activities at one or more licensee facilities. Proposed § 53.1270 would address the scope of these sections and is equivalent to § 52.151. Proposed § 53.1273 would address filing of applications for an ML and is equivalent to § 52.155(a). Proposed § 53.1276 would address general information requirements for the content of ML applications and is equivalent to § 52.156, with one exception. Proposed § 53.1276 would require each application for an ML to also include the information required by § 53.1109(e). This information includes the type of license applied for, the use to which the facility will be put, the period of time for which the license is sought, and a list of other licenses, except operator's licenses, issued or applied for in connection with the proposed facility to address the potential variations in how MLs might be formulated under the proposed part 53. Proposed § 53.1279 would address requirements for the technical content of applications for MLs to be included in the FSAR and is equivalent to § 52.157. In addition, the requirements in proposed § 53.1279(a) and (b) have been modified from the analogous requirements in § 52.157 to align with the technical requirements in proposed part 53. Proposed § 53.1279(a)(2) outlines the required content of the application addressing design information and states that the application must include design information equivalent to that required for a standard DC. This reference to the pertinent DC requirements is an efficiency that would prevent the need to repeat the same requirements for the content of an ML application. Proposed § 53.1279(c) would provide application requirements related to the deployment of the completed manufactured reactor. Proposed § 53.1279(c)(1) would require inclusion of information related to the procedures governing the preparation of the manufactured reactor for shipping to the site where it is to be operated, the conduct of shipping, and the verification of the condition of the shipped items upon receipt at the site. Proposed § 53.1279(c)(2) would require that the application include information on the interaction of the design, manufacture, and installation of a manufactured reactor within the applicant's organization and the manner by which the applicant will ensure close integration between the designer, contractors, and any licensee of a facility in which the manufactured reactor is to be installed. Finally, proposed § 53.1279(c)(3) would require that the application include a description of the measures used for the control of interfaces between the holder of the ML and the holder of the COL for the commercial nuclear plant at which the manufactured reactor is to be installed. This information is necessary for the NRC to determine whether the applicant would have appropriate controls in place to ensure coordination between parties involved in the design, manufacture, and eventual operation of any reactor manufactured under an ML. Proposed § 53.1279(d) would include additional requirements for application content for applicants seeking an ML for manufactured reactors that will be fueled at the factory under a 10 CFR part 70 license, consistent with the requirements in § 53.620(d). These provisions would require the application to include information related to loading fuel and the required independent physical mechanisms to prevent criticality and to otherwise provide assurance that the fueled manufactured reactor can be successfully transported, installed, and operated at a site for which the Commission has issued a COL that authorizes construction and operation of a commercial nuclear plant using the manufactured reactor. Proposed § 53.1282 would provide requirements for other application content for MLs and is equivalent to § 52.158. Proposed § 53.1282(a)(1) would provide requirements to include in the ML application the ITAAC within the scope of the ML that the COL holder referencing the ML must satisfy. Proposed § 53.1282(a)(2) would require that the ITAAC from a referenced standard design apply to the portions of the ML design within the scope of the referenced standard design. Proposed § 53.1282(a)(3) would state that the COL application may include a notification that required referenced standard DC ITAAC have been satisfied at the manufacturing facility. Proposed § 53.1282(b) would require an ML application to include an environmental report and, consistent with existing requirements, proposed § 53.1282(b)(2) would note that if the ML application references a standard DC, the environmental report need not contain a discussion of severe accident mitigation design alternatives for the manufactured reactor as used in a commercial nuclear plant. Proposed § 53.1285 would provide standards for review of applications and administrative review of applications for MLs, including hearings, and is equivalent to §§ 52.159 and 52.163. Proposed § 53.1286 would address referral of applications to the ACRS and is equivalent to § 52.165. Proposed § 53.1287 would address issuance of an ML and is equivalent to § 52.167. Proposed § 53.1288 would address finality of MLs and is equivalent to § 52.171. Proposed § 53.1291 would address the duration of MLs and is equivalent to § 52.173. Proposed § 53.1293 would address the transfer of MLs and is equivalent to § 52.175. Proposed § 53.1295 would address the renewal of MLs and is equivalent to §§ 52.177, 52.179 and 52.181, with a minor exception. Proposed § 53.1295(a)(3) would state that an ML 
                        ( print page 86946)
                        for which a timely application for renewal has been filed remains in effect until the Commission has made a final determination on the renewal application, provided, however, that the holder of an ML may not begin manufacture of a manufactured reactor less than six months before the expiration of the license. The proposed 6-month time frame for this provision is changed from the 3-year period in the equivalent provision in part 52 because future reactor applicants may present smaller, simpler designs, to include micro-reactor designs, in ML applications than those that were envisioned when the existing requirements were written. A 6-month time frame for this provision would provide greater flexibility for ML holders related to manufactured reactors being produced when the ML expires. Proposed §§ 53.1300 through 53.1348 would address licensing requirements for CPs. Proposed § 53.1300 would set out general requirements for CPs and is equivalent to § 50.23. Proposed § 53.1306 would address the general information requirements for the content of applications for CPs and is equivalent to § 50.33(f) and (h). Proposed § 53.1309 would address requirements for the technical content of applications for CPs and includes the requirement to submit a Preliminary Safety Analysis Report (PSAR) that describes the facility and presents a preliminary safety analysis of the facility as a whole. This is in contrast to an OL application which is required to include an FSAR that describes the facility and presents a final safety analysis of the facility as a whole. Proposed § 53.1309 is equivalent to § 52.17(a)(1)(iv) through (a)(1)(x) and 52.17(b), with two exceptions. First, proposed § 53.1309 would replace the analysis of the dose criteria required by § 52.17(a)(1)(ix) with analysis to demonstrate compliance with the safety criteria defined in §§ 53.210 and 53.220. Second, proposed § 53.1309(a)(2) would add a requirement for a CP application to include several categories of detailed design information, although § 53.1309(a)(2)(ii) would allow certain relaxations of this requirement in view of aspects of a design that may not yet be fully developed. Section 53.1309 would reference the requirements for the content of an ESP application to address application requirements related to siting and would reference the requirements for the content of a DC application to address application requirements related to design of the commercial nuclear plant. Proposed § 53.1309(a)(2)(ii) would address the treatment of preliminary design information and notes that information provided in the application may include some aspects of the design that are not fully developed. This provision would require that the completed design, including any changes during construction, be described in the FSAR in an application for an OL. This would include the requirement for a description of the PRA required by § 53.450(a) and its results. Probabilistic risk assessments developed for commercial nuclear plants prior to construction would be based on the design and other information available at the time of the CP application. PRAs performed in early design stages or prior to construction may be inherently less detailed and may include projected information that will be subsequently verified or revised when the plant is built. Proposed § 53.1309(a)(4) would address preliminary description of the plans for coping with emergencies. Proposed § 53.1312 would address other application content for CPs. Proposed § 53.1312(a)(1) is equivalent to § 52.80(b) but is adapted for a CP application. Proposed § 53.1312(a)(2) is equivalent to § 52.80(c) but is adapted for a CP application. Proposed § 53.1312(b)(1) is equivalent to § 52.79(b), (c), and (d) but is adapted for a CP application. Section 53.1312(b)(2) is equivalent to portions of §§ 52.63(b)(1), 52.79(b)(1) through (b)(3), (c), and (d)(1) and (d)(3), 52.80, and 52.93(b), but is adapted for a CP application. Guidance for equivalent requirements in parts 50 and 52 is also addressed in RG 1.206, “Applications for Nuclear Power Plants,” Revision 1, section C.1.7. Proposed § 53.1315 would address standards for review of applications and administrative review of applications, including hearings, and is equivalent to §§ 52.81 and 52.85, but is adapted for a CP application. Proposed § 53.1318 would address finality of NRC approvals, licenses, and certifications referenced in a CP application and is equivalent to § 52.83(a) but is adapted for a CP application. Proposed § 53.1324 would address referral to the ACRS and is equivalent to § 50.58(a) and to § 52.87 but is adapted for a CP application. Proposed § 53.1327 would address authorization to conduct LWA activities and is equivalent to § 52.91 but is adapted for a CP application. Proposed § 53.1327(a) is equivalent to § 52.91(a) but is adapted for a CP application. Proposed § 53.1327(b) is equivalent to § 52.91(b) but is adapted for a CP application. Proposed § 53.1330 would address exemptions, departures, and variances for CP applicants. Proposed § 53.1333 would address issuance of CPs. Proposed § 53.1333(a) is equivalent to § 50.35(a). Proposed § 53.1333(b) is equivalent to § 50.35(b) and to § 52.97(c) but is adapted for a CP application. Proposed § 53.1336 would address the effect of CPs and is equivalent to § 50.35(b). Proposed § 53.1342 would address the duration of CPs. Proposed § 53.1342(a) is equivalent to § 50.55(a). Proposed § 53.1342(b) is equivalent to § 50.55(b). Proposed § 53.1345 would address the transfer, assignment, and disposal of CPs and is equivalent to § 50.80. Proposed § 53.1348 would address the termination of CPs and is equivalent to §§ 52.3(b)(8) and 52.110(a)(1) but is adapted for a CP application. Proposed §§ 53.1360 through 53.1405 address requirements for OLs. Proposed § 53.1366 would address requirements for the general content of applications for OLs. It would refer to general content requirements in proposed § 53.1109 and would require supplemental information. Proposed § 53.1366(a) is equivalent to § 50.33(f). Proposed § 53.1366(b) is equivalent to § 50.33(k). Proposed § 53.1369 would provide requirements for the technical content of applications for OLs to be included in the FSAR and is equivalent to § 50.34(b) but has been modified to align with the technical requirements in part 53. It would require that the FSAR include and, as needed, update information provided in the PSAR that was submitted and reviewed to support the associated CP application. Similar to the proposed requirements for the content of CP applications, proposed § 53.1369(a) would reference the requirements for the content of an ESP application to address application requirements related to the site. Section 53.1369(b) would reference the requirements for the content of a DC application to address some of the application requirements related to design of the commercial nuclear plant. Proposed § 53.1369(c) is equivalent to § 50.34(b)(7). Proposed § 53.1369(d) would require a description of the Integrity Assessment Program that would be required by proposed § 53.870. Proposed § 53.1369(e) is equivalent to § 50.34(e). Proposed § 53.1369(g) would provide requirements for OL application content to support proposed § 53.730 related to the role of personnel in the operation of the commercial nuclear plant and is adapted from requirements in part 55 and § 50.34(f). Likewise, proposed § 53.1369(h) would provide 
                        ( print page 86947)
                        requirements for OL application content related to training programs to support proposed §§ 53.730(g) and 53.830 and includes requirements equivalent to § 50.34(b)(8), § 52.79(a)(33), and part 55. Proposed § 53.1369(i) would provide requirements for OL application content related to emergency plans to support proposed § 53.855 and is equivalent to § 50.34(b)(6)(v). Proposed § 53.1369(j) would provide requirements for OL application content related to the applicant's organizational structure and is equivalent to § 50.34(b)(6)(i). Proposed § 53.1369(k) would provide requirements for OL application content related to the applicant's proposed maintenance program to support proposed § 53.715 and is equivalent to § 50.34(b)(6)(iv). Proposed § 53.1369(l) would provide requirements for OL application content related to the applicant's quality assurance program to support proposed § 53.865 and is equivalent to § 50.34(b)(6)(ii). Proposed § 53.1369(m) would provide requirements for OL application content related to the applicant's proposed radiation protection program to support proposed § 53.850 and is equivalent to § 50.34(b)(3). Proposed § 53.1369(n) through (p) would provide requirements for OL application content related to the applicant's proposed physical security program to support proposed § 53.860(a) and are equivalent to § 50.34(c) and (d). Proposed § 53.1369(q) would provide requirements for OL application content related to the applicant's proposed cybersecurity plan to support proposed § 53.860(d) and is equivalent to §§ 52.79(a)(36)(iv) and 73.54. Proposed § 53.1369(r) would provide requirements for OL application content related to the implementation of proposed security, safeguards, and cybersecurity plans to support proposed § 53.860 and is equivalent to § 52.79(a)(35)(ii) and 52.79(a)(36)(iv) and (v). Proposed § 53.1369(s) would provide requirements for OL application content related to the applicant's proposed fire protection program to support proposed § 53.875 and is equivalent to § 52.79(a)(40). Proposed § 53.1369(t) would provide requirements for OL application content related to the applicant's proposed ISI and IST program to support proposed § 53.880 and is equivalent to part of § 52.79(a)(11). Proposed § 53.1369(w) would provide requirements for OL application content related to the applicant's general employee training program to support proposed § 53.830 and is equivalent to § 52.79(a)(33). Proposed § 53.1369(x) would provide requirements for OL application content related to the applicant's FFD program to support part 26 and is equivalent to § 52.79(a)(44). Proposed § 53.1369(y) would provide requirements for OL applicant's programs to demonstrate that any safety questions identified at the CP stage have been resolved and is equivalent to § 50.34(b)(5). Proposed § 53.1369(z) would provide requirements for OL applicants to describe how the performance of each safety design feature has been demonstrated capable of fulfilling functional design criteria considering interdependent effects through either analysis, appropriate test programs, prototype testing, operating experience, or a combination thereof to support proposed § 53.440(a). It is largely equivalent to §§ 50.34(b)(5) and 50.43(e). Proposed § 53.1369(aa) would provide requirements for OL application content related to the applicant's proposed TS to support proposed § 53.710(a) and is equivalent to § 50.34(b)(6)(vi). Proposed § 53.1372 would address requirements for the content of OL applications other than the FSAR. Proposed § 53.1372(a) would require submission of an environmental report and is equivalent to § 50.30(f) and § 51.53(b). Proposed § 53.1372(b) does not have a direct parallel in parts 50 and 52 and would require the inclusion of a description of availability controls that are not included in the FSAR to support proposed § 53.710(b). Proposed § 53.1375 would address standards for review of OL applications and the administrative review of applications, including hearings, and is equivalent to §§ 52.81 and 52.85, except that the NRC has omitted 10 CFR part 54, “Requirements for Renewal of Operating Licenses for Nuclear Power Plants,” from the list of standards in the proposed § 53.1375(a). Proposed part 53 does not include detailed requirements related to renewal of licenses, although a general provision and possible placeholder for future requirements has been included as proposed § 53.1595. The NRC will decide after the part 53 final rule is published whether this future section will be retained in part 53 to address license renewal or whether the agency will take another approach to address license renewal for part 53 licensees, such as amending part 54 to address part 53 licensees. Proposed § 53.1381 would address referral to the ACRS and is equivalent to §§ 50.58 and 52.87. Proposed § 53.1384 would address exemptions, departures, and variances for OL applicants. Section 53.1384(a) is equivalent to § 52.93 but is adapted for OLs. Proposed § 53.1384(b) is equivalent to §§ 52.39(d) (with respect to ESPs) and 52.93 but is adapted for OLs. Proposed § 53.1387 would address issuance of OLs. The proposed introductory paragraph is equivalent to § 50.56. Proposed § 53.1387(a)(1)(i) is equivalent to §§ 50.50 and 50.57(a)(1). Proposed § 53.1387(a)(1)(ii) is equivalent to § 50.50. Proposed § 53.1387(a)(1)(iii) is equivalent to § 50.57(a)(2). Section 53.1387(a)(1)(iv) is equivalent to § 50.57(a)(3). Proposed § 53.1387(a)(1)(v) is equivalent to § 50.57(a)(4). Proposed § 53.1387(a)(1)(vi) is equivalent to § 50.57(a)(6). Proposed § 53.1387(a)(1)(vii) is equivalent to § 50.57(a)(5). Proposed § 53.1387(a)(1)(viii) is equivalent to § 52.97(a)(1)(vi) but is adapted for OLs. Proposed § 53.1387(c) is equivalent to § 50.57(b). Proposed § 53.1387(d) is equivalent to §§ 50.36(b) and 50.50. Proposed § 53.1390 would address backfitting of OLs and is equivalent to § 52.98(a) but adapted for an OL application. Proposed § 53.1396 would address duration of an OL and is equivalent to § 50.51(a) and § 52.104. Proposed § 53.1399 would address transfer, assignment, and other disposition of an OL and is equivalent to § 50.80. Proposed § 53.1402 would address applications for renewal of an OL and refers to proposed § 53.1595. Proposed § 53.1405 would address continuation of an OL and is equivalent to § 52.109 but is adapted to address an OL. Proposed §§ 53.1410 through 53.1461 would address requirements for COLs. Proposed § 53.1410 is equivalent to § 52.71. Proposed § 53.1413 would address general information requirements for the content of applications for COLs and is equivalent to § 52.77, which references § 50.33. Most of the provisions from § 50.33 are restated in proposed § 53.1109. Some requirements in § 50.33 related to financial qualifications and construction timelines are addressed in other sections of part 53. Proposed § 53.1416 would address the technical content to be included in an FSAR for an application for a COL and is equivalent to § 52.79 except as modified to reflect the technical requirements in part 53 and with one addition. Proposed § 53.1416 includes the statement that the Commission will require, before issuance of a COL, that engineering documents, such as analyses, drawings, procurement specifications, or construction and installation specifications, be completed and available for audit if the more 
                        ( print page 86948)
                        detailed information is necessary for the Commission to verify the information in the application and make its safety determination. This statement is equivalent to DC application requirements in § 52.47 and is included in proposed § 53.1416 for clarity. Similar to the proposed requirements for the content of OL applications, proposed § 53.1416(a)(1) would reference the requirements for the content of an ESP application to address application requirements related to siting. Section 53.1416(a)(2) would reference the requirements for the content of a DC application to address some of the application requirements related to design of the commercial nuclear plant. The remaining items under proposed § 53.1416(a) are likewise similar to the required content for OL applications under proposed § 53.1369(a). Proposed § 53.1416(b) would require COL applicants to provide a report documenting the resolution of any safety questions for SSCs for which research and development was necessary to confirm the adequacy of their design and is equivalent to § 50.34(b)(5). Proposed § 53.1416(c) would provide requirements for COL applicants to describe how the performance of each safety design feature has been demonstrated capable of fulfilling functional design criteria considering interdependent effects through either analysis, appropriate test programs, prototype testing, operating experience, or a combination thereof to support proposed § 53.440(a). It is largely equivalent to §§ 52.79(a)(24) and 50.43(e). Proposed § 53.1416(d) would address the content of COL applications referencing an ESP. Proposed § 53.1416(e) would address the content of COL applications referencing a standard design approval. Proposed § 53.1416(f) would address the content of COL applications referencing a standard DC. Proposed § 53.1416(g) would address the content of COL applications referencing an ML. Proposed § 53.1419 would address other application content for COLs and is equivalent to § 52.80. Proposed § 53.1419(a)(2) is new and would require the inclusion of a description of availability controls that are not required to be included in the FSAR. Proposed § 53.1422 would address standards for review of applications and the administrative review of applications, including hearings, and is equivalent to §§ 52.81 and 52.85. The NRC has removed part 54 from the list of standards in proposed § 53.1422(a). Proposed part 53 does not include requirements related to renewal of licenses, in relation to proposed §§ 53.1422 and 53.1595. Proposed § 53.1425 would address the finality of NRC approvals referenced in a COL application and is equivalent to § 52.83(a). Proposed § 53.1431 would address the referral of COL applications to the ACRS for review and is equivalent to § 52.87. Proposed § 53.1434 would address the authorization to conduct LWA activities and is equivalent to § 52.91. Proposed § 53.1437 would address exemptions, departures, and variances and is equivalent to § 52.93. Proposed § 53.1440 would address issuance of COLs and is equivalent to § 52.97. Proposed § 53.1443 would address finality of COLs and is equivalent to § 52.98. Proposed § 53.1449 would address inspection during construction and is equivalent to § 52.99. Proposed § 53.1452 would address operation under a COL and is equivalent to § 52.103. Paragraph (a) of proposed § 53.1452 would include footnotes to provide that, for licensees installing fueled manufactured reactors under a COL, (1) the COL holder would notify the NRC of its scheduled date for initiating the physical removal of any one of the independent physical mechanisms to prevent criticality required under § 53.620(d)(1) rather than its scheduled date for the initial loading of fuel, and (2) the NRC would time its publication of the notice of intended operation based on the COL holder's schedule for initiating the physical removal of any one of the independent physical mechanisms to prevent criticality required under § 53.620(d)(1) rather than the COL holder's scheduled date for the initial loading of fuel. These footnotes are consistent with the provisions of proposed § 53.620(d)(1)(iv), which would state that, upon initiating the physical removal of any one of the independent physical mechanisms to prevent criticality in the manufactured reactor's place of operation, the fueled manufactured reactor has commenced operation. For reactors without the independent physical mechanisms to preclude criticality under proposed § 53.620(d)(1), operation begins with initial fuel load. In both cases, removal of the physical features to prevent criticality (for reactors with such features) and initial fuel load (for reactors without such features) put a fully constructed utilization facility in a position to sustain a nuclear chain reaction, and in both cases, the utilization facility cannot sustain a nuclear chain reaction (for lack of sufficient reactivity) until the action takes place. Therefore, the NRC proposes that initiating the physical removal of any one of the independent physical mechanisms to prevent criticality is the best analogue to initial loading of fuel for reactors without such features. The proposed footnote in § 53.1452(a) regarding timing of the notice of intended operation for fueled manufactured reactors with independent physical mechanisms to prevent criticality also addresses the requirements of Section 189a.(1)(B)(i) of the Act. This section requires, in part, that “[n]ot less than 180 days before the date scheduled for initial loading of fuel into a plant by a licensee that has been issued a combined construction permit and operating license under section 185b., the Commission shall publish in the 
                        **Federal Register**
                         notice of intended operation.” That section further requires that this notice provide a 60-day period in which to request a hearing “on whether the facility as constructed complies, or on completion will comply, with the acceptance criteria of the license.” In the case where a fueled manufactured reactor arrives at the site where it is to be operated by a COL holder, the manufacturer would have loaded fuel at the factory under its part 70 license. Therefore, at the site of operation, there would not be “initial loading of fuel into a plant 
                        *by a licensee that has been issued a combined construction permit and operating license*
                        ” (emphasis added). Under a literal reading of the entry condition in Act Section 189a.(1)(B)(i), this situation would not trigger its requirements. However, the purpose of the provision is to offer the hearing opportunity at least 180 days prior to when the fuel is loaded and ready for use at its authorized location. It would be contrary to that purpose if, in this situation, the Commission did not publish the notice of intended operation and opportunity for the public to request a hearing on conformance with the acceptance criteria in the COL for the site of operation. To fulfill the underlying purpose of the law, the NRC proposes to time the notice of intended operation based on the COL holder's schedule for initiating the physical removal of any one of the independent physical mechanisms to prevent criticality required under § 53.620(d)(1). This action by the COL holder would be the best analogue to initial fuel load by the COL holder for the reasons stated previously. This analogue is adopted in other sections of the proposed part 53 and related sections in parts 50 and 73 that use initial fuel loading to identify 
                        ( print page 86949)
                        a transition point for the applicability of regulatory requirements. To address the possible loading of fuel into a manufactured reactor for subsequent transport to and use at a commercial nuclear plant, multiple sections that determine the applicability of regulations have been drafted or revised to allow for either initial fuel load or initiating the physical removal of any one of the independent physical mechanisms to prevent criticality required under § 53.620(d)(1) for a fueled manufactured reactor to determine the applicability of the requirement, as appropriate. Proposed § 53.1455 would address duration of COL and is equivalent to § 52.104. Proposed § 53.1456 would address the transfer of a COL and is equivalent to § 52.105. Proposed § 53.1458 would address application for renewal and is equivalent to § 52.107. Proposed § 53.1461 would address continuation of COL and is equivalent to § 52.109. Proposed § 53.1470 would address standardization of commercial nuclear plant designs and licenses to construct and operate commercial power reactors of identical design at multiple sites and is equivalent to appendix N of part 52. This section would set out the particular requirements and provisions applicable to situations in which applications for CPs and subsequent OLs, or COLs, under this part are filed by one or more applicants for licenses to construct and operate nuclear power reactors of identical design (“common design”) to be located at multiple sites. Additional information related to this proposed section is provided in the final rule to revise part 52 (72 FR 49352; August 28, 2007). Subpart I—Maintaining and Revising Licensing-Basis Information Part 53 would establish requirements for the maintenance of licensing-basis information in subpart I. Section 53.1500 would describe the purpose of the subpart in terms of the definition of licensing-basis information in subpart A. Subpart I would be closely tied to the requirements in subpart H, which would provide the requirements for contents of applications for the various types of licenses issued under part 53. Subpart I would generally be organized into sections dealing with: (1) licensing-basis information that licensees are not authorized to change without NRC approval (
                        *e.g.,*
                         licenses, regulations); and (2) licensing-basis documents that licensees may change provided specified criteria are satisfied (
                        *e.g.,*
                         FSAR, program descriptions). The subpart would also capture certain general conditions on licenses and changes to the licenses related to the transfer and termination of licenses. Section 53.1502 would define specific terms and conditions of licenses. These terms and conditions would be equivalent to the regulations in: (1) § 50.54(h) stating that each license is subject to the provisions of the Act and requirements issued by the Commission; (2) § 50.54(s) stating the actions the Commission would take if it makes a finding that there is not reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency; (3) § 50.54(aa) stating that each license is subject to the specified sections of the Federal Water Pollution Control Act; and (4) § 50.54(dd) stating that a holder of an OL or COL may take reasonable actions that depart from the license in a national security emergency. Section 53.1505(a) would serve as an introduction to and overview of the sections that follow on changes to licensing-basis information requiring prior NRC approval, namely the elements of licensing-basis information defined by licenses, orders, and regulations. The related sections within these subparts would primarily deal with the process of how a licensee requests and the NRC issues an amendment to a license or issues an order that modifies a license. Another important element of licensing-basis information that a part 53 licensee would not be able to change or deviate from without NRC approval would be the NRC regulations themselves. Section 53.1505(b) would refer to § 53.080 in subpart A that would provide the criteria for a licensee or other party to satisfy when requesting an exemption from NRC regulations. Section 53.1510 would be equivalent to § 50.90 and would require that a licensee submit an application to request an amendment to a license. The required assessments that would be included within an application to amend a license under part 53 would need to address the safety criteria and analysis requirements of subparts B and C. As with parts 50 and 52, licensees would be required to include in their applications to amend a license an analysis of whether the amendment involves no significant hazards consideration using the standards in § 53.1520, which would be equivalent to the standards in § 50.92. Although this rulemaking provided an opportunity to revise the terminology related to no significant hazards consideration determinations, which dates to the early 1960s when applications were supported by final hazard summary reports, the NRC is proposing to maintain the same terminology used in part 50 to minimize the need for associated changes in other regulations, guidance, and public notices. Section 53.1515 would establish requirements for public notices and state consultations associated with the NRC's processing of a license amendment request. This section would be equivalent to § 50.91 for the NRC's processes related to applications to amend an OL or COL. Section 50.91(b) stipulates that the Commission will make available to the licensee the name of the appropriate State official designated to receive such amendments. While the Commission intends to continue following this practice, the Commission has not included this administrative matter in proposed part 53. Proposed § 53.1515(b)(3) contains some modifications compared to § 50.91(b)(3) for clarity; these revisions are not intended to revise the substance of the provisions in part 53 compared to part 50. Section 53.1520 would be based on § 50.92. The section would continue to use the criteria in § 50.92 for determining that a proposed amendment involves no significant hazards consideration. Although more specific terms such as event sequence are used throughout part 53, § 53.1520 would use the term “accident” to maintain consistency with the long history of making no significant hazards consideration determinations under part 50. Section 53.1525 would provide requirements for holders of an OL or COL requesting to revise information from a DC rule that was referenced in the initial license application and included in or incorporated by reference into the facility FSAR. In keeping with the current requirements in part 52, the portion of the part 53 facility licensing-basis information obtained from the certified design would be divided into two categories. The most significant design information and the ITAAC would be certified by rule and designated as “certification information.” The remaining information, which makes up the majority of the design information approved as part of the DC, would not be certified by rule and is not considered “certification information.” Part 52 refers to these categories of information as Tier 1 and Tier 2 information, respectively, and refers to a change made to that information on a plant-specific basis as a departure. Under part 52, a departure from Tier 1 information requires an exemption and, 
                        ( print page 86950)
                        for information incorporated into the license, a license amendment. Part 53 would dispense with the Tier 1 and Tier 2 terminology. Rather, § 53.1525 would use the term “certification information” in place of Tier 1, and a plant-specific departure from the certification information would require both a request for an exemption from the associated DC rule and, for information such as ITAAC incorporated into the license, a license amendment. However, as would be provided in § 53.1525(c), a plant-specific departure from the information approved by the NRC as part of the DC rule but which is not certification information (
                        *i.e.,*
                         Tier 2 information under part 52) would be assessed using the process and criteria defined in § 53.1550 for changes to a FSAR. An applicant or licensee would need to identify such a change as a departure from the referenced standard design in the updated FSAR. The process for making a generic change to a certified design would be described in the associated section in subpart H. Section 53.1530 would not allow the holder of an ML or the holder of a COL using a manufactured reactor to make changes to the design of the manufactured reactor without requesting a license amendment from the NRC. This section would provide the equivalent requirements as those in §§ 52.98 and 52.171. Section 53.1535 would establish requirements for license amendments during construction. The section would provide the equivalent options and requirements for the holders of a CP as those in § 50.35(b). The regulations would allow but do not require the holder of a CP or LWA to request an amendment under § 53.1510 if the licensee desires to obtain NRC approval of a specific design feature or specification. The requirements for obtaining an amendment to a COL to address changes during construction would also be provided in § 53.1535. The proposed process would differ from the current requirements in part 52 by adopting a requirement that would explicitly support a change process like that described in RG 1.237, “Guidance for Changes During Construction for New Nuclear Power Plants Being Constructed Under a Combined License Referencing a Certified Design Under 10 CFR part 52.” The proposed regulation would allow the holder of a COL to proceed at its own risk in making a change during the construction process and would require that licensee to submit a license amendment request no later than 45 days from the date the licensee begins to implement the change or departure requiring NRC approval. Section 53.1540 would serve as an introduction to the sections that follow on changes to licensing-basis information that are primarily under the control of a licensee but for which evaluations are made to determine if a submittal to the NRC requesting approval would be required. The section would also include definitions that would be applicable when using the processes in §§ 53.1545 through 53.1565. The definitions would be largely equivalent to those in § 50.59(a) but include some revision to reflect the structure and terminology in other subparts in part 53. For example, the definition of “
                        *Change*
                        ” in § 53.1540(b) would address a “design feature or related functional design criteria” rather than a “design function,” because the former are defined terms in part 53. Similarly, in § 53.1540(b), the phrase “design basis” from § 50.59(a)(2) would be replaced with functional design criteria for SR SSCs. Section 53.1545 would provide the proposed requirements for updating of FSARs. While the process-related requirements proposed under § 53.1545 would be largely the same as those in § 50.71, the specifics of information to be updated would differ due to the role of PRA in satisfying the requirements in subparts B and C. Additionally, the use of the risk-informed approach in subpart C would result in some but not all PRA information being in the FSAR or another licensing basis document and therefore a separate PRA update requirement similar to § 50.71(h) is not included in proposed subpart I. Proposed § 53.1239(a)(18) in subpart H and the related references to this proposed requirement for the holders of OLs and COLs would require a description of the PRA required by § 53.450(a) and its results to be included in FSARs. However, guidance documents are planned to clarify the division of PRA-related information that would need to be in the FSAR, in other possible licensing basis documents, and controlled as plant records subject to inspections and audits. At a minimum, the information from the PRA that would be needed to show compliance with subpart C would be included in the FSAR (
                        *e.g.,*
                         PRA summary and analytical results for LBEs). The submittal of voluminous PRA information was initially required under part 52, but that proved to be impractical and was revised in the 2007 revision of part 52. Guidance is being developed to ensure sufficient information is submitted to the NRC to support the licensing process and the NRC's regulatory findings under part 53 or similar applications using the LMP under parts 50 or 52. The NRC has posed a question in section VI, “Specific Requests for Comments,” of this document that asks about the appropriate level of detail for PRA-related information in an FSAR and whether other licensing basis documents might be more appropriate to both provide information to the NRC and ensure the PRA is maintained and updated as proposed in subpart C. The program document would provide more detail than the summaries in the FSAR but still be a much-condensed source of information in comparison to the documentation of the PRA. Section 53.1545(a)(3) and (4) would be based on the inclusion of at least a summary of PRA results and the related margins to safety criteria in the FSAR and would require updates to that information. The routine reporting of these margins would also inform application of the criteria for allowing changes without an amendment in the following section (§ 53.1550) in subpart I. Section 53.1550 would establish requirements for evaluating changes to a facility as described in its FSAR. This proposed section would provide the equivalent of the requirements in § 50.59 for evaluating changes to an FSAR (as updated) and determining if a license amendment is required to implement a change to a facility or procedures. The evaluation criteria proposed in § 53.1550 would reflect the role of the PRA in the safety analyses under part 53 and would include several measures related to the changes in plant risk resulting from a change in the plant design or plant procedures. Examples would include criteria that rely on the identification of risk-significant event sequences in accordance with the analysis requirements of § 53.450; exceeding the LBE evaluation criteria as defined in § 53.450; the consideration of potential reductions in margin between the estimated comprehensive risk metrics and associated risk performance objectives in the safety criteria in § 53.220; changes to the safety classification of SSCs; and consideration of reductions in defense in depth. Section 53.1550 would include a criterion related to a departure from a method of evaluation used in the safety analyses. The NRC has not yet developed draft guidance for use in applying proposed § 53.1550 but anticipates that the NRC and stakeholders will assess the potential need for such guidance and that such guidance would, if needed, be 
                        ( print page 86951)
                        developed as part of ongoing or future activities. Section 53.1550 would include certain concepts taken from existing guidance for § 50.59 in the proposed criteria related to DBAs. Specifically, criterion (iv) for changes made to a method of evaluation of DBAs under § 53.450(f) would be equivalent to a change in a method of evaluation under § 50.59, and criterion (viii) on assessing if a change creates a possibility for an accident of a different type than previously analyzed in the FSAR would be similar to the § 50.59 criterion (v). Guidance documents will be prepared to address the content of applications for PRA-related information under proposed part 53, and this guidance will also influence how potential changes in the evaluation of LBEs other than DBAs analyzed under § 53.450(e) are evaluated and reported under the proposed criterion (iv). Section 53.1550(a)(2)(x) would require evaluating plant changes to ensure they would not prevent satisfying the design requirements in § 53.440(j) related to the impact of a large commercial aircraft. The inclusion of a proposed requirement under § 53.1550 related to design features for protecting against aircraft impact would reflect the proposed design requirement in subpart C and related proposed requirements in subpart H to address the proposed design requirement in FSARs. Sections 53.1560 through 53.1565 in subpart I would define the processes for a licensee to evaluate changes to the program documents included in the licensing-basis information submitted to the NRC and to modify such programs without NRC prior approval. Section 53.1560 would include the proposed requirements for updating program documents included in licensing-basis information and would provide the equivalent of FSAR updates for key program documents. The proposed requirements in these sections would provide a uniform approach for updating program documents, which correspond to the programs required under subpart F. The proposed § 53.1565 would provide a process for licensees to make changes to program documents included in licensing-basis information without obtaining prior NRC approval. The proposed requirements would include several generic criteria that, if not satisfied, would prompt the need for NRC approval of a change to a program document. These generic criteria would include whether a change would comply with TS and NRC regulations. Another proposed criterion for evaluating changes to program documents would be conforming with program-specific requirements, including NRC-approved program documents with more specific criteria for a particular program, regulations, administrative controls sections of TS, and NRC-approved program documents. Proposed § 53.1565(d) would include specific criteria for evaluating changes to several program documents that have well established change processes and guidance for licensees under parts 50 and 52. The program documents specifically addressed in the proposed section would include quality assurance programs that would be equivalent to § 50.54(a), an emergency preparedness program that would be equivalent to § 50.54(q), and the security program that would be equivalent to § 50.54(p). The proposed § 53.1570 would establish requirements for the transfer of commercial nuclear plant licenses by providing the equivalent requirements of § 50.80 for the possible transfer of an ESP, CP, OL, or COL. Likewise, the proposed § 53.1575 would establish requirements for the termination of an OL or COL by providing the equivalent requirements of § 50.82. Other proposed requirements related to decommissioning and license termination would be included in subpart G. Section 53.1580 would establish requirements for information requests the NRC could send to the various types of licensees and would provide requirements that would be equivalent to requirements in § 50.54(f). The proposed § 53.1585 would provide the requirements that would be equivalent to requirements in § 50.100 to address revocation, suspension, modification of licenses, and approvals for cause. Section 53.1590 would propose to address backfitting requirements by providing requirements that would be equivalent to those in § 50.109. Proposed § 53.1595 would address license renewals under part 53 with simple statements that licenses may be renewed. This section would be expanded through future rulemakings to more fully describe or reference the processes related to requesting and processing applications to renew ESPs, OLs, and COLs issued under part 53 (if finalized). Subpart J—Reporting and Other Administrative Requirements Part 53 would address various reporting and administrative requirements in subpart J. Section 53.1600 would explain the organization of the various sections within the subpart related to providing unfettered access to NRC inspectors; maintaining certain records and reporting specified events or conditions; demonstrating compliance with financial qualification requirements and providing specified financial reports; and maintaining financial protections to address potential accidents. Section 53.1610 would establish requirements for the provision of facilities and unfettered access for inspections. These requirements would be equivalent to § 50.70 with only minor changes proposed to provide additional flexibilities and address possible differences related to reactors licensed under part 53 and the possibility that some commercial nuclear plants may not be assigned resident inspectors. Section 53.1620 would provide for maintenance of records and the making of various reports to the NRC. These requirements would be largely equivalent to § 50.71. This section is not intended to reflect all provisions in § 50.71; several important requirements in § 50.71 would be captured in other sections of part 53. For example, § 53.1545 within subpart I would provide requirements that would be equivalent to § 50.71(e), updating FSARs, and § 53.1680, “Annual financial reports,” would provide the equivalent of § 50.71(b), which covers financial reports. A reporting requirement related to completion of power ascension testing would be added to § 53.1620 to support the assessment of annual fees under 10 CFR part 171, “Annual Fees for Reactor Licenses and Materials Licenses, Including Holders of Certificates of Compliance, Registrations, and Quality Assurance Program Approvals and Government Agencies Licensed by the NRC,” which normally commence upon completion of those testing activities. Section 53.1630 would establish requirements for immediate notification requirements for operating commercial nuclear plants. These requirements would be equivalent to § 50.72 with minor changes proposed to make the reporting criteria technology inclusive. In addition, a new version of NRC Form 361 (NRC Form 361S) would be created for use by part 53 licensees, but without LWR-specific terminology to ensure technology inclusiveness. A separate rulemaking activity, “Reporting Requirements for Nonemergency Events at Nuclear Power Plants,” has been initiated to consider possible changes to the requirements in § 50.72. At a future date, the NRC may consider reconciling future changes to § 50.72 with the requirements proposed in part 53, which have been taken or derived from the current reporting requirements.
                        ( print page 86952) Section 53.1640 would address the licensee event report system. These requirements would be equivalent to § 50.73 with minor changes proposed to make the requirements inclusive of various reactor technologies and to reflect appropriate internal references to other sections in part 53. In addition, NRC Forms 366, 366A, and 366B would be revised to include corresponding check boxes for part 53 licensees. Section 53.1645 would require periodic reporting of the quantity of radionuclides released to unrestricted areas in liquid and gaseous effluents, doses to members of the public, and the results of environmental monitoring. These reporting requirements in the proposed part 53 would be largely equivalent to those in the TSs required by § 50.36a, “Technical specifications on effluents from nuclear power reactors.” The only difference would be that a § 50.36a requirement to specifically address conditions where the dose to the maximally exposed individual could be significantly above design objectives would refer to a design objective of 10 mrem/year total effective dose equivalent, instead of referring to the design objectives in appendix I to part 50. The proposed section would also include an equivalent to the reporting requirement in section IV of appendix I to part 50 if the radiation exposure to a member of the public in any calendar quarter exceeds one-half of the annual ALARA design objective. Section 53.1650 would include a reporting requirement to support safeguards agreements between the United States and the International Atomic Energy Agency (IAEA) and would be equivalent to § 50.78. Section 53.1660 through 53.1700 would address financial requirements and would be largely similar to existing regulations in parts 50 and 52. Section 53.1670 would be entitled “Financial qualifications” and would require applicants other than electric utilities to possess or have reasonable assurance of obtaining funds for the activities for which the license is being sought. The NRC is seeking feedback on these sections and their ramifications for merchant plants [3]
                        
                         in section VI, “Specific Requests for Comments,” of this document. The remaining financial reports in part 53 would be equivalent to § 50.71(b) for annual financial reports, § 50.76 for a change of status, § 50.54(cc) for the filing of a petition for bankruptcy, and § 50.81 for creditor regulations. Sections 53.1710 through 53.1730 would address financial protection requirements. Section 53.1720 would require insurance to stabilize and decontaminate a plant following an accident. These requirements would be taken from § 50.54(w) with the only notable change being the addition of a provision allowing plant-specific estimates of costs to stabilize and decontaminate a plant as an alternative to the $1.06 billion minimum coverage in § 50.54(w). Section 53.1730 is equivalent to § 50.57(a)(5) and would refer to the requirements in 10 CFR part 140, “Financial Protection Requirements and Indemnity,” related to financial protection requirements and indemnity agreements, including the financial protection requirements of the Price-Anderson Act. Subpart M—Enforcement Subpart M would contain two provisions, § 53.9000 and § 53.9010, which are analogous to provisions contained in other parts of 10 CFR Chapter I imposing requirements on regulated entities. Section 53.9000 would provide notice of the Commission's authority under the Act to obtain injunctions or other court orders for the enumerated violations. Paragraph (a) of § 53.9010 would provide notice to all persons and entities subject to part 53 that they are subject to criminal sanctions for willful violations, attempted violations, or conspiracy to violate certain regulations under part 53. Criminal sanctions would not apply to the regulations listed in paragraph (b). The regulations for which criminal penalties would apply are limited to those that establish either a regulatory obligation or prohibition. V. Changes to Other Parts of 10 CFR Chapter I 10 CFR Part 26 A. Introduction The NRC is proposing a technology-inclusive, risk-informed, and performance-based approach for the application of drug and alcohol testing and fatigue management requirements for facilities licensed under part 53. The proposed requirements applicable to these applicants, licensees, and other entities would be commensurate with the radiological consequences presented by the applicants' facilities and the operation of these facilities.[4]
                        
                         The proposed FFD framework would consist of a two-tiered graded approach similar to that currently in part 26 and an optional third tier for part 53 commercial nuclear plants that perform an analysis that demonstrates the facility and its operation would satisfy the criterion in proposed § 26.603(c), which refers to § 53.860(a). This proposed FFD framework would be established in subpart M, “Fitness for Duty Programs for Facilities Licensed Under Part 53,” of part 26. The NRC used operating experience to provide regulatory flexibility in the proposed subpart M of part 26 framework to help support a licensee's or other entity's response to changes in societal drug use, drug testing technologies and processes, and FFD program performance. The flexibility would also help in FFD program implementation because of the wide variety of staff sizes anticipated at commercial nuclear plants licensed under part 53 and the geographically remote locations in which commercial nuclear plants may be sited. The proposed first-tier FFD program requirements would apply to part 53 licensees and other entities of commercial nuclear plants under construction who satisfy the criterion in § 26.603(c) but elect not to implement proposed § 26.604, “FFD program requirements for facilities that satisfy the § 26.603(c) criterion,” or who do not satisfy the criterion in § 26.603(c), and to holders of MLs who are assembling or testing manufactured reactors. These requirements would be provided in proposed § 26.605(a) and would be essentially equivalent to those requirements in subpart K, “FFD Program for Construction,” of part 26 as supplemented by select requirements from subparts E, “Collecting Specimens for Testing,” and I, “Managing Fatigue,” of part 26, and the requirements in subparts A, “Administrative Provisions,” and O, “Inspection, Violations, and Penalties,” of part 26. The first-tier requirements would involve policies, procedures, behavioral observation, fatigue management, drug and alcohol testing, determinations of fitness, appeals, training, sanctions, auditing, change control, performance monitoring, recordkeeping, and reporting. These requirements would help deter individuals subject to this section from illicit drug and/or alcohol use and from being impaired from any cause including fatigue. These proposed requirements would also help licensees 
                        ( print page 86953)
                        and other entities identify individuals as users of impairing substances and demonstrate compliance with § 26.23, “Performance objectives.” The proposed second tier would include all the proposed first-tier requirements, plus the more comprehensive set of FFD program requirements in current subparts C, “Granting and Maintaining Authorization,” D, “Management Actions and Sanctions to be Imposed,” H, “Determining Fitness-for-Duty Policy Violations and Determining Fitness,” and N, “Recordkeeping and Reporting Requirements,” of part 26. These requirements would be provided in proposed § 26.605(b) and would be applicable to licensees and other entities satisfying the § 26.603(c) criterion, at their discretion. These requirements would also apply to licensees or other entities not satisfying the § 26.603(c) criterion that implement an FFD program under subpart M of part 26, before the loading of fuel onsite into a reactor vessel; before receiving a manufactured reactor; or before operating, testing, performing maintenance of, or directing the maintenance or surveillance of security-related equipment or equipment that a risk-informed evaluation process has shown to be significant to public health and safety. The second-tier requirements are based on the additional risk presented by nuclear reactor assembly, testing, fueling, and operation and the necessity for human actions in certain event sequences. The inclusion of the current part 26 requirements would align proposed part 53 FFD and AA program requirements with the current FFD and AA programs required for facilities licensed under parts 50 and 52. This approach would ensure effective and consistent AA and FFD program implementation across the commercial nuclear power industry, thereby ensuring uniform requirements for individuals who may perform roles and responsibilities for multiple facilities regardless of facility licensure. Proposed § 26.604 would offer an alternate option for an applicant implementing an FFD program under subpart M of part 26. If the applicant demonstrates that the criterion in proposed § 26.603(c) is met, then the applicant (and the subsequent licensee or other entity) must still implement an FFD program described in subpart M of part 26; however, drug and alcohol testing would not be required unless FFD performance declines or the applicant, licensee, or other entity elects to implement drug and alcohol testing. The proposed § 26.604 requirements are equivalent to those proposed in § 26.605(a) except for required drug and alcohol testing. This proposed framework would focus on the human performance of individuals while they are performing those duties and responsibilities that make them subject to the FFD program. This performance would be verified through behavioral observation, evaluation of any FFD concerns, performance monitoring, fatigue management, and determinations of fitness. Applicants that do not satisfy the criterion in proposed § 26.603(c), or elect not to perform the analysis required to demonstrate that the criterion in § 26.603(c) is met, would be subject to an FFD program described in § 26.605, “FFD program requirements for facilities that do not implement § 26.604,” or an FFD program that implements all part 26 requirements, except for those requirements in subparts K and M of part 26. In establishing the minimum FFD program requirements in § 26.604, the NRC reviewed current advanced reactor designs against that of a non-power production or utilization facility (NPUF) that is not required to implement an FFD program for those individuals who have unescorted access to the controlled access area (and vital area for some facilities), including NRC-licensed operators.[5]
                        
                         This review was performed because commercial nuclear plants licensed under part 53 could be designed with similar power levels and radiological consequences as the currently licensed NPUFs. From this review, three principal considerations supported the minimum set of requirements for the § 26.604 FFD program. First, the radiological consequences presented by a part 53 licensed facility and its operation that satisfy the criterion in § 26.603(c) may present a greater potential radiological consequence to workers and the public in the vicinity of the facility than does an NPUF. Second, the operating characteristics of a part 53 licensed facility are unlike that of an NPUF because there may be a higher reliance on individuals at the part 53 site to safely and competently operate, maintain, surveil, and secure SSCs that may not be required at an NPUF, such as systems that provide secondary heat transfer, reactor coolant flow, pressure control, and at-power core refueling. Differences in operating characteristics could include, for example: long-term, full power operation with automated reactivity control systems for load-following; active and passive safety and security systems; innovative non-light-water heat transfer systems; and energy storage and hazardous chemical systems. The individuals at part 53 facilities may also be required to communicate to individuals both onsite and offsite, such as electrical load dispatchers, any conditions adverse to safety, security, or quality. Third, part 53 licensed facilities may be sited in geographically remote locations that may not have a physically available administrative or corporate support team to provide face-to-face oversight, engineering expertise, and maintenance support like that at NPUFs. This places a higher reliance on those individuals required at a part 53 facility being fit for duty and trustworthy and reliable because a replacement individual may not be readily available. The NRC proposes to exclude drug and alcohol testing from the proposed § 26.604 framework for five reasons: (1) the § 26.23 performance objectives can be met through effective implementation of the defense-in-depth regulatory framework established by behavioral observation, reporting of legal actions, the proposed performance monitoring and review program (PMRP), FFD training, and requirements from the physical protection, AA, cyber protection, and licensed operator programs; (2) the PMRP would require the licensee or other entity to monitor its FFD program performance (both qualitatively and quantitatively) against its historical site performance, fleet-level performance, if applicable, and industry performance. The licensee or other entity would be required to implement corrective actions if site FFD performance meets a licensee- or other entity-established threshold or to resolve a finding resulting from a qualitative review or audit in a manner that restores performance and corrects root causes, contributing causes, or both; (3) the requirements in proposed § 26.609, “Behavioral observation,” are more robust than those in § 26.407, “Behavioral observation,” of subpart K of part 26 and are proposed to synchronize with and reinforce the AA behavioral observation requirements in § 73.56, “Personnel access authorization requirements for nuclear power plants,” or the proposed requirements under § 73.120, “Access authorization program for commercial nuclear plants”; (4) a part 53 commercial nuclear plant that satisfies the § 26.603(c) criterion will be designed, operated, and secured with a radiological risk profile that is lower than that described in § 53.860(a)(2) and 
                        ( print page 86954)
                        perhaps will approach the radiological risk profile of an NPUF (which does not implement an FFD program); and (5) the NRC is aware that a part 53 commercial nuclear plant could be designed and constructed in such a manner to reduce reliance on an onsite security force to protect SSCs, NRC-licensed materials, and sensitive information, with enhanced capabilities for the detection, assessment, and delay of a DBT adversary. Regarding fatigue management requirements, work hour controls would be required for personnel at utilization and manufacturing facilities in accordance with the existing scoping criteria in § 26.4, “FFD program applicability to categories of individuals,” as revised in this proposed rule. The amended § 26.4 also would be used to determine whether an individual would be subject to drug and alcohol testing. The applicability of these scoping criteria for certain individuals (such as operators and maintenance personnel) would be determined by the licensee or other entity through its risk-informed evaluation process performed to assess the risk significance of the SSC upon which work is being performed or directed by the individual. These requirements also would be scaled based on the potential radiological consequences presented by the facility. However, fatigue management would be applied to all individuals subject to the FFD program, similar to FFD program implementation by the current fleet of commercial nuclear plants because fatigue management is a proactive requirement designed to help prevent on-shift impairment through work hour scheduling and time off. The behavioral observation program (BOP) would be the principal requirement to provide reasonable assurance that individuals on shift are not mentally or physically impaired due to fatigue, which in any way could adversely affect their ability to safely and competently perform their duties. The NRC is proposing subpart M of part 26 for facilities licensed under part 53, in lieu of subjecting all part 53 licensees to the same part 26 requirements that apply to facilities licensed under part 50 or 52, for four principal reasons. First, subpart M of part 26 would apply FFD requirements in a risk-informed manner commensurate with the radiological consequences presented by facilities licensed under part 53. This regulatory strategy is consistent with the current part 26, which provides a comprehensive set of deterministic requirements for licensees and other entities at facilities that are operating. This approach is also consistent with the current subpart K of part 26, which provides a more flexible framework for nuclear power reactors under construction, where the probabilities of serious radiological accidents are lower and consequences from such accidents are less severe than at operating plants. Second, subpart M of part 26 would enable a part 53 licensee or other entity to implement innovative drug testing technologies and behavior observation techniques while continuing to demonstrate compliance with the part 26 performance objective in § 26.23(b) of providing reasonable assurance that individuals are not under the influence of any substance or mentally or physically impaired from any cause, which in any way adversely affects their ability to safely and competently perform assigned duties. These technologies include drug and alcohol testing using oral fluid, urine, and hair specimens; screening using point of collection testing and assessment (POCTA) devices; and monitoring using passive drug and alcohol detection instrumentation. Part of the basis to enable the use of innovative drug and alcohol testing technologies is to maintain FFD program effectiveness should the staff size at a part 53 commercial nuclear plant be small and challenge the effective implementation of the behavioral observation and drug and alcohol testing programs. Also, a commercial nuclear plant that is sited at a geographically remote location may present additional challenges to behavioral observation and drug and alcohol testing that are not presented by traditional LWR facilities licensed under part 50 or 52, such as: efficiency of postal services for shipping and controlling biological specimens; proximity to drug and alcohol collection facilities that are reasonably equivalent to that described in subpart E of part 26; availability of internet and cellular services to enable same-time discussions among the Medical Review Officer (MRO), donor, and laboratory; accessibility to substance abuse treatment services described in subpart H of part 26; and proximity to an MRO (or management and clinical staff) to evaluate potential impairment caused by fatigue and/or substance use or abuse, for-cause and post-event occurrences, and the individual's potential to return to duty. A part 53 commercial nuclear plant that is sited in a geographically remote location and has a small staff size may present implementation challenges and the potential for small group dynamics to impact FFD program effectiveness. Particularly in isolated environments, psychological phenomena known as “groupthink” may take effect and could impact the effectiveness of BOPs and the ability to effectively manage safety culture. For example, in circumstances where small staffs are drawn from the same small town and thereby have a potentially narrow experience base, it could be challenging to maintain a safety conscious work environment in which personnel feel free to raise safety concerns without fear of retaliation, intimidation, harassment, or discrimination, and organizations may resultingly experience groupthink-like effects. Groupthink is particularly prevalent among cohesive and insulated groups that experience high levels of decisional stress.[6]
                        
                         Small staffs at part 53 commercial nuclear plants may therefore be more susceptible to groupthink if they are working in an isolated environment where decision-making pressures may be high. Groupthink could have adverse effects on workplace safety culture, as studies show that individuals will be more hesitant to speak out against practices they deem unsafe for fear of deviating from group norms.[7]
                        
                         Individuals may also be unaware of systematic biases in the group decision-making process and may then be less likely to scrutinize the potential risks of the group's decision or sufficiently contemplate alternative paths of action.[8]
                        
                         Furthermore, the literature indicates that groups make riskier decisions than individuals acting alone due to the diffusion of responsibility among group members.
                        9( print page 86955)
                        This phenomenon, known as “the risky shift,” also runs counter to a safety culture. Accordingly, “groupthink” and “the risky shift” may lead to group behaviors that render behavioral observation less effective. As such, alternative approaches to behavior observation programs, such as the utilization of video-based surveillance by individuals separate from the onsite work unit, could serve to mitigate potential issues associated with groupthink. The incorporation of remote observation, performed by individuals physically separate from the site, could help to bring in independent and objective perspectives and help to break patterns of thought and communication that may result in groupthink. Even without the influence of small group dynamics, there are other practical constraints to implementing FFD requirements, such as random drug and alcohol testing, among small staffs. Random testing is less effective when applied to small staff sizes because it may be easier for staff to communicate and predict when individuals will be subject to drug and alcohol testing. Furthermore, if a facility is sited in a remote location, program implementation could be challenged by the following factors: limited mail services to laboratories certified by the U.S. Department of Health and Human Services (HHS), availability of local clinical or medical options for treatment and determinations of fitness by an MRO or Substance Abuse Expert, and use of offsite drug and alcohol collection facilities. The increased potential for small staff sizes to impact FFD policy compliance warrants an approach to FFD that emphasizes performance over prescriptive requirements that may be ineffective or infeasible at these facilities. Therefore, the NRC proposes the subpart M of part 26 framework to provide a performance-based approach to FFD. For example, proposed § 26.603(d) would use existing part 26 auditing requirements and the reporting requirement in § 26.717, “Fitness-for-duty program performance data,” and clarify how FFD performance data would be used to maintain or improve, if necessary, FFD program effectiveness. Specifically, § 26.603(d) would require each licensee and other entity that elects to implement subpart M of part 26 to monitor and assess their site-specific performance against the preceding year's site performance, the licensee's most recent fleet-level performance, and the most recent industry performance. Licensees and other entities would use these datapoints to develop performance measures, which would be qualitative descriptions of the specific FFD program elements, and threshold values for each performance measure that, if exceeded, would indicate a performance deficiency. Each licensee and other entity would compare its site's current performance data against the performance measures and, if a threshold is exceeded, the licensee or other entity would be required to take corrective actions to restore performance. Also, the NRC proposes a change control requirement to allow a licensee or other entity to change its subpart M of part 26 FFD program while ensuring that FFD program effectiveness is maintained. Lastly, subpart M of part 26 would consolidate the applicable FFD requirements by placing in one subpart all proposed part 26 requirements (either new requirements or cross-references to existing part 26 requirements) for part 53 licensees and other entities. This should help licensees and other entities implement the requirements because it would enable easy cross-reference to similar requirements in other subparts that are being implemented by non-part 53 licensees and entities subject to part 26. Understanding how other licensees or other entities implement similar FFD requirements may facilitate the sharing of operating experience in program implementation. The use of innovative technologies and a risk-informed performance-based framework parallels the considerations presented in the Advanced Reactor Policy Statement. As stated in the policy statement, “[S]implified systems should facilitate operator comprehension, reliable system function, and more straightforward engineering analysis.” Furthermore, these same attributes may reduce potential radiation exposures, help prevent the theft of nuclear materials, and use technology and design innovations. Should these components and systems be designed, implemented, and maintained to minimize reliance on human actions and leverage technology and innovation, then the robust and prescriptive FFD requirements in, for example, subparts B, “Program Elements,” and E of part 26 could be scaled to the part 53-licensed facility and its operation. This strategy would be implemented in the subpart M of part 26 framework. Even though current subpart K of part 26, provides for FFD requirements commensurate with the radiological consequences presented by a nuclear power plant construction site, proposed subpart M of part 26 would not allow part 53 licensees and other entities to implement the requirements in subpart K. The principal reasons are that (without significant changes to subpart K that would be outside the scope of this rulemaking): (1) subpart K does not apply to holders of MLs who assemble or test a reactor; (2) subpart K only applies during construction, whereas subpart M would apply during construction, operation, and decommissioning through implementation of the insider mitigation program (IMP) required by § 73.55 or proposed § 73.100; (3) subpart K does not address training, authorization as defined in § 26.5, and MRO performance; (4) subpart K does not expressly authorize the use of innovative drug and alcohol testing technologies; (5) subpart K does not describe the use of time-dependent alcohol limits or special analysis testing of dilute urine specimens; and (6) subpart K has less rigor in the protection of worker rights and sensitive information than that proposed in subpart M. Despite the differences between subparts K and M of part 26, the requirements in subpart M would be essentially equivalent to many in subpart K that were implemented by the licensees of Vogtle Nuclear Station and V.C. Summer Nuclear Station when they were constructing four commercial nuclear power reactors and NRC inspection and operating experience evaluation determined that the use of subpart K contributed to adequately protecting the public health and safety and the common defense and security. Further, given the risk profile posed by facilities licensed under part 53 and the proposed additional requirements in subpart M of part 26 that were developed from operating experience and other part 26 subparts (but are not included in subpart K of part 26), the NRC concludes that if licensees and other entities effectively implement the proposed requirements in subpart M of part 26, then individuals subject to the rule should be fit for duty and trustworthy and reliable. B. Proposed Changes to Part 26, Subparts A Through E and I Section 26.3(d) is the applicability paragraph for contractor/vendors (C/Vs) who implement FFD programs or program elements, to the extent that the licensees and other entities specified in § 26.3(a) through (c) rely on those C/V FFD programs or program elements to meet the requirements of part 26. Section 26.3(d) would be amended to 
                        ( print page 86956)
                        address part 53 licensees and other entities in proposed § 26.3(f). Proposed § 26.3(f) would place part 53 licensees or other entities within the scope of part 26. For licensees and other entities of a part 53 commercial nuclear plant, except a holder of an ML, the FFD program would be required to be implemented no later than the start of construction activities. The holder of an ML would need to implement its FFD program before commencing activities that assemble a reactor. Current § 26.4 describes FFD program applicability to categories of individuals. These categories are based on the duties, responsibilities, and the types of access an individual may possess. The NRC proposes to amend § 26.4 to include licensees and other entities described in § 26.3(f). The NRC expects that not all categories of individuals described in current § 26.4 would be applicable to all part 53 facilities. The NRC is proposing regulatory guidance in DG-5073, “Fitness-of-Duty Programs for Commercial Nuclear Plants and Manufacturing Facilities Licensed Under 10 CFR part 53,” and DG-5078, “Fatigue Management for Nuclear Power Plant Personnel at Commercial Nuclear Plants Licensed Under 10 CFR part 53,” to help address program applicability to certain individuals. Section 26.4(a)(1) and (a)(4) would be amended to account for the possibility that certain individuals may perform or direct the performance of operational and maintenance activities from a remote facility (for example, a remote-control station) for licensees or other entities licensed under part 53. The framework of the current part 26 does not account for individuals who perform operating and maintenance duties at remote facilities. Although current § 26.4(a)(1) does not limit the operating of applicable SSCs to onsite operating, § 26.5 limits the definition of “
                        *Maintenance,*
                        ” for the purposes of § 26.4(a)(4), to include only “onsite maintenance activities.” In the 2008 part 26 final rule preamble, the NRC explained that the work hour requirements apply to those individuals who perform maintenance activities within the licensee's owner-controlled area. Furthermore, regarding the direction of applicable operations and maintenance activities, current § 26.4(a)(1) and (4) address only individuals who perform “onsite direction.” Under the proposed amendments to part 26, the limitation of “onsite” activities to those performed within the owner-controlled area would still apply to facilities licensed under part 50 or 52. However, for licensees and other entities described in § 26.3(f), the NRC would remove the “onsite” limitation to include activities performed both within the owner-controlled area as well as operations and maintenance duties performed at remote facilities where safety-significant systems and components are expected to be operated within the design basis of the commercial nuclear plant. In the 2008 part 26 final rule, the purpose of limiting “directing” activities to those “directing” activities that are conducted onsite was to avoid requiring work hour controls for individuals performing incidental duties, consistent with § 26.205(b)(5), from an offsite location in instances where those duties might be considered to be “directive” in nature. Under the proposed amendments to part 26, the exclusion of incidental duties while calculating work hours would still be applicable for licensees and other entities licensed under part 53. However, for these licensees and other entities, beyond instances of incidental duties, the direction of operations and maintenance activities associated with safety-significant SSCs, when performed at remote facilities, would be considered in an equivalent fashion as direction performed at non-remote facilities, for the purposes of administering work hour controls. Proposed § 26.4(b) would include in an FFD program individuals who are granted unescorted access to the protected area of a facility licensed under part 53 and do not perform or direct the performance of the duties described in § 26.4(a). This requirement would contribute to the defense-in-depth regulatory framework that helps provide that individuals who have unescorted access are fit for duty, trustworthy, and reliable. For example, the NRC is proposing amendments to part 73 to require a part 53 licensee to subject individuals to a series of reviews to help determine whether those individuals are trustworthy and reliable before granting them unescorted access to the facility's protected area. The NRC would amend § 26.4(c) to include in an FFD program individuals who are assigned to physically report to the part 53 licensee's emergency response facility (or facilities) or participate remotely in emergency response activities, and individuals without unescorted access to the part 53 facility who, remotely or otherwise, make decisions and/or direct actions regarding plant safety or security. Part 53 commercial nuclear plants may be licensed for and rely upon offsite facilities to fulfill the role of a Technical Support Center or Emergency Operations Facility. Therefore, the proposed rule would account for such offsite facilities or remotely performed activities. Further, the use of personnel to operate systems and components, maintain and surveil SSCs, and respond to plant conditions and security events may be different than those included in the Technical Support Center or Emergency Operations Facility team for power reactors currently licensed under part 50 or part 52. For the individuals whose duties for the licensees and other entities in § 26.3(c) require the individuals to have the types of access or perform the activities listed in § 26.4(e)(1) through (6) at the location where the commercial nuclear plant will be constructed and operated, current § 26.4(e) requires them to be subject to an FFD program that satisfies all the requirements of part 26 except subparts I and K. The NRC would amend § 26.4(e) to except subpart M as well as subparts I and K. The NRC would also amend § 26.4(e) to include in an FFD program the individuals whose duties for the licensees and other entities in § 26.3(f) require the individuals to have the types of access or perform the activities listed in § 26.4(e)(1) through (6) or perform construction activities as defined in § 26.5. Section 26.4(e)(4) would be revised to include in an FFD program individuals who witness or determine inspections, tests, and analyses certifications required under part 53 because current § 26.4(e)(4) includes the individuals who perform the same duties under part 52. The proposed rule would amend § 26.4(f) to require individuals who construct or direct the construction of safety- or security-related SSCs at facilities licensed under part 53 to be subject to an FFD program under subpart M of part 26 or an FFD program that demonstrates compliance with all of the requirements of part 26 except for subparts I, K, and M of part 26. Section 26.4(g) is the applicability paragraph for FFD program personnel (
                        *e.g.,*
                         the FFD manager, MRO, and technicians) and persons who perform AA determinations (
                        *e.g.,*
                         the licensee- or other entity-designated Reviewing Official). This section would be amended to address part 53 licensed facilities. Specifically, a part 53 licensee or other entity would use FFD program personnel to implement its FFD program as well as other assigned individuals who are not involved in the day-to-day operations of the program to implement specific elements of its FFD program, such as the collection of a 
                        ( print page 86957)
                        specimen for drug or alcohol testing. These individuals would be held accountable for program implementation, including consistent implementation of protections afforded to all individuals subject to the FFD program. Section 26.4(h) would be amended to include subpart M of part 26. The NRC proposes to include several new definitions in § 26.5, “Definitions,” and amend some existing definitions. The NRC is proposing to add a definition for “
                        *Biological marker.*
                        ” The proposed definition would be consistent with “
                        *Biomarker*
                        ” defined by the HHS in its Mandatory Guidelines for Federal Workplace Drug Testing (HHS Guidelines) using oral fluid as the biological specimen to be tested (84 FR 57554; October 25, 2019). However, the proposed definition for § 26.5 would add that the endogenous substance used to validate that the biological specimen “was produced by the donor” because subpart M of part 26 proposes to have the MRO evaluate any discrepant biological marker identified in a biological specimen collected from a donor. The NRC is proposing a definition for the word “
                        *Change*
                        ” as used in the proposed § 26.603(e), “FFD program change control,” process. The proposed definition would be consistent with the definition of “
                        *Change*
                        ” for a part 50 or 52 licensee's emergency plans in § 50.54(q)(1)(i). The NRC proposes to revise the definition of “
                        *Constructing or construction activities*
                        ” to clarify that for licensees or other entities in § 26.3(f), the definition of “
                        *Construction*
                        ” would be that as proposed in § 53.020. The definitions of “
                        *Contractor/vendor*
                        ” (C/V) and “
                        *Other entity*
                        ” would be revised to make them applicable to part 53 licensees. A holder of an ML under part 53 could be a C/V under the proposed C/V definition. The NRC is proposing a definition for “
                        *Illicit substance*
                        ” because this phrase is used in subpart M of part 26 and would address substances that cause impairment and possible addiction but are not an “illegal drug” as defined in § 26.5. This proposal is based on operating experience where individuals have admitted to using common household, non-drug substances to achieve a high or satisfy an addiction. These common household items include, but are not limited to nitrous oxide, butane, propane, glue, paint vapors, lighter fluid, nail polish remover, degreasers, permanent markers, and methyl alcohol (which is found in hand sanitizer and mouthwash). The definition of “
                        *Questionable validity*
                        ” would be revised to make it applicable to an FFD program implemented under subpart M of part 26, which would include all biological specimens. The NRC is proposing a definition for “
                        *Reduction in FFD program effectiveness*
                        ” because this phrase, similar to the proposed definition for “
                        *Change,*
                        ” is used in proposed § 26.603(e). The proposed definition is generally consistent with the definition of “
                        *Reduction in effectiveness*
                        ” provided for emergency plans in § 50.54(q)(1)(iv). The proposed rule would make the current definition of “
                        *Reviewing official*
                        ” applicable to those licenses and other entities in § 26.3(f). The current part 26 definition of “
                        *Safety-related structures, systems, and components*
                        ” would be amended to use the NRC's proposed definition in § 53.020 for the part 53 licensees and other entities described in § 26.3(d) and (f). The NRC would amend the definition of “
                        *Security-related SSCs*
                        ” in § 26.5 to make it applicable to a licensee or other entity described in § 26.3(d) and (f). The NRC proposes a definition for “
                        *Special Nuclear Material*
                        ” that would refer to the definition in § 70.4, “Definitions,” of part 70 to ensure consistency. The NRC is proposing a revision of the definition of “
                        *Unit outage*
                        ” to account for the potential use of commercial nuclear plants for purposes other than electricity generation. Section 26.21, an applicability statement for part 26 FFD programs, would be amended to include licensees and other entities described in § 26.3(f) that choose to implement an FFD program that implements all part 26 requirements, except those in subparts K and M of part 26. Section 26.51, “Applicability,” would be amended to apply to licensees and other entities described § 26.3(f) that elect not to implement the requirements in subpart M of part 26 for the categories of individuals in § 26.4 and those licensees and other entities that elect to implement the requirements in § 26.605. Section 26.53(e), (e)(1) and (3), and (g) through (i), which are general provisions for granting and maintaining authorization, would be amended to apply to licensees and other entities described § 26.3(f). Section 26.63(d), a suitable inquiry requirement, would be amended to apply to licensees and other entities described § 26.3(f). Section 26.73, the applicability statement for subpart D of part 26, would be amended to apply to licensees and other entities described § 26.3(f) that elect not to implement the requirements in subpart M of part 26 for the categories of individuals in § 26.4 and those licensees and other entities that elect to implement the requirements in § 26.605(b). Section 26.81, the purpose and applicability statement for subpart E of part 26, would be amended to apply to licensees and other entities described in § 26.3(f) that elect not to implement the requirements in subpart M of part 26 for the categories of individuals in § 26.4 and those licensees and other entities that implement proposed § 26.605(a) or (b). The subpart E requirements to be implemented are listed in proposed § 26.607(c)(2)(i) and (c)(2)(ii) and (c)(3). Section 26.201, the applicability statement for subpart I of part 26 would be amended to apply to licensees and other entities described in § 26.3(f). Also, the applicability statement would be divided into two paragraphs for clarity. The NRC proposes to add § 26.202, “General provisions for facilities licensed under part 53,” for licensees or other entities described in proposed § 26.3(f) that elect to implement the requirements in subpart I of part 26 in accordance with § 26.604 and § 26.605. Section 26.202 would establish requirements equivalent to those in current § 26.203, “General provisions,” which is applicable to part 50 and 52 licensees. The NRC would add the separate § 26.202 because § 26.203 refers to various requirements under subpart B of part 26, which would not be applicable to facilities licensed under part 53 that implement subpart M of part 26. Additionally, § 26.202(c), “Training and assessments,” unlike § 26.203(c), “Training and examinations,” would not include a comprehensive examination requirement because trainee assessment is conducted as part of a SAT that would be required as proposed under the FFD program training requirements in § 26.608. Proposed changes in §§ 26.205, 26.207, and 26.211 would add references to new requirements in subparts I and M of part 26 that would be applicable specifically to licensees and other entities in § 26.3(f). The NRC would not change the specific provisions for work hour requirements in current § 26.205(d). However, as addressed in the discussion of proposed changes to § 26.4(a), whether a licensee or other entity under part 26 would need to implement work hour controls 
                        ( print page 86958)
                        for certain individuals or groups would be dependent, in part, on determinations reached by that licensee's risk-informed evaluation process. Proposed changes to §§ 26.207(a)(1)(ii) and 26.211(b) would allow licensees and other entities in § 26.3(f) to perform face-to-face assessments to support the approval of work hour control waivers and the conduct of fatigue assessments, respectively, using electronic communications. These proposals would allow supervisors to conduct such assessments from a remote location under appropriate circumstances. Such remotely conducted assessments would need to be supported by someone who is present in-person with the individual being assessed and who is trained in accordance with the requirements of either § 26.29 and § 26.203(c) or § 26.608 and § 26.202(c). The reasoning for these proposals and the associated need for in-person support to augment electronic communications is addressed further in the preamble discussion of proposed § 26.619. C. Proposed Requirements for Part 26, Subpart M The proposed rule would add a new subpart M to part 26 that would provide alternative FFD requirements for part 53 licensees and other entities. Proposed § 26.601 would make subpart M of part 26 applicable to part 53 licensees and other entities, at their discretion. If a licensee or other entity in § 26.3(f) does not elect to implement an FFD program that demonstrates compliance with the requirements of subpart M, then the individuals specified in § 26.4 would be subject to an FFD program that demonstrates compliance with all part 26 requirements, except for those requirements in subparts K and M. Proposed § 26.603(a) would require an applicant to provide a description of its FFD program and its implementation within its application for a license. This requirement is equivalent to the existing requirements in §§ 26.401(b) and 52.79(a)(44). The entities that would be required to submit these FFD program descriptions are certain applicants that would comply with the part 53 application requirements in subpart H. In subpart H, § 53.1309(a)(6) would require an applicant for a CP to provide a description of its FFD program in its PSAR. Under §§ 53.1279(b)(4), 53.1369(x), and 53.1416(a)(24), an applicant for an ML, OL, and COL, respectively, would be required to provide a description of its FFD program in its FSAR. Unlike an application for a license, a description of an FFD program does not receive NRC review for possible approval. The applicant provides the NRC with information about the applicant's proposed FFD program to inform the NRC's inspection program and to demonstrate that the FFD program will be effectively implemented before a licensee or other entity commences any activity making individuals at the NRC-licensed facility subject to the FFD program. Proposed § 26.603(a)(1) would require a summary description of the analysis described in § 26.603(c), if performed. The analysis should describe the operation of the facility. This would include informing the Commission of: (1) the principal individuals assigned by job title (work category) and a summary description of the human actions (
                        *e.g.,*
                         monitoring, operating, responding, surveillance, oversight, etc.) that they perform to maintain the facility in a safe operating or shutdown condition; (2) the principal individuals by job title and a summarized description of the human actions to secure and protect the facility (without providing sensitive information); (3) the estimated total population of individuals subject to the FFD program and per shift by job description; and (4) references to supporting documentation. The purpose of these descriptions is to enable an NRC assessment of the licensee's or other entity's analysis and the required human actions to operate, monitor, surveil, maintain, and secure the facility within its design and licensing basis so that if an operational or security-related event were to occur, the facility would respond as designed and licensed and the calculated radiological dose consequences would not exceed the consequences described in § 53.860(a)(2). This is important because facilities that implement § 26.604 are expected to have very small staff sizes and may be sited in geographically remote locations, both of which could challenge effective implementation of the FFD program. Proposed § 26.603(a)(2) would require the applicant to state what FFD program it plans to implement. Proposed § 26.603(a)(3) would require a discussion that informs the NRC of the applicability of the applicant's FFD program to individuals who perform safety- or security-significant activities. This description should summarize any key differences between the staff at the site and any remote facility and the categories of individuals in § 26.4. The principal purpose of providing this description would be to inform the NRC of any substantial differences in the applicability of the FFD program to the categories of individuals in § 26.4. Proposed § 26.603(a)(4) would require a description of the drug and alcohol testing and fitness determination process to be implemented through the licensee's or other entity's procedures, including the collection and testing facilities to be used, biological specimens to be collected, and sanctions to be imposed upon a confirmed FFD policy violation. This process includes how individuals who test positive for a drug or alcohol will be evaluated before being afforded unescorted access to the protected area to perform or direct those duties or responsibilities making them subject to the FFD program. The principal purpose of describing this return-to-duty process is to inform the NRC of the behavioral observation strategy (for those facilities that implement § 26.604) and/or drug screening and testing strategy. Proposed § 26.603(a)(5) would require a summary description of the applicant's planned PMRP. This description must provide the performance measures and thresholds that the applicant intends to use. Proposed § 26.603(b) would establish when the FFD program must be implemented and the longevity of the FFD program. This proposal is equivalent to the current § 26.3, which states, in part, when licensees and other entities must begin implementing their FFD programs. Unlike the current part 26 regulations, proposed § 26.603(b) would expressly state that an FFD program would not be applicable during decommissioning of a part 53 facility for licensees and other entities specified in § 26.3(f). However, licensees of facilities licensed to operate a reactor should be aware that the physical protection program under § 73.55, “Requirements for physical protection of licensed activities in nuclear power reactors against radiological sabotage,” and under proposed § 73.100 include a requirement for the implementation of an IMP, even during decommissioning. Proposed § 26.603(b) would also require the holder of an ML to implement its FFD program no later than the start of activities that assemble a reactor. The holder of the ML should establish in its procedures when reactor assembly commences and what constitutes assembly. For example, the FFD program would not need to be implemented for the receipt, storage, inspection, and staging of components and systems used to assemble (
                        *i.e.,*
                         build or fabricate) the reactor because this is not a current requirement for LWR facilities licensed under part 50 or 52. Furthermore, the NRC currently does 
                        ( print page 86959)
                        not require that an FFD program be applied to the assembly or manufacturing of components (or basic components as defined in § 21.3), or systems that were fabricated or assembled outside the footprint of a commercial power reactor, and this regulatory position would also apply to a manufacturing facility. Proposed § 26.603(c) would require the applicant, licensee, or other entity seeking to implement an FFD program under § 26.604 to perform a site-specific analysis to determine whether the facility and its operation satisfy the criterion in § 53.860(a)(2). If the analysis is performed and demonstrates that the radiological consequences presented by the facility and its operation satisfy the criterion, then the licensee or other entity could implement the FFD program detailed in § 26.604. If the analysis does not demonstrate that the facility and its operation satisfy the criterion, then the licensee or other entity must implement the FFD program described in either § 26.605 or subparts A through I, N, and O of part 26. Proposed § 26.603(c) would also require licensees and other entities that implement proposed § 26.604 to update the technical analysis used to justify compliance with the criterion in § 53.860(a)(2). This analysis would be updated to reflect changes made to the staffing, FFD programs, or offsite support resources described in the analysis to show that the facility and its operation continue to satisfy the criterion. This is important because facility, operation, or staffing changes outside FFD program implementation (
                        *e.g.,*
                         changes in the facility safety analysis, physical protection strategies, or the security plan, implementing procedures, or contingency response strategies) could adversely impact the licensee's or other entity's documented analysis demonstrating that the facility and its operation satisfy the criterion if event sequences require human action. Proposed § 26.603(d) would require the establishment of a PMRP. The concept of a PMRP is not new. This requirement would consolidate for part 53 the requirements in current §§ 26.41, “Audits and corrective actions”; 26.415, “Audits”; 26.717, “Fitness-for-duty program performance data”; and 26.183(c), which describes MRO responsibilities. The proposal would state that the licensee or other entity must monitor the effectiveness of its FFD program by comparing performance data against performance measures and thresholds. The development of quantitative thresholds would be new, but this is born from licensees and other entities with facilities licensed under parts 50 or 52 already collecting, reviewing, and reporting FFD performance data. Additionally, the benefit of quantitatively measuring FFD program performance against established thresholds benefits a licensee's and other entity's determination of whether they are maintaining FFD program performance in a manner that demonstrates compliance with the performance objectives in § 26.23. The NRC is proposing the PMRP because the subpart M of part 26 requirements would enable a high degree of flexibility in FFD program implementation (
                        *e.g.,*
                         drug testing). A licensee or other entity would not only have options in the type of FFD program they may implement under part 26, but they would have options in the types of biological specimens they may test for drugs, where to collect the biological specimens (
                        *e.g.,*
                         at the NRC-licensed facility or offsite at a local hospital or clinic), and the use of collection and assessment devices to screen individuals for drugs and alcohol. These FFD program flexibilities could cause FFD programs under subpart M of part 26 to become very site-specific, necessitating performance measures to enable the licensee or other entity to maintain the effectiveness of its FFD program. Fitness-for-duty program effectiveness would be determined by comparing actual performance against the performance measures and thresholds. The result of that comparison would inform licensee or other entity decisions whether to change FFD program elements to address a performance deficiency. Also, the thresholds would have sufficient margin, based on operating experience, before conditions adverse to safety and security may occur should an individual be identified as impaired or not trustworthy and reliable. The potential of a human-related failure causing a condition adverse to safety and security is dependent on the duties and responsibilities of the individual and the defense-in-depth designed to prevent or mitigate an adverse consequence. The PMRP would account for this by requiring the review of FFD performance data, in part, by work category, C/V, and individuals employed by the licensee who are not a C/V as defined in § 26.5 (
                        *i.e.,*
                         a licensee employee). Proposed § 26.603(d)(1) would require the licensee or other entity to document and maintain its PMRP. Proposed § 26.603(d)(1)(i) would require that the performance measures be identified and designed to monitor FFD program performance. Proposed § 26.603(d)(1)(i)(A) would require the FFD program of a licensee or other entity subject to the requirements of § 26.604 to include monitoring of the BOP. The purpose of this monitoring is to help ensure that individuals subject to the FFD program are observing the behaviors of others, are being observed themselves, and are reporting FFD concerns to licensee- or other entity-designated individuals. The other performance measures would include occurrence of FFD policy violations evaluated by licensee employee, C/V, and labor category, and occurrence of individuals with potentially disqualifying information or who possessed an FFD prohibited item. Proposed § 26.603(d)(1)(i)(B) would require the FFD program of a licensee or other entity that is either subject to the requirements of § 26.604 and has implemented a drug testing program at its discretion, or is subject to the requirements of § 26.605, to include the performance measures identified in § 26.603(d)(1)(i)(A) and those necessary to monitor the effectiveness of the drug and alcohol testing program. The drug and alcohol measures would include the monitoring of FFD performance data for pre-access and random testing and subversion attempts by the categories of licensee employee, C/V, and labor category. Proposed § 26.603(d)(1)(ii) would require the licensee or other entity to establish thresholds for each performance measure. Initial thresholds must be based on FFD performance data from comparable facilities subject to part 26, the licensee's or other entity's fleet-level program performance if applicable, and industry FFD performance data. This provision introduces the requirement to “maintain FFD program effectiveness.” This terminology describes a performance-based regulatory strategy in which the licensee or other entity must initially establish a level of performance that is representative of other facilities in the licensee's fleet of facilities subject to part 26, if applicable, and the FFD performance of comparable facilities subject to part 26. Proposed § 26.603(d)(1)(iii) would require that the licensee or other entity evaluate FFD data as it is received to determine whether a threshold has been exceeded. Historical FFD performance data for the current LWR fleet indicates that, for particular work categories and employment types, few FFD policy violations occur per year. Therefore, for work categories that may be significant to worker safety (
                        *e.g.,*
                         radiation protection technicians), physical 
                        ( print page 86960)
                        protection (
                        *i.e.,*
                         security personnel), or safety (
                        *i.e.,*
                         NRC-licensed operators and individuals who perform or direct the performance of activities that a risk-informed evaluation process has shown to be significant to public health and safety), a single FFD policy violation could be a significant occurrence and warrant corrective actions. Based on licensee-submitted FFD-related reports under §§ 26.417, 26.419, 26.717, and 26.719, licensees and other entities with facilities licensed under parts 50 or 52 implement some form of corrective action that is typically scaled to the significance of the violation. These corrective actions have included counseling, follow-up drug and/or alcohol testing, remedial training, generic announcements to the workforce, and reviews of recently performed or directed work by the individual suspected of being impaired. Proposed § 26.603(d)(1)(iii) would require that the PMRP include a year-to-year comparison of FFD performance data to help provide assurance that an adverse trend in FFD program performance would be identified if occurring. This proposed requirement was developed from the annual FFD performance data reporting requirements in §§ 26.417(b)(2) and 26.717. In particular, the proposed year-to-year comparison of FFD performance data is equivalent to § 26.717(c), which requires, in part, licensees and other entities to analyze their performance data at least annually and take appropriate actions to correct any identified program weaknesses. Proposed § 26.603(d)(1)(iv) would require the licensee or other entity to perform and document quantitative and qualitative reviews. These reviews would be performed in three program areas: protections afforded to individuals subject to the FFD program, laboratory test results and MRO performance, and change control. The purpose of these reviews would be to specifically target performance within the three program areas to assess whether the outcomes resulting from the implementation of procedure requirements are contributing to FFD program effectiveness. The proposed reviews would not require the establishment of measures and thresholds because the reviews are expected to result in qualitative findings regarding program effectiveness. Qualitative findings and observations could still result in the consideration of corrective actions in the targeted program areas. Proposed § 26.603(d)(1)(iv)(A) would require the licensee or other entity to monitor whether its FFD program is affording appropriate protections to individuals subject to the FFD program. The review of these protections would include, in part, assessing the licensee's or other entity's protection of the following: privacy during the specimen collection process; specimen integrity, custody, and control; information gathered from FFD program implementation; and due process during appeals of FFD policy violations. Proposed § 26.603(d)(1)(iv)(B) would require, in part, a review of laboratory test results and MRO performance. Effective performance by the laboratory (
                        *e.g.,*
                         obtaining and communicating accurate test results) and MRO (
                        *e.g.,*
                         correct evaluation of the laboratory test results based on § 26.185 or HHS Guidelines) would result in three significant outcomes: (1) protection of the donor from an inaccurate FFD policy violation determination; (2) protection of the donor, other individuals, and the facility from potential harm should the donor be impaired or not trustworthy and reliable; and (3) a performance-based assessment of both the laboratory and MRO. This last outcome could facilitate actions to improve laboratory performance, MRO training under § 26.607(m), or both. Proposed § 26.603(d)(1)(iv)(B) would also require a comparative analysis between the POCTA screening result(s) and the corresponding specimen test results obtained from the HHS-certified laboratory if the POCTA indicated a positive, adulterated, substituted, or invalid screening result or discrepant biological marker, to assess the effectiveness of the POCTA and to inform MRO decisions under § 26.185 or § 26.607(m)(6). The results of this biennial review could also inform the conduct of laboratory audits. Proposed § 26.603(d)(1)(iv)(C) would require that the change control requirement in proposed § 26.603(e) be included in the biennial program review to help ensure that changes implemented over the life of the facility do not result in a reduction in program effectiveness even if a mitigating action was implemented for the specific change. This requirement was developed from §§ 26.137(f) and 26.713(d). This part of the review would require an assessment of all changes since the last review and their potential aggregated impact on FFD program effectiveness. For example, if last year the licensee elected to contract with a different MRO and this year the licensee implemented a new type of POCTA device, each of those program changes probably would not have resulted in a recognizable reduction in FFD program effectiveness. But, if the drug testing positivity rate (or FFD policy violations) for C/Vs decreased markedly during a future maintenance outage that required many C/Vs, then the reduction could indicate, for example, that the POCTA device was not as effective as determined by a forensic toxicologist review under §§ 26.603(e) and 26.607(h) or that the new MRO was improperly crediting prescription medication for laboratory-confirmed positive test results. Proposed § 26.603(d)(2) would state when the licensee or other entity must implement corrective actions. This requirement would be equivalent to the requirement in current § 26.415(b) and was developed from requirements contained in §§ 26.41(a) and (f), 26.127(e), 26.129(b)(1)(i), 26.137(f)(3) through (5), 26.155(a)(6), 26.157(e), 26.159(b)(1)(i), and 26.203(e)(2). Corrective actions must be implemented to correct root causes, contributing causes, or both. There is margin built into the FFD performance thresholds and qualitative factors (
                        *e.g.,*
                         to account for potential changes in drug and alcohol testing performance data when there is a large influx of C/Vs to perform maintenance) that may influence a licensee or other entity's causal determination for an occurrence. Thus, generalized or qualitative corrective actions may be implemented like informing management and placing a sufficiently descriptive summary of the occurrence in a corrective action program for future monitoring to assess recurrence. However, should the occurrence challenge safety or security or significantly exceed a performance threshold even when considering qualitative factors and margin, the licensee or other entity should implement more robust corrective actions to resolve the cause. An example of a challenge to safety or security would be the situation when an NRC-licensed operator or maintenance professional had operated, surveilled, or maintained safety-significant SSCs and was determined to have been impaired by behavioral observation or potentially under the influence of a narcotic as determined by an alcohol or drug test or screening result. Immediate corrective actions could include, but would not be limited to, a licensee or other entity assessment of the duties and responsibilities recently performed by the individual. Operating experience within the LWR operating reactor community demonstrates few FFD policy violations per year per site have been caused by individuals who perform or direct the performance of 
                        ( print page 86961)
                        safety or security-significant activities. Therefore, any such violations of the FFD policy in a particular work category in one year could be a significant performance deficiency. These violations could be even more significant at part 53 facilities that have a very small workforce subject to part 26. Proposed § 26.603(d)(3) would require the licensee or other entity to biennially assess and document its FFD performance monitoring program; this requirement was developed from § 26.41(b). This documented review would demonstrate that the performance measures and thresholds are appropriate based on site- and licensee's fleet-level program performance, if applicable, and industry performance and adjusted to maintain FFD program effectiveness. Also, as a result of this effort, the licensee or other entity would be in possession of lessons learned from fleet-level performance, if applicable, and industry performance that could contribute to their own performance assessment to maintain program effectiveness. Under proposed § 26.603(d)(3)(i), the identified program weaknesses and corrective actions resulting from the biennial review would be required to be summarized in the licensee's or other entity's annual report to the NRC in compliance with either § 26.417(b)(2) or § 26.717, as applicable. This information would inform the NRC of FFD program weaknesses to facilitate regulatory oversight and enable the NRC to aggregate industry data for use in a licensee or other entity PMRP. Proposed § 26.603(d)(3)(ii) would establish when the biennial PMRP review must be completed and when corrective actions from the review must be implemented. The NRC selected the May 15th date of odd-numbered years to help ensure that all FFD programs will maintain their previously determined performance measures and thresholds or reset them based on FFD program performance early in the year in which the biennial review was conducted. This would assist in obtaining quality FFD performance data over two annual reporting cycles and evaluating whether previous corrective actions were effective. In proposed § 26.603(e), the NRC proposes a change control requirement for subpart M of part 26 FFD programs. Requiring licensees and other entities to demonstrate compliance with certain requirements before implementing changes to their FFD programs would be necessary for two primary reasons. First, proposed changes to a licensee's or other entity's FFD program could affect the analysis performed by the licensee or other entity under proposed § 26.603(c), which helps determine the FFD program requirements that must be implemented. If this analysis changes, then the licensee's or other entity's FFD program requirements might change. Second, the requirements in subpart M of part 26 are performance based. Therefore, FFD program implementation may change periodically in response to societal changes in substance abuse or from PMRP implementation. Change control therefore relies on the licensee or other entity maintaining its procedures in a manner that details how its FFD program is to be implemented while incorporating changes, with documentation that justifies the changes to support the PMRP, audits, and NRC inspection. Proposed § 26.603(e)(1) would permit the licensee or other entity to implement changes to its FFD program if it performs and retains an analysis demonstrating that the change does not reduce the effectiveness of the FFD program or the change was necessitated or justified by a change to part 26, laboratory processes, or guidance issued by the HHS or NRC. The proposed change control requirement would enable flexibility in program implementation should the NRC or HHS change its drug testing procedures (as implemented by the licensee or other entity through its procedures) in response to changes in societal substance abuse or drug testing technologies. The proposed change control requirement was developed from the change control requirements in § 50.54(p) and (q)—the change control requirements for security and emergency plans, respectively. However, unlike these two requirements, the NRC does not review and approve a licensee's or other entity's FFD program or its implementing procedures, and the FFD program is not licensing-basis information as described in § 53.1300. Proposed § 26.603(e)(2) would require that if a change reduces FFD program effectiveness, then the licensee must implement a mitigating strategy so the FFD program, as revised, will continue to demonstrate compliance with the performance objectives in § 26.23 and not result in a reduction in program effectiveness. Proposed § 26.603(e)(3) would prohibit, with one exception, the use of the change control process to reduce the minimum panel of drugs to be tested and would reference the drugs listed in proposed § 26.607(c)(1). Proposed § 26.607(c)(1) would reference current § 26.31(d)(1), which states that, at a minimum, licensees and other entities shall test for marijuana metabolite, cocaine metabolite, opioids (codeine, morphine, 6-acetylmorphine, hydrocodone, hydromorphone, oxycodone, and oxymorphone), amphetamines (amphetamine, methamphetamine, methylenedioxymethamphetamine, and methylenedioxyamphetamine), phencyclidine, and alcohol. The testing of these drugs and drug metabolites, except phencyclidine, and alcohol is necessary for the FFD program to remain effective. Also, there is no proposed subpart M of part 26 requirement stating that this panel of drugs and drug metabolites needs to consist of only scheduled drugs.[10]
                        
                         This flexibility would account for the situation where an impairing substance becomes prevalent in society and a licensee or other entity elects to add the substance to their panel of substances to be tested prior to it being scheduled by the Drug Enforcement Administration. The exception in proposed § 26.603(e)(3) would be that, should HHS elect to remove phencyclidine from the panel of drugs and drug metabolites to be tested, a licensee or other entity could make this change in its FFD program without resulting in a reduction in FFD program effectiveness. This outcome would be justified based on the very infrequent occurrence rate of FFD policy violations due to phencyclidine use since 2010. However, if HHS proposes to remove a class of drugs from the panel of drugs to be tested that is listed in § 26.31(d)(1), except for phencyclidine, then a licensee or other entity may not make a similar change to its panel of drugs to be tested, because this change would be a reduction in FFD program effectiveness even with a mitigative strategy implemented. Changes in the HHS panel of drugs and drug metabolites to be tested may also shift from one metabolite to a 
                        ( print page 86962)
                        different metabolite for the same drug class (
                        *e.g.,*
                         amphetamines, opioids) to be tested. Should HHS issue such a change to its panel, this would not be expected to result in a reduction in FFD program effectiveness because HHS would be targeting a more prevalent or effective metabolite in its drug testing program. This situation could occur as HHS gathers more operating experience from Federal Government implementation of its HHS Guidelines, or data generated by drug testing laboratories and federally mandated drug testing programs required by Federal agencies such as the NRC and U.S. Departments of Transportation, Energy, and Defense. Proposed § 26.603(e)(4) would require that change control records be maintained for a 5-year record retention period based on the current NRC practice to conduct triennial inspections of licensees' and other entities' FFD programs. This would afford the NRC an opportunity to review the licensee's or other entity's determination that FFD program changes have not reduced the effectiveness of their FFD program. Licensees and other entities would also be required to summarize each change made under proposed § 26.603(e) in their annual FFD performance reports required by § 26.617(b)(2) or § 26.717, as applicable. Proposed § 26.604 would establish the minimum set of FFD program requirements for licensees and other entities who have a documented analysis that demonstrates that the facility and its operation satisfy the criterion in § 53.860(a)(2). For these licensees, compliance with the performance objectives in § 26.23 would be ensured through the BOP; defense-in-depth measures proposed in subpart M of part 26 like the PMRP, change control, and audits; and other requirements, such as those for AA, physical protection, and licensed operators. The adequacy of these measures in satisfying the performance objectives is supported by operating experience, which demonstrates margin between an FFD-related occurrence and a condition adverse to safety or security, as illustrated by for-cause, post-event, and random testing data. A facility that satisfies the criterion in proposed § 53.860(a)(2) would present a smaller potential radiological consequence than a facility that does not satisfy the criterion, so the requirements in proposed § 26.604 are scaled to the lower risk presented consistent with the Commission's Advanced Reactor Policy Statement. The disadvantages of implementing the FFD program described in proposed § 26.604 would be few. Since drug and alcohol testing would not be required, behavioral observation would be the keystone requirement in this performance-based framework to provide that individuals are fit for duty, trustworthy, and reliable, and can safely and competently perform the duties and responsibilities making them subject to the FFD program. If not, the individuals would be assessed in accordance with the licensee's or other entity's procedures similar in manner to that required by subpart K of part 26, and the proposed PMRP would require corrective actions should a threshold be exceeded. If a licensee or other entity elects not to perform the analysis in proposed § 26.603(c) to determine whether it satisfies the criterion in proposed § 53.860(a)(2); performs the analysis and finds that the facility and its operation does not satisfy the criterion in proposed § 26.603(c); or is a holder of an ML, the licensee or other entity could not implement the FFD program described in § 26.604. Instead, the licensee or other entity would implement either the program described in proposed § 26.605 or an FFD program that demonstrates compliance with all the requirements in current subparts A through I, N, and O of part 26. Proposed § 26.605 would establish requirements in a graded manner similar to the regulatory framework established by the requirements in subparts A through I, N, O, and K of part 26. This existing graded approach consists of an FFD program for construction of a commercial nuclear plant and a more robust program that must be implemented before reactor operation. The former is the FFD program in proposed § 26.605(a), and the latter is proposed § 26.605(b). Like that for an FFD program under § 26.604, the FFD program under § 26.605 would include FFD program elements similar to those in subpart B of part 26, but the proposed requirements are less prescriptive, enabling more flexibility in program implementation like that offered in subpart K of part 26. For example, the requirements in subpart B of part 26 are explicit requirements for, in part, the collection and analysis of urine specimens. Subpart B of part 26 does not enable the use of oral fluid for drug testing or screening, except under very limited situations as described in subpart E of part 26, or the use of hair specimens, unlike proposed § 26.605. Proposed § 26.605 would require drug and alcohol testing based on either the requirements in part 26 or the HHS Guidelines. The principal benefit of the proposed § 26.605 FFD program is that it would provide a regulatory framework that is consistent with the radiological consequences for a facility that does not satisfy the criterion in proposed § 53.860(a)(2) while affording flexibilities in the conduct of drug and alcohol testing. Proposed § 26.605(a) would apply to licensees and other entities who perform the § 26.603(c) analysis and satisfy the criterion in § 53.860(a)(2) but decide not to implement the FFD program described in proposed § 26.604, licensees and other entities who do not perform the § 26.603(c) analysis, and licensees and other entities who perform the analysis but their analysis does not demonstrate that their facility and its operation satisfy the criterion in § 53.860(a)(2). These entities must establish, implement, and maintain an FFD program under § 26.605(a) either during construction activities as defined in § 26.5, or during activities performed under an ML that allows the assembly, testing, or both, of a manufactured reactor. This FFD program implements all the FFD program requirements in § 26.604 plus drug and alcohol testing. The timing element of the proposed applicability statement of § 26.605(a) is equivalent to that for an LWR licensee or other entity who is performing those same activities at a facility licensed under part 50 or 52 and helps provide assurance that those individuals who assemble, test, or perform construction activities as defined in § 26.5 or direct these activities are fit for duty and trustworthy and reliable. This is important because assembly and testing a manufactured reactor and the construction and testing of SSCs required for facility operation require, in part, adherence to procedures, possible implementation of unique and precise assembly techniques, and quality assurance and controls. Additionally, SSCs within a manufactured reactor may not be accessible, testable, or available for quality assurance and verification after the reactor is assembled. This requirement is also proposed to address solo-assembly activities that may cause latent failures and passive SSCs located internal to a reactor (for example, a fusible link designed to melt at a particular temperature to trigger an actuation mechanism) that are relied upon for safe operation but cannot be inspected or tested for proper installation, configuration, or operation after installation. A § 26.605(a) FFD program for these types of activities is equivalent to the FFD program applicable to the assembly of the reactor vessel internals and testing of the SSCs internal to the reactor at an LWR licensed under part 50 or 52.
                        ( print page 86963) Proposed § 26.605(b) would apply to the same licensees and other entities as in proposed § 26.605(a) but before the loading of fuel onsite into a reactor vessel; before receiving a manufactured reactor; or before individuals subject to part 26 operate, test, perform maintenance of, or direct the maintenance or surveillance of security-related equipment or equipment that a risk-informed evaluation process has shown to be significant to public health and safety. These entities must establish, implement, and maintain an FFD program that implements all the requirements in § 26.605(a), except proposed §§ 26.610, “Sanctions”; 26.617, “Recordkeeping and reporting”; and 26.619, “Suitability and fitness determinations”; plus additional requirements due to the increased radiological consequences presented by a part 53 commercial nuclear plant as the licensee readies it for operation. These additional requirements include those in subparts C, D, H, and N of part 26, some of which would replace §§ 26.610, 26.617, and 26.619. Proposed § 26.605(b) would also enable the licensee or other entity to better integrate its facility into the LWR fleet and Category I fuel cycle facilities because subparts C, D, and H of part 26 would be required. These subparts would be required, in part, because it is expected that: (1) individuals will be able to work at any part 50, 52, or 53 commercial nuclear plant and will possess a nuclear safety culture and desirable qualifications, skills, expertise, or services; and (2) licensees and other entities of facilities licensed under parts 50, 52, and 70 may venture to construct or operate a facility licensed under part 53. Therefore, the implementation of these subparts would help ensure that all individuals subject to part 26, except those individuals subject to an FFD program under § 26.604, § 26.605(a), or subpart K of part 26, would be subject to FFD programs that provide reasonable assurance that the individuals are fit for duty, trustworthy, and reliable. Proposed § 26.606, “Written policy and procedures,” would require licensees and other entities to implement and maintain an FFD policy and procedures for their FFD programs. This section would establish requirements equivalent to those in current § 26.403, “Written policy and procedures,” of subpart K. However, a principal difference is that proposed § 26.606 is written to enable the use of urine, oral fluid, and hair for drug testing and screening. Proposed § 26.606(a)(1) would require each licensee and other entity to provide a written FFD policy statement to individuals subject to the FFD program before the individuals are subjected to behavioral observation and any FFD program drug and alcohol test. This would be a protection measure afforded to individuals subject to the FFD program to help ensure that they know what is expected of them before being subject to the FFD program and potential consequences should they violate the FFD policy or procedures. This requirement would also contribute to safety and security because understanding FFD program responsibilities may enhance an individual's safety culture or the individual may self-select out of the licensee's or other entity's hiring process. Proposed § 26.606(a)(2) would require that the FFD policy statement describe the performance objectives in § 26.23, which are the same FFD program performance objectives required for facilities licensed under parts 50, 52, or 70. Having a standard performance outcome based on a licensee or other entity satisfying the § 26.23 performance objectives would enhance consistency in FFD program implementation across all entities subject to part 26. It would also generate confidence that individuals subject to part 26 will safely and competently perform their duties and responsibilities and use NRC-licensed materials in a manner that will protect the public health and safety and common defense and security. Proposed § 26.606(a)(3) would require that the FFD policy statement describe the minimum days off requirements in § 26.205(d)(3) or maximum average work hours requirements in § 26.205(d)(7). Proposed § 26.606(a)(4) would require the FFD policy statement be written in sufficient detail to provide affected individuals with information on what is expected of them and what consequences may result from a lack of adherence to the policy, including those elements described in § 26.603(b), part 26-required sanctions, and required medical/clinical treatment and follow-up testing for FFD policy violations. This requirement is equivalent to § 26.403(a) of subpart K but includes an additional description of what the policy statement must include. For example, the policy would describe the NRC-required sanctions to help deter substance abuse and required medical/clinical treatment and follow-up testing for FFD policy violations. This provision would provide a protection measure by helping the individual get the assistance they need and help ensure that the individual refrains from substance abuse. Proposed § 26.606(a)(5) would require that the FFD policy statement describes the individual's responsibilities to report for work in a physiological and psychological condition that enables the safe and competent performance of assigned duties and responsibilities and inform a licensee- or other entity-designated representative when the individual determines that this cannot be accomplished. Proposed § 26.606(b) would require licensees and other entities implementing a FFD program in accordance with subpart M of part 26 to establish, implement, and maintain written procedures for their FFD programs. This requirement would be equivalent to that in § 26.403(b) of subpart K. Proposed § 26.606(b)(1) would establish requirements for a subpart M of part 26 FFD program in which the licensee or other entity implements a drug and alcohol testing program. This provision would be equivalent to the requirements in current § 26.403(b)(1) of subpart K, but § 26.606(b)(1)(i) through (iv) proposes additional clarity and specificity that licensees and other entities must detail in their procedures to address new testing methods in subpart M of part 26 that are not permitted under the current part 26 framework. Clarity and specificity in procedural instructions would support consistent program implementation, which protects all individuals subject to the program. Proposed § 26.606(b)(1)(iv) would require that if the licensee or other entity elects to use the HHS Guidelines for the conduct of drug testing, the FFD program procedures must include the name of the specific HHS Guideline and revision being implemented by the licensee or other entity and a description of the specific sections in the guideline that are being implemented, including specimen collections, drug testing, laboratory procedures, and evaluation of test results. This requirement would help ensure the following: the validity and accuracy of drug testing because the specimens would be subject to laboratory testing that has been certified by the HHS; protection of worker rights equivalent to the privacy, information, and due process protections afforded to Federal workers under the HHS Guidelines because the HHS Guidelines are used in the Federally mandated drug testing programs; consistency in program implementation because all individuals subject to the FFD program would be subject to the same collection, 
                        ( print page 86964)
                        testing, and evaluation processes; and FFD program effectiveness because the effectiveness of the HHS Guidelines have been verified by HHS's National Laboratory Certification Program (NLCP). Detailed procedures would enhance MRO and FFD program personnel reviews of individual test results because instructions would be provided for, in part, the evaluation of specific test results (
                        *e.g.,*
                         positive, negative, biological markers), the conduct of additional testing for invalid or dilute specimens, and the assessment of subversion attempts (
                        *e.g.,*
                         adulterated or substituted). This would benefit FFD program effectiveness and help prevent misunderstanding of program requirements and processes. Proposed § 26.606(b)(2) would require licensees and other entities to include in their written procedures the immediate and follow-up actions that would be taken, and the procedures that would be used, in certain situations specified in proposed § 26.606(b)(2)(i) through (vi). Proposed § 26.606(b)(2) would be equivalent to the requirements in current § 26.403(b)(2), which provides the same requirement under an FFD program for construction for part 50 or 52 licensees and other entities. This would help ensure the effectiveness of the FFD program and its consistent implementation, because part 53 licensed facilities would be implementing procedures to address the same requirements and with individuals who would understand what is expected of them no matter what part 53 facility they were assigned. The situation specified in proposed § 26.606(b)(2)(i) would arise when individuals subject to the FFD program have been involved in the use, sale, or possession of illegal substances, illegal drugs, or illicit substances. This provision would be equivalent to current § 26.403(b)(2)(i), except that the phrase “illegal drugs” would be replaced with “illegal substances, illegal drugs, or illicit substances.” Illegal substances would include legal substances used in a manner inconsistent with Federal or State law. The situation specified in proposed § 26.606(b)(2)(ii) would arise when individuals who are subject to the FFD program are impaired by any substance or the consumption of alcohol as determined by behavioral observation or a test that measures blood alcohol concentration, as defined in § 26.5. Except for a few differences, this provision would be equivalent to current § 26.403(b)(2)(ii) of subpart K. The NRC would not include the phrases “to excess” and “accurately” in proposed § 26.606(b)(2)(ii). Subpart M of part 26 is a performance-based framework that focuses on impaired human performance, and for alcohol, impairment is determined by behavioral observation or by blood alcohol concentrations exceeding the limits in § 26.103, “Determining a confirmed positive test result for alcohol,” using an evidentiary breath testing (EBT) device for alcohol (not whether an individual drank “to excess”). If impairment is determined by an individual's behavior, it must be based on physiological indications of alcohol impairment. These indications are well established in medical, clinical, and law enforcement organizations, and could be used by the licensee or other entity through its procedures and training.[11] The NRC would include the phrase “illegal substances, illegal drugs, and illicit substances” in proposed § 26.606(b)(2)(ii) based on operating experience and the terminology in current § 26.23(b). There are far more substances that may cause impairment than just drugs, drug metabolites, and alcohol. The phrase “before or while constructing or directing construction of safety- or security-related SSCs” in current § 26.403(b)(2)(ii) would not be included in proposed § 26.606(b)(2)(ii) because proposed § 26.606 would apply during construction, operation, and decommissioning, if applicable. The NRC would include the term “behavioral observation” in proposed § 26.606(b)(2)(ii) because impairment can be visibly or audibly observed in an individual, and individuals subject to subpart M of part 26 would be trained in behavioral observation under proposed § 26.608. The situation specified in proposed § 26.606(b)(2)(iii) would arise when individuals who are subject to an FFD program that includes drug and alcohol testing attempt to subvert the testing process by adulterating or diluting specimens (
                        *in vivo*
                         or 
                        *in vitro*), substituting specimens, or by any other means. Except for one difference, this provision would be equivalent to current § 26.403(b)(2)(iii). The NRC would include the phrase “if drug and alcohol testing is conducted” to address the licensee or other entity who implements § 26.604, which does not require drug and alcohol testing. The purpose underlying this requirement has increased in significance since issuance of the 2008 part 26 final rule because subversion attempts have accounted for about one-third of all FFD policy violations every year since 2016. The situation specified in proposed § 26.606(b)(2)(iv) would arise when individuals, who are subject to an FFD program that includes drug and alcohol testing, refuse to provide a specimen for analysis or refuse to follow instructions provided by FFD program personnel. Except for two differences, this provision would be equivalent to current § 26.403(b)(2)(iv). As with proposed § 26.606(b)(2)(iii), the NRC would include the phrase, “if drug or alcohol testing is conducted,” to account for an FFD program implemented under § 26.604. The NRC would include the phrase “or follow the instructions provided by FFD program personnel” based on an existing requirement in § 26.89(c) that the collector must inform the donor that if the donor refuses to cooperate in the specimen collection process, then such refusal will be considered a refusal to test and sanctions for subverting the testing process will be imposed. The situation specified in proposed § 26.606(b)(2)(v) would arise when individuals who are subject to an FFD program had legal action taken relating to drug or alcohol use. This requirement would be equivalent to current § 26.403(b)(2)(v). The situation specified in proposed § 26.606(b)(2)(vi) would be when individuals subject to an FFD program demonstrated character or actions indicating that the individual cannot be trusted or relied upon to perform those duties and responsibilities or maintain access to NRC-licensed facilities, SNM, or sensitive information. This includes character traits beyond those attributed to drug or alcohol use. This proposal would help ensure that the licensee or other entity will implement an FFD program designed to demonstrate compliance with the § 26.23(c) performance objective that FFD programs must provide “reasonable measures for the early detection of individuals who are not fit to perform the duties that require them to be subject to the FFD program.” An individual who is not trustworthy and reliable is not fit to perform or direct the performance of those duties and responsibilities or be afforded those types of access that make the individual subject to an FFD program. This proposed requirement also would help to align the subpart M of part 26 BOP with the BOP implemented under § 73.56(f) and proposed § 73.120 and the purpose of the IMP as described in § 73.55(b)(9) and proposed 
                        ( print page 86965)
                        § 73.100(b)(9).[12]
                        
                         The demonstrated character and actions of an individual can indicate whether the individual can be trusted and relied upon to safely and competently perform assigned duties and responsibilities or be afforded those types of access making the individual subject to the FFD program. This holds true for any demonstrated adverse character indication or action on- or offsite. The phrase “character or actions” would be used in proposed § 26.606(b)(2)(vi) to focus on observed examples that indicate an individual subject to subpart M of part 26 may not be fit for duty or trustworthy and reliable. Character traits include but are not limited to personality, temperament, honesty, carelessness, apathy, psychosis, and commitment to safety culture. Assessment of an individual's character should consider the potential for changes in these traits when compared to a previous baseline. Actions would include a physical or verbal demonstration of a character trait that could call into question an individual's fitness, trustworthiness, or reliability. For example, the individual does something physically, verbally, or in writing (
                        *e.g.,*
                         falsifying records, driving while impaired, or harming or threatening to harm oneself, others, or property) that compels another individual to conclude that the observed individual cannot be trusted or relied upon. Unlike the background investigation and reviews of “character and reputation” in § 73.56(d)(6) and (k)(1)(v) and proposed § 73.120, which are principally retrospective reviews of an individual and may be based on third-party information (
                        *i.e.,*
                         information from individuals not subject to NRC requirements), the “character or action” focus of proposed § 26.606(b)(2)(vi) would be a present observation of an individual subject to the FFD program and performed by an individual who is also subject to the FFD program. Whether the information would be received from an individual subject to the FFD program or someone who is not subject to the FFD program, the licensee or other entity would need to review this information (
                        *i.e.,*
                         determine if the information and its source are credible) to determine whether the individual should maintain authorization. Proposed § 26.606(b)(3) would require licensees and other entities to address in their procedures the process, including the duties and responsibilities of FFD program personnel, to be followed if an individual's behavior or condition raises an FFD concern. This provision would also require a process to be conducted when credible information is received by the licensee or other entity that the individual is not fit for duty, trustworthy, and reliable. With a few exceptions, proposed § 26.606(b)(3) would be equivalent to current § 26.403(b)(3). Instead of the phrase “while constructing or directing the construction of safety- or security-related SSCs” in current § 26.403(b)(3), the NRC would use “on the NRC-licensed facility” in proposed § 26.606(b)(3) because this provision would apply during commercial nuclear plant construction, operation, and decommissioning, if applicable, in addition to holders of an ML as described in § 26.3(f). The requirement that the roles and responsibilities of FFD program personnel be described was developed from current §§ 26.4(g) and 26.31(b) and operating experience, which has demonstrated that clear job descriptions help ensure that individuals know who is designated by the licensee or other entity to make decisions regarding FFD program implementation and who can be approached when physiological or psychological help is needed. This is principally a protection consideration afforded to individuals subject to the FFD program. The proposed requirement would also include two conditions not found in current § 26.403(b) that would clarify the initiation of the fitness determination process should an individual's behavior or condition raise an FFD concern. The phrase, “impairment from any cause that in any way could adversely affect the individual's ability to safely and competently perform the individual's duties,” would reflect the § 26.23(b) performance objective. The condition, “the receipt of credible information indicating that the individual cannot be trusted or relied on to perform those duties and responsibilities making the individual subject to this part,” would reflect the § 26.23(a) performance objective. In either case, as required by § 26.23(c), the FFD program must provide reasonable measures for the early detection of individuals who are not fit to perform the duties that require them to be subject to the FFD program. Proposed § 26.606(b)(4) would require licensees and other entities to have written procedures that address the operation and oversight of an onsite or offsite collection facility. This requirement would be equivalent to current §§ 26.403(b) and 26.405(e) and is developed from § 26.41(b), which states that each licensee and other entity who is subject to subpart B of part 26, shall ensure that the entire FFD program is audited, which is part of a licensee's or other entity's oversight of the facility, and § 26.87(a), which states that each FFD program must have one or more designated collection sites that have all necessary personnel, materials, equipment, facilities, and supervision to collect specimens for drug testing and to perform alcohol testing. Having procedures for the operation and oversight of the onsite or offsite collection facility would enhance consistency in program implementation, protect individuals subject to testing, and account for the flexibilities afforded in the types of biological specimens than may be collected under an FFD program subject to subpart M of part 26. Section 26.606(b)(4), when used with the PMRP described in § 26.603(d) and the proposed audit requirement in § 26.605(a), would help maintain FFD program effectiveness and prevent subversion attempts at facilities that may not be under the direct day-to-day oversight of FFD program personnel. Proposed § 26.606(b)(5) would require licensees and other entities to have written procedures that address the fatigue management requirements in § 26.202(b), “Procedures,” and either § 26.205(d)(3) or (d)(7). Proposed § 26.606(b)(6) would require licensees and other entities to have written procedures that provide measures to prevent subversion of drug and alcohol tests conducted onsite and offsite. This proposal was developed from § 26.27(c)(1). Proposed § 26.607, “Drug and alcohol testing,” would establish drug and alcohol testing requirements for licensees and other entities implementing proposed § 26.604, at their discretion, and licensees and other entities implementing proposed § 26.605. Except for a few differences, proposed § 26.607 would be equivalent to current § 26.405, which requires licensees and other entities implementing an FFD program under subpart K of part 26 to have a drug and alcohol testing program that demonstrates compliance with the requirements in § 26.405(b) through (g). The differences are commensurate with the risk consequences presented by a part 53-licensed facility as compared to a part 50 or 52 nuclear power plant. These proposed requirements would improve flexibility in the conduct of 
                        ( print page 86966)
                        drug and alcohol testing while maintaining protections afforded to individuals subject to the FFD program. Proposed § 26.607(a) would require licensees and other entities to obtain a split specimen for all drug tests using oral fluid or urine for all test conditions in § 26.607(b), (h) and (j). Neither current subpart K nor current subparts B or E of part 26 require a split specimen. However, the majority of the LWR fleet uses split specimens for drug testing and commercially available drug screening products use a split specimen technique. Since publication of the 2008 part 26 final rule, the HHS has issued guidelines for urine and oral fluid that require split specimens, and the draft proposed HHS Guidelines for hair requires split specimens, as well. The required use of a split specimen process would protect the individual because, upon a donor-alleged discrepant or questionable test result, the donor may provide permission to test the split specimen (specimen B) in an effort to refute the laboratory test results for specimen A. The requirement also would enable the MRO to direct laboratory testing of specimen B if specimen A were invalid; though the NRC expects specimens becoming invalid at the laboratory to be a rare occurrence as testing would be conducted in HHS-certified laboratories with trained collectors. In the event that a specimen is determined to be invalid, then the occurrence would likely warrant further investigation by the MRO and laboratory to identify the cause. This protocol would be equivalent to the special analysis testing in current § 26.163(a)(2) for dilute specimens in that additional laboratory analysis is performed because of a questionable test result. If a split specimen is tested by an HHS-certified laboratory, then the test result from specimen B must be used as part of the determination for an FFD policy violation as required by § 26.185(n), “Evaluating results from a second laboratory.” However, this is not to say that the test results from specimen A should be discarded. Since the HHS-certified laboratory should report all test results from all specimens tested to the MRO, like the information described in § 26.169, “Reporting results,” test result differences between specimens A and B can be used to inform the MRO as to what should be reported to the licensee or other entity to either facilitate medical or clinical assistance for the individual, inform an FFD-policy violation determination, or both. The proposed § 26.607(a) requirement would also state that if the licensee or other entity elects to use a POCTA device for screening during random testing or portal area monitoring (
                        *e.g.,*
                         pre-access screening), a split specimen would not need to be taken. The reason for this exception would be that the requirements in § 26.607(h)(4) establish the process to be implemented when a screening test indicates a presumptive positive, adulterant, or a discrepant biological marker, if applicable. This process includes collecting and testing a specimen for analysis at an HHS-certified laboratory. Proposed § 26.607(b) would require the licensee or other entity to subject individuals identified in § 26.202 to drug and alcohol testing under the five conditions listed in § 26.607(b)(1) through (5). Proposed § 26.607(b) would be equivalent to current § 26.405(c). Proposed § 26.607(b)(1) would require pre-access testing similar to current § 26.405(c)(1), which requires testing before assignment to construct or direct the construction of safety- or security-related SSCs. Unlike current § 26.405(c)(1), the proposed requirement would not include the phrase, “construct or direct the construction of safety- or security-related SSCs,” because, for licensees or other entities under part 53, the pre-access test condition applies to construction, operation, and decommissioning, if applicable, to help inform a licensee's or other entity's authorization determination. The proposal also would use “pre-access” instead of “pre-assignment,” which is used in current § 26.405(c)(1). A pre-access test would require the collection of an oral fluid or a urine specimen no more than 14 days before the individual is granted unescorted access. Although this change has roots in the 2008 part 26 final rule, which reduced the period within which pre-access testing must be performed from 60 days to 30 days or less, the 14-day proposal is based on three lessons learned from operating experience. First, the 14-day period would be a large enough window of time to collect the specimen and evaluate test results because licensees or other entities typically receive laboratory test results within 5 business days of laboratory receipt of the biological specimen. At the same time, the 14-day period would be small enough to help ensure that the test results are representative of the individual's forensic toxicology before being granted authorization. Second, the 14-day window would enable the licensee or other entity to conduct an unannounced pre-access drug and alcohol screening using a hair specimen or a POCTA. This would help prevent an individual from attempting to subvert the drug and alcohol test by temporarily abstaining from drug or alcohol abuse or adulterating or substituting their specimen to obtain a non-positive test result. Third, the NRC does not expect licensees and other entities licensed under part 53 to have the large and periodic influxes of individuals (either licensee employees or C/Vs) that LWRs have to support facility operation, maintenance, engineering design changes, or nuclear refueling. Therefore, these licensees or other entities would not be periodically challenged to in-take a large workforce within the proposed 14-day pre-access testing window. Proposed § 26.607(b)(2) would require the licensee or other entity to conduct random drug and alcohol testing of all individuals subject to the FFD program. With one exception, this proposed requirement would be equivalent to current § 26.405(b). Section 26.405(b) gives licensees and other entities that implement an FFD program subject to subpart K of part 26 the option to impose random drug and alcohol testing. Proposed § 26.607(b)(2) would not offer that option because subpart M of part 26, unlike subpart K, would not allow a licensee or other entity to implement a fitness monitoring program under current § 26.406 instead of a random testing program. The principal reasons for not allowing this flexibility would be that no licensee or other entity has ever implemented a fitness monitoring program (
                        *i.e.,*
                         there is no operating or regulatory experience on which to judge the effectiveness of a fitness monitoring program) and the proposed subpart M framework already uses behavioral observation to help ensure FFD program effectiveness. Supplementing the proposed § 26.609 BOP with an additional observation technique (
                        *i.e.,*
                         the fitness monitoring program) would not result in a level of deterrence or detection equivalent to that which would be obtained through behavioral observation and random drug and alcohol testing. Proposed § 26.607(b)(2)(i) through (v) would provide specific requirements for the conduct of a random testing program. These paragraphs would be equivalent to § 26.405(b)(1) through (4), although with a few differences. The similar provisions would be proposed in § 26.607(b)(2)(i), (b)(2)(iii), and (b)(2)(iv). The differing provisions would include proposed § 26.607(b)(2)(ii), which would refer to an “FFD program procedure” instead of the reference to an “FFD program policy” in § 26.405(b)(2) because procedures 
                        ( print page 86967)
                        contain the instructions that implement FFD program requirements, but the FFD policy need not contain specific instructions. Section 26.607(b)(2)(ii) would also require individuals who are selected for random testing to report to the onsite collection site, as opposed to the collection site in § 26.405(b)(2) because alcohol metabolism necessitates a relatively timely alcohol test. This change is also proposed because the NRC expects that part 53 licensees and other entities may use a combination of onsite (for random, for-cause, and post-event testing) and offsite (for pre-access, post-event, and follow-up testing) collection facilities for drug and alcohol testing and may have to afford reasonable accommodation to certain individuals, which would add complexity in the licensee's or other entity's procedurally determined time period in which an individual must report to the collection facility. Another difference from § 26.405(b) would be proposed § 26.607(b)(2)(v), which would establish the random testing rate for the population of individuals subject to testing. Subpart K of part 26 does not establish a random testing rate. The proposed requirement would be equivalent to current § 26.31(d)(2)(vii), which requires that the sampling process used to select individuals for random testing provides that the number of random tests performed annually is equal to at least 50 percent of the population that is subject to the FFD program. The NRC would revise that slightly for proposed § 26.607(b)(2)(v) to require a 50 percent random testing rate for the licensee employee population and a 50 percent random testing rate for the C/V population. The NRC proposes this change for two reasons. First, although operating experience has demonstrated that § 26.31(d)(2)(vii) helps provide reasonable assurance that individuals are fit for duty and trustworthy and reliable through the detection and deterrence of substance abuse, this same operating experience demonstrates that, on many occasions, the C/V population has been tested at a rate lower than 50 percent, even though this population results in the majority of all FFD policy violations. This bias occurs because C/Vs are available for testing only during short periods of time or periodically throughout the year, whereas licensee employees are essentially always available for a test. A second reason why the NRC is proposing a different 50 percent random testing protocol than in the current part 26 requirements is that the flexibilities afforded to part 53 licensees or other entities in subpart M of part 26 are not afforded to licensees or other entities that must implement an FFD program under subparts A through I, N, and O of part 26. These flexibilities include enabling the use of a POCTA device to screen individuals during the random testing process and the use of offsite collection facilities for pre-access testing. The potential reduction in FFD program effectiveness caused by licensee or other entity implementation of these options would be offset by subpart M requirements that mitigate possible challenges to the FFD program, such as the 50 percent random testing rate for the licensee employee population and 50 percent random testing rate for the C/V population. Proposed § 26.607(b)(3) would require for-cause testing equivalent to that used in current FFD programs implementing § 26.405(c)(2). The NRC would require for-cause testing, like random testing, to be conducted onsite to ensure that the test is conducted as soon as reasonably practicable. This is an important consideration when for-cause testing for alcohol or using oral fluid for drug screening or testing because human metabolism continually lowers the concentrations of the drugs, drug metabolites, and alcohol perhaps to concentrations lower than the initial or confirmatory testing cutoffs. Additionally, for facilities that are sited in geographically remote locations, an offsite collection facility might be too far away or not readily accessible. Proposed § 26.607(b)(4) would require post-event testing in a manner equivalent to current § 26.405(c)(3) with a few adjustments. For part 53 licensees or other entities, the NRC proposes post-event testing under two conditions: events involving human errors that may have caused or contributed to the events (proposed § 26.607(b)(4)(i)), and events not involving human error that result in adverse health consequences or damage to any safety- or security-related SSC (proposed § 26.607(b)(4)(ii)). The word “significant” would not be used in § 26.607(b)(4)(ii)(A) to describe the “illness or personal injury” as used in § 26.405(c)(3)(i) because § 26.607(b)(4)(ii)(A) would describe which illnesses or injuries are covered. Proposed § 26.607(b)(4)(ii)(B), unlike § 26.405(c)(3)(ii), would not use the word “significant” to describe the damage to safety- or security-related SSCs because any damage to safety- or security-related SSCs would require testing within four hours of the event unless immediate medical intervention precludes the conduct of the test on the individual(s) who caused or contributed to the event. Proposed § 26.607(b)(4)(ii)(B) also would not use the word “construction” as in § 26.405(c)(3)(ii) because § 26.607(b)(4) would apply to construction, operation, and decommissioning, if applicable. Proposed § 26.607(b)(4)(i) would require the licensee or other entity to define in its procedures the terms “human error” and “event.” These terms may take on various meanings and they are not defined in the current or proposed rule, so the licensee or other entity would be required to describe or define these terms to help ensure consistent implementation of subpart M of part 26 and that the post-event test condition would be consistently applied to all individuals subject to the FFD program. The § 26.405(c)(3)(i) requirement that “the event is recordable under the Department of Labor standards contained in 29 CFR 1904.7, and subsequent amendments thereto,” would not be carried over to proposed § 26.607(b)(4). Instead, the NRC proposes to prescribe the post-event test conditions in § 26.607(b)(4), in part so they would not change unless the NRC amends the requirement. Proposed § 26.607(b)(5) would require follow-up testing. This requirement would be equivalent to current § 26.405(c)(4), although the proposed § 26.607(b)(5) would further describe follow-up testing. The NRC proposes to describe follow-up testing as part of a series of tests for drugs, alcohol, or both, which are performed after an individual subject to part 26 has violated the FFD policy on substance use or abuse, or the sale, use, or possession of illegal drugs. Follow-up testing would be used to verify an individual's continued abstinence from substance abuse. The NRC would not include a reference to a follow-up plan as in § 26.405(c)(4) because the intent of a follow-up plan is to conduct a series of drug tests, alcohol tests, or both, to verify continuing abstinence from substance abuse. Nevertheless, individuals who violate an FFD policy on substance use or abuse, or the sale, use, or possession of illegal drugs, should have a follow-up plan that includes a definition of “abstinence” from the medical professional prescribing the plan. Proposed § 26.607(c) would provide additional testing requirements. This proposed requirement would be equivalent to § 26.405(d) and would require implementation of select requirements from current subpart E of part 26. The proposed requirements would govern directly observed collections, shy bladder situations, special analysis testing, and alcohol testing. These requirements would be necessary to maintain FFD program 
                        ( print page 86968)
                        effectiveness equivalent to that currently implemented by the LWR fleet. Proposed § 26.607(c)(1) would require validity testing and establish the minimum panel of drugs and drug metabolites to be tested. This panel would be the same as those in §§ 26.31(d)(1) and 26.405(d) because, based on operating experience from LWR FFD program implementation, this panel has been determined to contribute to a licensee or other entity satisfying the FFD performance objectives in § 26.23(a) through (d). Proposed § 26.607(c)(1) would differ from § 26.405(d) because it would require testing of oral fluid and urine specimens for validity, including at least one biological marker (developed from an HHS Guidelines provision) and one adulterant (equivalent to current validity testing for urine specimens in part 26). Section 26.405(d) requires that urine specimens collected for drug testing be subject to validity testing. The addition of oral fluid validity testing is important because, just as there are publicly available kits to subvert a urine drug test, kits that may be used to subvert a drug test that uses oral fluid as a biological specimen are also readily available. Proposed § 26.607(c)(2) would include requirements that already exist in the part 26 framework that provide protections for individuals subject to the FFD program and contribute to testing effectiveness when collecting and assessing a urine specimen. Specifically, current § 26.115, “Collecting a urine specimen under direct observation,” describes the exclusive grounds for performing a directly observed collection and the process to be followed to protect the privacy of the individual. Section 26.119, “Determining `shy' bladder,” establishes the process to be followed when a donor is not able to produce a sufficient amount of urine for testing, and § 26.163(a)(2) requires special analysis testing when a specimen is dilute to help prevent a subversion attempt. Proposed § 26.607(c)(3) would require implementation of all the current alcohol testing requirements in § 26.91, “Acceptable devices for conducting initial and confirmatory tests for alcohol and methods of use,” through § 26.103, “Determining a confirmed positive test result for alcohol.” Using the same alcohol testing framework for parts 50, 52, 70, and 53 licensees and other entities would provide for regulatory consistency, protections for individuals subject to the FFD program (
                        *e.g.,*
                         the quality controls and verification applied to the EBT device), and FFD program effectiveness (
                        *e.g.,*
                         accuracy of test results). For alcohol testing, unlike drug testing, there is a preponderance of evidence that correlates blood alcohol concentrations to impairment and intoxication. Furthermore, FFD performance data has demonstrated that the time-dependent alcohol cutoffs in § 26.103 have increased the detection of individuals who are under the influence of alcohol. For these reasons, the current alcohol requirements in part 26 are proposed for FFD programs under subpart M. Proposed § 26.607(c)(4) would establish additional testing requirements. This proposal would be equivalent to current § 26.405(f) for facilities licensed under part 53 for the conduct of drug testing. Unlike § 26.405(f), proposed § 26.607(c)(4) would not reference validity screening and initial drug and validity tests at licensee testing facilities as this would be required in proposed § 26.607(c)(1). Another minor difference between § 26.405(f) and proposed § 26.607(c)(4) would reflect the requirement in subpart M of part 26 to use an HHS-certified laboratory for all biological specimens collected and not just for urine specimens. Consistent with § 26.405(f), proposed § 26.607(c)(4) would require the use of an HHS-certified laboratory for all test conditions listed in § 26.607(b), MRO-directed tests, and the testing of a split specimen. Further, HHS-certified laboratory test results using urine or oral fluid would be required for the issuance of an FFD policy violation and part 26-required sanction. All drug testing would need to be performed at an HHS-certified laboratory to help ensure FFD program effectiveness and to protect the donor from a false positive test result and an unwarranted FFD policy violation. The donor would be protected because laboratory procedures for specimen accessioning, testing, custody and control, and evaluation of test results and the training and qualification of laboratory personnel are evaluated by HHS as part of the NLCP. This provides assurance that the drug testing results are accurate and attributed to the donor. Urine, oral fluid, and hair specimens may also be screened and tested for drugs and alcohol as described in § 26.607. Drug and alcohol screening results obtained from urine and oral fluid specimens collected and analyzed using a POCTA device and screening results obtained from a hair specimen or a portal monitor may only be used as potentially disqualifying information for a licensee's or other entity's authorization determination (
                        *i.e.,*
                         used to assess the fitness, trustworthiness, and reliability of the individual). These screening results may not be used for the administration of an FFD policy violation and sanction, except as proposed §§ 26.607(i)(3) and 26.610 for subversions, as defined in § 26.5, of the drug and alcohol screening process. There are three phrases or requirements in § 26.405(f) that the NRC does not propose to use in § 26.607(c)(4). The first is the phrase, “consistent with its standards and procedures for certification,” regarding the operation of an HHS-certified laboratory, because the laboratory would not be HHS-certified if it were not following “its standards and procedures for certification.” The second is the requirement that urine specimens that yield positive, adulterated, substituted, or invalid initial validity or drug test results must be subject to confirmatory testing by the HHS-certified laboratory, except for invalid specimens that cannot be tested. This requirement would not be used because, under subpart M of part 26, licensees or other entities would be required to use an HHS-certified laboratory. For a laboratory to be HHS-certified, it must follow the HHS Guidelines and include procedures that describe when a specimen cannot be tested. Lastly, the § 26.405(f) requirement that other specimens that yield positive initial drug test results must be subject to confirmatory testing by a laboratory that demonstrates compliance with stringent quality control requirements that are comparable to those required for certification by the HHS, would not be used because subpart M of part 26 would require the use of an HHS-certified laboratory. Proposed § 26.607(c)(4) would require the licensee or other entity to contract with a primary and backup HHS-certified laboratory. This provision would help ensure that specimens are processed and tested to maintain FFD program effectiveness should the primary laboratory be unable to perform specimen testing. This would help maintain protections afforded to individuals subject to the FFD program (
                        *e.g.,*
                         should the donor or MRO request testing of the split specimen, a different laboratory could be used). This requirement also would state that the primary and backup laboratories must have a different certifying scientist. Having a back-up HHS-certified laboratory and a different certifying scientist would benefit the program and donor because the drug testing instruments, technicians, and certifying scientist would be independent of the 
                        ( print page 86969)
                        primary laboratory testing and review process. The back-up HHS-certified laboratory may be of the same corporate entity as the primary laboratory. Proposed § 26.607(c)(4) would also state that the laboratory would be subject to inspection or audit by the licensee or other entity and that records and documents must be provided and/or able to be photocopied and removed from the premises to support the inspection or audit. This requirement would be equivalent to current § 26.41(d) except that laboratories would not be able to limit the use and dissemination of documents copied or taken from the laboratory by a licensee or other entity. This is necessary to ensure the continuing effectiveness of FFD programs, because NLCP findings and audit results could adversely impact FFD program effectiveness. Pertinent information includes and should not be limited to NLCP-identified weaknesses (
                        *e.g.,*
                         custody and control, accessioning, instrumentation, procedures, training, supervision, review of test results, and resolution of previously identified corrective actions) that may impact the effectiveness of FFD programs. Proposed § 26.607(d) would help protect the donor from mistakes made during the drug and alcohol testing processes and help ensure FFD program effectiveness. The rule would require the licensee or other entity to protect the individual's privacy and the integrity of the specimen and to implement quality controls to ensure that test results are valid and attributable to the correct individual. This requirement would be equivalent to the first sentence of current § 26.405(e), except that the word “stringent” was removed from the phrase “stringent quality controls,” because the word “stringent” is not defined. Proposed § 26.607(e) would describe the requirements for licensees and other entities that use offsite collection facilities. Consistent with current § 26.405(e), a licensee or other entity would be able to conduct specimen collections and alcohol testing at a local hospital or other facility. Unlike § 26.405(e), proposed § 26.607(e) would not restrict licensees and other entities to use hospitals and other facilities that meet the requirements in 49 CFR part 40, “Procedures for Transportation Workplace Drug and Alcohol Testing Programs,” because subpart M of part 26 is intended to provide flexibilities beyond those in the current part 26 framework. Licensees and other entities may use these Department of Transportation requirements to inform their procedures under § 26.606(b)(1) as long as the procedures do not conflict with the requirements in part 26 or the HHS Guidelines. Proposed § 26.607(e) would also require licensees and other entities to audit offsite collection facilities before their use and biennially to confirm that the facility procedures are comparable to those described in subpart E of part 26 or the HHS Guidelines for urine and oral fluid. This proposed requirement is based on current § 26.41(a) and (b). The § 26.607(e) audit requirement would be a program effectiveness consideration because offsite collection facilities may not require vigilance of their collectors (
                        *e.g.,*
                         identification of subversion attempts), diligence in the protection of worker rights (
                        *e.g.,*
                         privacy and specimen custody and control), or procedural compliance. The offsite facility used by a licensee or other entity under proposed § 26.607(e) would have to be licensed to conduct specimen collections and perform alcohol testing, and be audited, by the State or a State-designated entity. This requirement would help provide assurance of adequate collection facility performance and may help reduce the burden on the licensee or other entity and the collection facility. Crediting a State audit (or State licensure, oversight, or regulation) is established in §§ 26.4(i)(4) and (j), 26.91(e)(5), 26.153(f)(1), and 26.183(a). Proposed § 26.607(f) would provide the requirements for initial drug testing. This provision would be equivalent to § 26.405(f) except to account for the alternative biological specimens that may be tested under subpart M of part 26. For the testing of all biological specimens, the licensee or other entity under part 53 would be required to use a device that employs an immunoassay screening technique, or an alternative technology that the licensee or other entity has incorporated into its FFD program through the § 26.603(e) change control process, that demonstrates compliance with the requirements of the U.S. Food and Drug Administration (FDA) for commercial distribution. Examples of alternative technologies include liquid or gas chromatography and mass spectrometry. Licensees and other entities would use the § 26.603(e) change control process to evaluate and document a change to their collection and analysis procedures to enable the use of a better or perhaps more cost-effective collection and/or testing technology. Another difference from § 26.405(f) would be changing the word “urine” in § 26.405(f) to “biological specimens” in § 26.607(f). Lastly, proposed § 26.607(f) would include the phrase “discrepant biological marker” as a drug screening result that must be analyzed by an HHS-certified laboratory and evaluated by the MRO to help inform the MRO's determination of a subversion attempt. Proposed § 26.607(g) would enable a part 53 licensee to use oral fluid as a biological specimen for testing. This requirement would be equivalent to § 26.31(d)(5), which enables the MRO to conduct drug and alcohol testing using alternative methods, and § 26.405, which does not preclude the use of oral fluid specimens for FFD programs that implement subpart K of part 26 requirements. In order to provide assurance that drug testing is effective and protects the worker, § 26.607(g) would require that the licensee's or other entity's procedures incorporate the HHS Guidelines or the requirements in part 26 for the conduct of urine or oral fluid testing. The proposed § 26.607(g) requires that the oral fluid collection device must have received premarket approval from the FDA and must not expire before laboratory testing. Also, the drugs, drug metabolites, initial and confirmatory testing cutoffs, and biological markers, if applicable, must be those established by HHS for oral fluid drug testing and the alcohol cutoffs in part 26. If they are not established by HHS or the NRC for the paneled drugs and drug metabolites, then they would be determined and documented by a forensic toxicologist review. This forensic toxicologist review would help ensure that the device accurately tests for the drug, drug metabolite, biological markers, adulterants, and/or alcohol and that the results from the device are comparable to those established in the HHS Guidelines for oral fluid testing. Proposed § 26.607(h)(1) and (2) would enable the use of a POCTA device during the random and pre-access testing processes. These requirements are adopted from § 26.97, “Collecting oral fluid specimens for alcohol and drug testing,” and § 26.405(f), which does not preclude the use of oral fluid testing. To use a POCTA device for urine, oral fluid, or other biological indicators (breath, sweat, etc.), a forensic toxicology review would be required to ensure that the device is forensically effective. If the POCTA device is forensically effective, then the donor would be reasonably protected from a false positive test result, the licensee or other entity would be reasonably protected from false negative test results, and the FFD program would remain effective. For a POCTA device to be forensically effective, the forensic toxicologist would need to document an evaluation that the performance of the 
                        ( print page 86970)
                        POCTA device must be comparable to the requirements in § 26.161(b) for a urine specimen or the procedures in the HHS Guidelines for urine or oral fluid, as implemented by the licensee or other entity through its procedures. The use of POCTA for oral fluid and urine specimens for the pre-access and random testing processes would be acceptable because individuals in the pre-access process would be subject to an oral fluid or urine specimen collection and possible drug screening using a hair specimen, which are both required to be sent to an HHS-certified laboratory. For random testing, the individual would have also been granted authorization under the AA and FFD requirements and have been subject to behavioral observation and physical protection screening (
                        *e.g.,*
                         verification of identify, and screening for explosives and contraband). Proposed § 26.607(h)(3) would require that procedures be developed that ensure the effectiveness of the POCTA collection process, assessment of the screening results, and prevention of subversion attempts. This requirement would be equivalent to current § 26.403(b)(1) and would help ensure protections afforded to individuals subject to the FFD program and program effectiveness. The subpart M of part 26 framework enables the use of POCTA for random screening of individuals for any part 53 facility, so the licensee or other entity should exercise due diligence and implement risk management strategies to ensure the efficacy of random screening and its contribution to an effective FFD program. Proposed § 26.607(h)(4) would provide that an individual donor who screens positive (or whose specimen is invalid or indicates a discrepant biological marker or adulterant) is removed from all duties and responsibilities making the donor subject to subpart M of part 26. Under proposed § 26.607(h)(4)(i), the donor then would be immediately subject to a drug and alcohol test that provides quantified confirmatory test results from which an FFD policy violation may be issued. Similar to other requirements for specimen collections, except for biological specimens analyzed by a passive detection system, the licensee or other entity would be required to implement procedures that ensure that all specimens collected are uniquely assigned to the donor (
                        *i.e.,*
                         procedures that provide for custody and control of the specimen). If the individual shows signs of impairment during the POCTA process, proposed § 26.607(h)(4)(ii) would require the temporary removal of the individual's authorization until the MRO reviews the laboratory test result(s), and interviews the individual, and a determination of fitness finds that authorization may be restored. Section 26.607(h)(4) is equivalent to § 26.77(b) and was informed by the requirements in §§ 26.419, 26.75(c) and (d), and 26.185(c). Proposed § 26.607(i) would enable the collection of hair specimens for drug testing to supplement pre-access testing that uses urine or oral fluid specimens. Hair testing would be a new feature in the part 26 framework. The NRC proposes to permit the use of hair testing for only Schedule I or II drugs or their metabolites to inform a licensee's or other entity's determination whether the individual is trustworthy and reliable. For example, if an individual stated no prior use of illegal drugs or potentially addictive habits, a hair screening test could be performed during the pre-access process to ascertain the validity of the individual's statement. However, if the HHS-certified laboratory communicates a laboratory-confirmed positive test result, an FFD policy violation may not be administered. This laboratory information must be treated as potentially disqualifying FFD information, unless the individual subverts the screening process, in which case a permanent denial of authorization must be issued under proposed § 26.610. To provide assurance of testing effectiveness and protections afforded to individuals subject to the FFD program, proposed § 26.607(i) would require that an HHS-certified laboratory must be used to analyze the hair specimen, a forensic toxicologist must review the licensee's or other entity's hair screening process, the test kit must be cleared by the FDA, and hair screening must be conducted in accordance with the HHS Guidelines. The forensic toxicologist review would be necessary if the panel of drug or drug metabolites to be tested and their cutoffs are not established by HHS or the NRC for hair. Proposed § 26.607(j) would allow the use of portal area screening for drugs, alcohol, or both. This provision would result in a substantial contribution to a licensee or other entity satisfying the § 26.23 performance objectives by helping ensure that 100 percent of all individuals who arrive at the NRC-licensed facility to perform or direct those duties and responsibilities or maintain those types of access making them subject to the FFD program are fit for duty and deterred from arriving onsite in a physiological condition that may be adverse to safety and security. Additionally, screening could be conducted when an individual exits the NRC-licensed facility to provide assurance that substance abuse had not occurred on the site (see § 26.23(d)). The screening device could be electronically linked to temporarily prevent ingress or egress and could automatically inform licensee- or other entity-designated officials of the portal area alarm. The proposed requirement would enable the licensee or other entity to use innovative technologies to maintain FFD program effectiveness when their PMRP compels the licensee or other entity to implement mitigative strategies to maintain program effectiveness. The use of portal screening technologies may also represent cost savings because, for NRC-licensed facilities that have small staff sizes or are geographically remote, passive drug and alcohol screening technologies could be an innovative alternative to a random testing program, although the license or other entity would need to request and receive an exemption. Proposed § 26.607(j) would also provide that if the portal area screening instrument detects a substance that exceeds the instrument's established setpoint, the individual would be tested with either a collection kit that must be analyzed by an HHS-certified laboratory or a POCTA. This situational screening would be equivalent to a for-cause test. The requirements would not allow an individual to be rescreened by the portal area screening instrument following an initial screening detection that exceeded an established setpoint in order to prevent a subversion attempt. Similar to other drug and alcohol testing technologies enabled for use by subpart M of part 26, a forensic toxicology review would be required before using passive screening technology to help ensure the effectiveness of the instrument by protecting against false positive or negative screening results, which would place an unwarranted burden on the individual, licensee, or other entity. These instruments and alcohol screening devices, already in the marketplace, may also be used to determine true identity to facilitate implementation of the FFD BOP, which may be very practicable at facilities that operate with small staff sizes. Proposed § 26.607(k) would enable the use of a blood specimen for drug, alcohol, or other testing for certain medical conditions as determined by the licensee- or other entity-designated MRO. This requirement would be equivalent to current § 26.31(d)(5). The use of a licensee- or other entity-designated MRO and not one designated by a third party, such as an MRO 
                        ( print page 86971)
                        employed by an offsite specimen collection facility, is important because the MRO must be familiar with the subpart M of part 26 requirements. To help ensure testing effectiveness and protect the worker, the blood test would need to be conducted by a laboratory that demonstrates compliance with quality control requirements that are comparable to those required for certification by the HHS, such as a hospital or clinic certified by the State, Commonwealth, or territory. Proposed § 26.607(l) would require licensee and other entities to use a Federal custody-and-control form (CCF) approved by the OMB for the collection and packaging of a hair, oral fluid, or urine specimen. This proposed requirement is based on the CCF documentation requirements in current subpart E of part 26 because subpart K of part 26 does not require the use of a CCF under § 26.117(e). Additionally, when using a POCTA device, the licensee or other entity would be required to implement a licensee- or other entity-approved and -maintained procedure that ensures the reliability of the tracking, handling, and storage of a specimen from the point of specimen collection to final disposition of the specimen and the reliability of an identification system to uniquely assign the specimen to the donor. Both requirements would help protect the worker by helping ensure chain of custody and by contributing to program effectiveness. Proposed § 26.607(m) would establish requirements for the licensee- or other entity-designated MRO. Section 26.607(m)(1) would be equivalent to § 26.405(g), however, the word “designated” would be added to the first sentence to clarify that the MRO would be designated by the licensee or other entity, and not by a third party. As stated with regard to proposed § 26.607(k), this change would clarify that it is the licensee's or other entity's responsibility, through their designated MRO, to determine whether an individual is fit for duty and trustworthy and reliable. This would be consistent with the description of FFD program personnel in current § 26.31(b) and help provide FFD program effectiveness and protections to individuals subject to the FFD program. The paragraph was also modified from § 26.405(g) to address the determinations of FFD policy violations and fitness required by subpart H for a part 53 licensee or other entity that implements the FFD program described in § 26.605(b). Proposed § 26.607(m)(2) would help ensure that MRO reviews are consistent with those MRO reviews conducted at other NRC-licensed facilities subject to part 26 and that the MRO maintains knowledge of drug collection, testing processes and procedures, and evaluation of testing results. The NRC also proposes that if an MRO performed the duties and responsibilities in §§ 26.185 and 26.187 for at least three continuous years in the last 10 years prior to being hired or contracted by the licensee or other entity, then the MRO would not need to repeat the initial training and examination requirements. The basis for 3 years is that the MRO would have experienced three annual cycles of evaluating drug and alcohol test results, contributed to the FFD annual report to the NRC, experienced a refueling or maintenance outage, understood the duties and responsibilities of individuals subject to the FFD program to make informed determinations of fitness, demonstrated a safety culture that helps ensure FFD program effectiveness, and been subject to NRC inspection. The basis for 10 years is the relatively long periods between significant changes to part 26 and the HHS Guidelines. Proposed § 26.607(m)(3) would require that the MRO attend a medical- or clinical-based training session on a triennial basis. This proposal was developed from Section 13.1 of the HHS Guidelines for urine and oral fluid with two substantial differences: the HHS Guidelines state that “requalification training,” including an exam, must be conducted “at least every 5 years from initial certification,” whereas the proposed § 26.607(m)(3) would require a training session every three years. The proposed requirements are justified because changes in societal drug use or forensic toxicology could occur more frequently than every 5 years, which could compel MROs to attend training in areas of forensic toxicology, determinations of fitness, or other part 26 technical areas on a more frequent periodicity than every 5 years to improve their knowledge and expertise. Proposed § 26.607(m)(4) would require the MRO to evaluate drug testing results by implementing the requirements in § 26.185 or the HHS Guidelines through the licensee's or other entity's procedures. This requirement would help ensure FFD program effectiveness and enhance consistency across the commercial nuclear industry for the evaluation of drug testing results. This also would help protect individuals because they would be subject to the same evaluation criteria. If § 26.185 provides insufficient information for an MRO to make a determination on a drug testing result (including adulterant and discrepant biological markers), the guidance issued by a State agency in the state in which the NRC-licensed facility is located, Federal agency, or nationally recognized MRO training and certification organization may be used to inform an MRO determination. This provision would ensure that the MRO has the flexibility to inform their evaluation of the drug testing results and fitness determination, if necessary, considering the drug- and alcohol-related flexibilities afforded in subpart M of part 26. The proposed requirement would also state that an MRO need not review a confirmed alcohol positive test result determined by an EBT device under § 26.607(c)(3)(vi) and (vii), which are equivalent to the current requirements in §§ 26.101 and 26.103, respectively. The results of an EBT device are precise and accurate enough to support the issuance of an FFD policy violation without an MRO review of an EBT test result if the instrument demonstrates compliance with the requirements in § 26.91. The NRC acknowledges that there are physiological conditions that may cause an abnormally high blood alcohol concentration, such as diabetes, acid reflux, gastroesophageal reflux disease, and perhaps certain diets (high protein and low carbohydrates). However, operating experience has not demonstrated a compelling need to require an MRO review of all EBT test results. For consistency, a licensee or other entity may elect to require its MRO to review all EBT test results when a donor communicates a testing concern or physiological condition. If the donor has a testing concern, the occurrence could be appealed under the proposed § 26.613. If the donor presents a physical condition to the MRO that may have caused an elevated EBT test result, the MRO may direct an alternative testing process (see § 26.607(m)(5)) should it be medically necessary. Proposed § 26.607(m)(5) would require the licensee- or other entity-designated MRO to determine and approve the use of oral fluid or urine as an alternative biological specimen when the donor cannot provide a requested specimen for testing. This proposed requirement is equivalent to § 26.31(d)(5), which enables the use of an alternative specimen collection if a medical condition makes the collection of the biological specimen difficult. This determination and the retest must be completed as soon as reasonably practicable and documented to support recordkeeping, auditing, and NRC inspection.
                        ( print page 86972) Proposed § 26.607(m)(6) would require that the MRO review all specimens screened or tested associated with a drug-related FFD policy violation. This includes POCTA, split specimens, and all specimens taken to resolve a discrepant condition, such as a possible subversion attempt, impairment without a known cause, or a donor-requested or MRO-directed retest. To resolve a discrepant condition, the MRO is authorized to test a specimen for a biological marker, adulterants, or additional drugs. The broad scope of this MRO evaluation would be necessary because of the variety of different screening and testing methods that may have been associated with the FFD policy violation. All information learned from the conduct of part 26 drug and alcohol screening and testing should be used in the evaluation of an individual's trustworthiness and reliability, issuance of a sanction, and development of a follow-up treatment and testing plan, if administered. Proposed § 26.607(n) is equivalent to current § 26.31(d)(6) and would establish limits on the screening and testing of biological specimens. This is a protection consideration afforded to individuals subject to the FFD program and was not provided in subpart K of part 26. This requirement states that specimens collected under NRC regulations may only be designated or approved for screening and testing as described in this part and may not be used to conduct any other analysis or test without the written permission of the donor. Analyses and tests that may not be conducted include, but are not limited to, deoxyribonucleic acid (
                        *i.e.,*
                         DNA) testing, serological typing, or any other medical or genetic test used for diagnostic or specimen identification purposes. The NRC proposes to require that no biological specimens may be passively sampled and analyzed in a manner different than described in subpart M of part 26 to ensure workers are protected from non-consensual passive screening. The subpart M framework enables passive detection of drugs and alcohol, whereas passive detection is not afforded in subparts A through I, N, and O of part 26. Proposed § 26.607(o) is equivalent to current §§ 26.31(b)(1)(iii)(A) and 26.89 and would require that all specimen collections be conducted by a licensee- or other entity-designated and -trained individual. For subpart M of part 26, this would include onsite specimen collections, except a collection by a portal area screening instrument in § 26.607(j). Proposed § 26.608 would require licensees and other entities to provide FFD program training to individuals subject to the FFD program. The proposed performance-based § 26.608 requirement was developed from the prescriptive training requirements in current § 26.29 and modeled on current § 50.120 and the proposed requirements in §§ 53.725 and 53.830 because there is no training requirement in subpart K of part 26. Proposed § 26.608(a)(1) would require an FFD training program that includes the licensee's or other entity's FFD policies and procedures, including fatigue management, and the individuals' FFD program responsibilities. Individuals who collect specimens for testing or screening must also be trained in specimen collector duties and responsibilities, including, at a minimum, specimen collection, custody and control, identification and response to subversion attempts, and privacy. The fatigue management training must include the knowledge and abilities described in § 26.202(c). For individuals specified in § 26.4, a licensee or other entity of a commercial nuclear plant would be required to use a SAT as defined in proposed in § 53.725. These requirements are based on requirements in § 26.29(a)(2), (3), (9), and (10). Proposed § 26.608(a)(2) would require training on the BOP. This requirement would be based on §§ 26.29(a)(8), (9), and (10) and 26.33. The proposal would require individuals to be trained in the detection of behaviors or conditions related to not only illegal drugs, as in the current § 26.33 BOP requirements, but also illicit drugs and substance abuse onsite and offsite. Also, in reference to impairment from fatigue or any cause if left unattended, the phrase in § 26.33, “may constitute a risk to public health and safety or the common defense and security,” would be replaced in § 26.608(a)(2)(iii) with “could result in inattentiveness or human errors,” because subpart M of part 26 is focused, in part, on ensuring individuals are fit for duty to safely and competently perform or direct the performance of assigned duties and responsibilities. Proposed § 26.608(a)(2)(iv) focuses on training to inform individuals that they are responsible for their own conduct, as well as observing others. Specifically, individuals would be trained to recognize when they feel unable to safely and competently perform assigned duties and responsibilities or act in a trustworthy and reliable manner. The proposed training requirement and the proposed reporting requirement in § 26.606(a)(5) are in the interest of safety and security because the individual is proactively announcing that assistance may be necessary. This would be consistent with the performance objectives in § 26.23(b) and (c) where certain behavior or stress conditions may be indicative of an individual not being fit for duty, trustworthy, and reliable. Proposed § 26.608(a)(3) would help ensure that individuals subject to the FFD program understand that FFD policy violations would result in an FFD program sanction and that program information learned or generated by FFD program implementation would be used to aide licensee or other entity authorization determinations and be shared, as requested, with other licensees or other entities subject to parts 26, 53, and 73. This proposed requirement is equivalent to § 26.29(a)(1). Proposed § 26.608(a)(3) would be a protection measure afforded to individuals subject to the FFD program because they would understand that licensees and other entities subject to parts 26, 53, and 73 would be informed of, in part, an individual's character, reputation, and ability to follow policies, procedures, and instructions to safely and competently perform assigned duties and responsibilities in a trustworthy and reliable manner. Fitness-for-duty-related information would include drug and alcohol testing results (not quantitative testing values), issuance of any sanctions, FFD-determinations regarding trustworthiness and reliability, testing programs, treatment, and other remedial or corrective action. Proposed § 26.608(b) would require individuals be trained and receive a trainee assessment before pre-access testing and that refresher training and trainee assessments be conducted periodically thereafter. These requirements would be equivalent to § 26.29(c)(1). However, § 26.608(b) was developed from the SAT-based training requirements in § 50.120 and training elements from the annual training requirements in § 26.29(c)(2). The term “systems approach to training” would have the meaning in proposed § 53.725(c). A trainee assessment would be the same as in currently required SAT-based training programs. Proposed § 26.608(c) would require licensees and other entities to periodically evaluate their FFD training programs and revise them as appropriate. This training focus is not required by subpart K of part 26 or § 26.29 but is proposed to address the flexibilities afforded in subpart M of 
                        ( print page 86973)
                        part 26. This section would be equivalent to § 50.120(b)(3). Proposed § 26.609 would require the implementation of a BOP. The proposed requirement would be equivalent to that in §§ 26.33 and 26.407, “Behavioral observation,” and would apply during construction, operation, and decommissioning, if applicable. Because subpart M of part 26 would apply during decommissioning through a licensee's IMP, proposed § 26.609(a) and (b) were developed, in part, from proposed § 73.100(b)(9) and current §§ 73.55(b)(9) and 73.56(f) to help ensure consistency in the conduct of behavioral observation whether conducted for FFD or security purposes. Under the FFD program, the purpose of the BOP would be to help ensure that individuals subject to the FFD program are fit for duty and trustworthy and reliable to perform or direct those duties and responsibilities and maintain those types of access that make the individual subject to the FFD program. This assurance is accomplished by requiring each individual subject to subpart M of part 26 to be subject to behavioral observation, and by requiring all individuals to perform behavioral observation of others and report FFD concerns to the licensee- or other entity-designated representative(s). The intent of the BOP requirement is not to require that all individuals be observed at all times by others; NRC-licensed operators, maintenance professionals, security officers, and others routinely perform solo operations periodically throughout the day. However, individuals must be subject to observation while they are performing or directing the performance of duties and responsibilities or maintaining the types of access making them subject to the FFD program. Observing behavior only at the beginning of a work shift is not sufficient to ascertain whether an individual is fit for duty, trustworthy, and reliable. Controlled substances may have a delayed effect between use (
                        *e.g.,*
                         ingestion) and the onset of physiological or psychological effects, and fatigue accumulates with time. Behavior must be continually observed throughout the work shift to detect any changes from baseline human performance characteristics, including mental or physical health and mannerisms, or any activities that may indicate that the individual is not trustworthy and reliable. Proposed § 26.609(a) would differ from §§ 26.33 and 26.407 in that it would place the responsibility for performing behavioral observation on “all individuals subject to this subpart,” rather than only those “individuals specified in § 26.4(f) [who] are constructing or directing the construction of safety- or security-related SSCs” in § 26.407 or “individuals who are trained under § 26.29 to detect behaviors” in § 26.33 to improve clarity. Proposed § 26.609(b) would require all individuals subject to the FFD program to report to the licensee- or other entity-designated representative any onsite or offsite behaviors or activities by individuals subject to this part that may constitute an unreasonable risk to the safety or security of the NRC-licensed facility or SNM or may cause harm to others. The NRC proposes this description of reportable conduct because an individual's activities (
                        *e.g.,*
                         use of illegal substances) and communications (
                        *e.g.,*
                         hate speech or threats of violence) offsite are a direct indication of the individual's fitness, trustworthiness, and reliability and must be evaluated as to whether authorization should be granted or maintained. Proposed § 26.609(b) would include a description of this conduct instead of the § 26.33 undefined phrase, “FFD concerns,” to enhance the clarity of the requirement. This proposed BOP reporting requirement would include any information relating to character or reputation of the individual indicating that the individual cannot be trusted or relied upon to perform those duties and responsibilities or maintain access to NRC-licensed facilities, SNM, or sensitive information. This would better align with the proposed § 73.120 BOP requirement, which states that each person subject to behavioral observation must communicate to the licensee or applicant observed behaviors or activities of individuals that may constitute an unreasonable risk to the health and safety of the public and common defense and security. Proposed § 26.609(a) and (b) were written broadly to include offsite conduct that the reporting individual considers serious enough to call into question the character or reputation of the subject individual. Proposed § 26.609(c) would require that licensees and other entities perform behavioral observation visually, in-person, and, when necessary, remotely by live video and audible streaming and capture. This requirement was developed from the security observation requirements in § 73.55(e)(7)(i)(B) and (C), (h)(2)(v), and (i)(2) and (i)(5)(ii). Conducting an in-person observation of another individual is the preferred method to ascertain whether the observed individual can safely and competently perform assigned duties and responsibilities. When in-person observations are not feasible (
                        *e.g.,*
                         during solo operations), the proposed requirement would enable the use of video monitoring. This is addressed, for example, in proposed § 26.609(d) regarding NRC-licensed operator manipulation of reactor controls. Additionally, certain duties (such as maintenance activities performed by a single worker outside of a control room) may not present an opportunity for video monitoring; in these situations, behavioral observation should be conducted on a sampling basis (
                        *i.e.,*
                         a planned observation of the work activity) as outlined in a licensee's or other entity's FFD program. In situations involving small staff sizes, facilities sited in geographically remote locations, or both, additional observers would enhance the effectiveness of a BOP. Technological developments in automated safety and security systems may enable licensees or other entities to reduce staff sizes to 10 to 40 percent of the staff size of an LWR facility licensed under part 50 or 52. Smaller staff sizes may translate into more solo operations, less teamwork, fewer peer checks, or infrequent management oversight of field activities, leading to fewer behavioral observations. Therefore, a licensee or other entity would have fewer opportunities to observe whether individuals are fit for duty. Enabling video and audible streaming and capture to enhance the BOP would be consistent with the security-related behavioral observation requirement in proposed § 73.120(c)(2)(ii), which would also enable video conferencing or other acceptable electronic means promoting face-to-face interaction for those individuals working remotely. Proposed § 26.609(d) would require that licensees or other entities perform behavioral observation of NRC-licensed operators who manipulate the controls of any commercial nuclear plant licensed under part 53, remotely by live video and audible streaming capture for those part 53 facilities where individual task loading does not allow for the effective conduct of behavior observation in addition to assigned operational tasks. The purpose of this paragraph would be similar to that of proposed § 26.609(c), where the possibility of in-person observation is significantly diminished because of solo operations or because the facility may only require a minimum staff size onsite. Proposed § 26.610 would be equivalent to § 26.409, “Sanctions,” and would require the licensee or other entity to establish sanctions for FFD 
                        ( print page 86974)
                        policy violations that, at a minimum, prohibit the individuals specified in § 26.4 from being assigned to perform or direct those duties and responsibilities or maintaining authorization making them subject to subpart M of part 26. To be consistent with § 26.75, “Sanctions,” the severity of the sanction as described in § 26.610 would escalate with the number of occurrences and severity of the FFD policy violation. The sanction would be long enough to help deter future FFD policy violations and facilitate counseling and treatment before the licensee reinstates the individual's access to the facility. The NRC proposes this requirement because the 14-day denial described in § 26.75 may not allow sufficient time for counseling and treatment based on the particular FFD policy violation. Equivalent to § 26.75(c), proposed § 26.610 would also require a minimum 5-year denial of access to the NRC-licensed facility for certain violations of the FFD policy within the protected area of a commercial nuclear plant and by an individual or individuals who are the operators of the conveyance to transport or use formula quantities of strategic SNM. Equivalent to § 26.75(b), proposed § 26.610 would require a permanent denial of authorization be issued for any subversion attempt. Proposed § 26.611 would protect information collected from FFD program implementation and would be equivalent to current § 26.411, “Protection of information.” The protected information would include, but not be limited to, privacy and medical information. Section 26.611 would not include the § 26.411 requirement that FFD programs must maintain and use the personal information with the highest regard for individual privacy because such a requirement would be unnecessary in light of the proposed § 26.611(a) requirement that licensees and other entities must establish and maintain a system of files and procedures to prevent unauthorized disclosure. Proposed § 26.611(b), although equivalent to § 26.411(b), would require licensees and other entities to have all individuals sign a consent to be subject to the FFD program before subjecting the individual to the FFD program (
                        *e.g.,*
                         before being subject to a pre-access test in § 26.607(b)(1), unlike § 26.411(b)). The purpose of this proposal would be to enhance protections afforded to individuals subject to the FFD program and their knowledge of, in part, why they are subject to drug and alcohol testing, behavioral observation, information collection, MRO reviews, and other FFD program elements. Like the consent required by § 26.411(b), the consent would authorize disclosure of the collected information. Consent would not be needed for disclosures to the individuals and entities specified in § 26.37(b)(1) through (b)(6), (b)(8), and persons deciding matters under review in proposed § 26.613, “Appeals process.” Proposed § 26.613 would be equivalent to § 26.413, “Review process.” The proposed title was changed to an appeal process to clarify that § 26.613 would be the process implemented when an individual elects to appeal a licensee or other entity determination that the individual had violated the FFD policy. The proposal would also require that the process include a schedule for the completion of the review of the determination that the individual had violated the FFD policy. The NRC proposes this requirement because operating experience demonstrates that workers may not be protected from a continuous review process that does not result in an outcome. Proposed § 26.615 would require licensees and other entities to perform audits of the FFD program. The proposed section would be equivalent to § 26.415, “Audits.” Under proposed § 26.615(a), audits would be performed at a frequency that ensures the FFD program's continuing effectiveness. This would be particularly important for FFD program elements that are not part of the FFD PMRP required by § 26.603(d). Corrective actions would be taken as soon as reasonably practicable to resolve any problems identified and preclude recurrence. Proposed § 26.615(b) would require the subject matter, scope, and frequency of audits be revised as necessary to improve or maintain program performance based on findings resulting from licensee or other entity implementation of its FFD PMRP. These requirements were developed from appendix B to part 50, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants”; criterion X, “Inspection”; and criterion XVIII, “Audits.” Proposed § 26.615(c) would be equivalent to § 26.415(b) and would enable licensees and other entities to conduct joint audits or accept audits of C/Vs so long as the audit addresses the relevant services of the C/Vs. Proposed § 26.615(d) would be equivalent to § 26.415(c) by establishing requirements for the auditing of HHS-certified laboratories. Unlike § 26.415(c), the proposal would not contain a reference to the Department of Transportation drug and alcohol testing requirements. This would broaden the regulatory flexibility afforded to a licensee or other entity in that they may use an offsite collection or testing facility that does not meet the Department of Transportation requirements. Proposed § 26.615(d) would state that licensees and other entities need not audit an HHS-certified laboratory if the licensee's or other entity's panel of drugs and drug metabolites to be tested is equivalent to the panel by which the laboratory is certified by HHS or is subject to the standards and procedures for drug testing and evaluation used by the laboratory under the HHS Guidelines. The NRC would afford this flexibility because the NRC is aware that HHS desires to streamline changes in its guidelines to its panel of drugs and drug metabolites to be tested. Therefore, if a licensee or other entity elects to implement the HHS Guidelines in its procedures and maintains the minimum panel of drugs and drug metabolites to be tested as required by subpart M of part 26, a licensee or other entity may still use (and not audit) the HHS-certified laboratory because the § 26.603(e) change control process would maintain FFD program effectiveness. To help ensure FFD program effectiveness, § 26.615(d) would also require that collection facility procedures are comparable to those required in subpart E of part 26, including a proposed requirement that the offsite facility's specimen collection and testing procedures are audited on a biennial basis, which is also a protection consideration afforded to individuals subject to the FFD program. Conducting this audit on a biennial basis would be equivalent to that required in § 26.41(b) and would help ensure that the specimen collection process at the facility remains effective. Proposed § 26.617 would establish recordkeeping and reporting requirements equivalent to those in current § 26.417. However, § 26.617 would require retention of records pertaining to administration of the FFD program and FFD performance data required by § 26.717 until license termination, which is based on current § 26.711(a) because § 26.417 does not provide for a retention period. Proposed § 26.617(b)(1) would be identical to the reporting requirements in § 26.417(b)(1) regarding the licensee's or other entity's FFD program. Proposed § 26.617(b)(2) would require the reporting of annual (
                        *i.e.,*
                         January through December) program performance information to the NRC before March 1 of the following year. This reporting would be equivalent to 
                        ( print page 86975)
                        the annual program performance requirement in § 26.417(b)(1), and the March 1 due date is based on the reporting deadline in § 26.717(e). Licensees and other entities would be required to report FFD performance information using new NRC Forms 893, “Single FFD Policy Violation Form,” and 894, “10 CFR part 26, subpart M, Annual Reporting Form for FFD Performance Information.” Proposed § 26.617(c) would require that FFD-related information be shared within the commercial nuclear industry when requested to support authorization determinations. This requirement would help individuals seeking employment by another NRC-licensed facility subject to subpart C of part 26, complete their NRC-required sanctions and licensee-administered or -directed drug and/or alcohol abuse treatment plans before the restoration of authorization by a licensee or other entity. Information sharing may also enhance FFD program effectiveness because FFD-related lessons learned from, for example, substance testing, subversion attempts, and laboratory and MRO performance must be shared when requested. Proposed § 26.619 would require licensees or other entities to establish a process to evaluate individuals when their fitness or trustworthiness and reliability are in question. Section 26.619 would be equivalent to § 26.419, “Suitability and fitness determinations,” but, unlike § 26.419, would apply during the construction and operation phases. Also, proposed § 26.619 would require that a suitability or fitness determination conducted for cause be conducted face-to-face. This proposed requirement is based on current § 26.189(c); however, unlike § 26.189(c), proposed § 26.619 would not prohibit augmenting determinations via electronic means of communication. Instead, § 26.619 would explicitly permit determinations to be performed via electronic means, so long as those determinations are supported by an appropriately trained individual who is present in-person with the individual being assessed. In considering the current restriction on the use of electronic means of communication for determinations of fitness conducted for cause, the NRC finds that since publication of the 2008 part 26 final rule, there have been developments in using electronic means of communication (
                        *i.e.,*
                         “videoconferencing”) as an alternative to conducting face-to-face interactions. To address these considerations, the NRC contracted the Pacific Northwest National Laboratory (PNNL), DOE, to study whether a medical and mental health assessment via electronic communication could be an acceptable alternative to an in-person, face-to-face assessment.[13]
                        
                         Based on this study, if electronic means were to be used to conduct a face-to-face assessment, an in-person element would still be integral to the assessment process. However, under certain circumstances, face-to-face determinations and assessments conducted as part of an FFD program for an entity licensed under part 53 (
                        *i.e.,*
                         those determinations and assessments performed in accordance with § 26.619, § 26.207, or § 26.211) may be augmented via electronic communications. Such remotely conducted determinations and assessments would be required to be conducted with someone who is present in-person with the individual being assessed and who is trained in accordance with the requirements of either § 26.29 and § 26.203(c) or § 26.608 and § 26.202(c). Permitting the use of electronic communications would help ensure FFD program effectiveness, especially in instances where the part 53 commercial nuclear plant is sited in a geographically remote location or when the facility has a small staff size. D. Proposed Changes to Part 26, Subpart N Proposed § 26.709 would make the recordkeeping and reporting requirements in subpart N of part 26 applicable to licensees and other entities of facilities licensed under part 53 that elect not to implement the requirements in subpart M of part 26 or elect to implement the requirements in § 26.605(b). Proposed § 26.711(c) and (d) would be amended to make these requirements applicable to licensees or other entities described in § 26.3(f). Section 26.711(c) provides protection to individuals subject to part 26 by enabling an individual's right to review FFD-related information and correct any inaccurate or incomplete information. Section 26.711(d) requires, in part, that any FFD-related information shared with other licensees or other entities is correct and complete. E. Proposed Changes to Part 26, Subpart O The vast majority of the proposed changes to part 26 would be new or revised substantive provisions that would establish a regulatory obligation or prohibition or would be conforming edits to reflect the addition of part 53. The only new provision that would not be substantive, such that violation of it would not result in a criminal penalty, would be proposed § 26.601. Therefore, the NRC proposes to add § 26.601 to the list of regulations in § 26.825(b) to which criminal sanctions do not apply. 10 CFR Part 50 A. Section 50.160: Emergency Preparedness for Small Modular Reactors, Non-Light-Water Reactors, and Non-Power Production or Utilization Facilities This proposed rule would revise § 50.160(b)(3) and (c)(2) to make that section applicable to applicants and licensees under part 53. Section 50.160 provides an alternative to other part 50 emergency preparedness requirements focused on large light-water reactors to provide an optional emergency preparedness framework specifically for small modular reactors (SMRs) and other new technologies. These alternative emergency preparedness requirements adopt a performance-based, technology-inclusive, risk-informed, and consequence-oriented approach. Commercial nuclear reactor applicants complying with § 50.160 would be required to submit as part of the application the analysis used to determine whether the criteria in § 53.1109(g)(2)(i)(A) and (B) are met and, if they are met, the size of the plume exposure pathway emergency planning zone (EPZ). An EPZ bounds the area surrounding a facility within which detailed planning is needed to implement predetermined, prompt protective actions. The criterion in proposed § 53.1109(g)(2)(i)(A) is that public dose, as defined in § 20.1003, is projected to exceed 10 mSv (1 rem) TEDE over 96 hours from the release of radioactive materials from the facility considering accident likelihood and source term, timing of the accident sequence, and meteorology. The criterion in proposed § 53.1109(g)(2)(i)(B) is that pre-determined, prompt protective measures are necessary. These are the same criteria that are in § 50.33(g)(2)(i)(A) and (B) and are used to assess the need for and size of an EPZ in applications under parts 50 and 52. Applicants choosing to comply with § 50.160 must determine the radiological releases from the facility that are evaluated in the determination of the plume exposure pathway EPZ. Consistent with other Federal guidelines such as the Federal Emergency Management Agency “Radiological Emergency Preparedness Program Manual,” issued in 2023, and the 
                        ( print page 86976)
                        Environmental Protection Agency “PAG Manual: Protective Action Guides and Planning Guidance for Radiological Incidents,” issued in 2017, applicants should consider quantitative and qualitative information on the potential radiological releases that make up the spectrum of accidents used to develop the basis for the applicant's site-specific EPZ. This information is derived from the licensing basis. The NRC plans to update the risk-informed approach in RG 1.242 for part 53 while maintaining its flexibility for using information already developed and available in licensing basis documents, including PRA results, deterministic dose quantities, accident timing, target set analyses, mitigation capabilities, and site-specific factors such as meteorology. In its safety analysis report, the applicant would describe the LBEs relevant to the facility and would consider these LBEs as candidates for the spectrum of accidents used to develop the site-specific EPZ. The LBEs assessed include a wide range of events that are appropriate for considering in the facility's emergency preparedness and response planning. In addition, § 50.160(b)(1)(iv)(A)(
                        *2*) requires licensees to be capable of implementing their approved emergency response plan in conjunction with their safeguards contingency plan. Radiological sabotage events are typically factored into EPZ determinations by considering consequences to be bounded by LBEs and by crediting protection against the DBT in reducing the likelihood of a release. The provisions in proposed § 53.860(a) provide an alternative to applicants and licensees by not requiring them to protect against the DBT of radiological sabotage in accordance with §§ 73.55 and 73.100 if they can demonstrate that the consequences from unmitigated radiological sabotage events are below the safety criteria in proposed § 53.210. The deployment of some commercial nuclear plants under part 53 may involve new scenarios where the source terms and consequences of sabotage-related events are not bounded by the consequences of the unlikely and very unlikely event sequences analyzed under subpart C. Accordingly, the NRC plans to develop guidance for part 53 applicants and licensees choosing to comply with the alternative emergency preparedness requirements in § 50.160 to address this new class of reactors. In Section VI of this document, the NRC is asking for stakeholder feedback on the clarity of the regulations and guidance for various scenarios that might arise in implementing graded approaches for security and emergency planning for some commercial nuclear plant designs. B. Appendix B to Part 50: Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants Appendix B to part 50 would be amended to make it applicable to applicants and licensees under part 53. This results in the need for some revisions to recognize differences in terminology between parts 50 and 53. Namely, the term “design bases,” which is defined in § 50.2, is not used in part 53. For this reason, text is added in both Section III, “Design Control,” and Section IV, “Procurement Document Control,” to refer to “functional design criteria, as defined in § 53.020,” as the part 53 equivalent of the term “design bases.” 10 CFR Part 73 A. Section 73.100: Technology-Inclusive Requirements for Physical Protection of Licensed Activities at Commercial Nuclear Plants Against Radiological Sabotage Proposed § 73.100 would provide a performance-based regulatory framework for the design, implementation, and maintenance of a physical protection program and security organization for certain commercial nuclear plants licensed under part 53. The current § 73.55 physical security requirements for nuclear power reactors licensed under part 50 and part 52 use a combination of performance criteria (
                        *e.g.,*
                         § 73.55(b)(1) through (3)) and numerous prescriptive requirements developed to achieve performance objectives (
                        *e.g.,*
                         § 73.55(k)(5)(ii)). By contrast, in the proposed performance-based approach to physical security for part 53, performance objectives and requirements would be the primary bases for regulatory decision-making, giving the licensee the flexibility to determine how to demonstrate compliance with the established performance criteria for an effective physical protection program. This proposed physical protection program would provide an optional pathway for licensees that elect not to demonstrate compliance with the provisions in § 73.55 and do not satisfy the criterion as described in proposed § 53.860(a)(2). This proposed physical protection program would provide that activities involving SNM are not inimical to the common defense and security and do not constitute an unreasonable risk to the public health and safety. Section 73.100(a) would require each part 53 licensee that elects to demonstrate compliance with this section rather than § 73.55 to implement the requirements therein through its physical security plan, training and qualification plan, safeguards contingency plan, and cybersecurity plan (referred to collectively hereafter as “security plans”) prior to initial fuel load into the reactor (or, for a fueled manufactured reactor, before initiating the physical removal of any one of the independent physical mechanisms to prevent criticality required under § 53.620(d)(1)). The security plans would need to identify, describe, and account for site-specific conditions that affect the licensee's capability to satisfy the requirements of § 73.100. Based on experience from recent new reactor licensing reviews, the NRC recognizes that licensees may seek to receive unirradiated fuel onsite before carrying out the security requirements in § 73.100. However, these security requirements would have to be implemented at some point before reactor operation to address the increased risk arising from irradiated fuel onsite. This proposed rule would make clear that part 53 applicants and licensees using § 73.100 may bring unirradiated nuclear fuel onsite and protect it in accordance with the NRC's requirements for physical protection of SNM of moderate and low strategic significance under § 73.67 until initial fuel load into the reactor (or, for a fueled manufactured reactor, until initiating the physical removal of any one of the independent physical mechanisms to prevent criticality required under § 53.620(d)(1)). Section 73.100(b) would outline the general performance objective and design requirements of the licensee physical protection program. A licensee's program would be required to provide protection against any deliberate act within the DBT of radiological sabotage, including spent fuel sabotage, which could directly or indirectly endanger the public health and safety by exposure to radiation. The physical protection program is supported by the AA program, cybersecurity program, and IMP to demonstrate compliance with the general performance objective of § 73.100(b). Section 73.100(b)(2) was developed, in part, from § 73.55(b)(3). To satisfy the general performance objective of § 73.100(b)(1), the physical protection program would need to protect against the DBT of radiological sabotage. The existing fleet of LWR satisfies this objective by preventing significant core 
                        ( print page 86977)
                        damage and spent fuel sabotage. Some non-LWR reactor licensees' physical protection programs may be designed to prevent a significant release of radionuclides from any source. Therefore, the proposed performance objective would focus on radiological sabotage in general, rather than a specific focus on core damage or spent fuel sabotage, to be technology inclusive and allow for flexibility for different reactor technologies. Under the proposed § 73.100(b)(2)(ii), licensees must provide defense in depth in achieving performance requirements through the integration of engineered systems, administrative controls, and management measures. This requirement would apply defense-in-depth concepts as part of the physical protection program to ensure the capability to demonstrate compliance with the performance objective of the proposed § 73.100(b)(1) is maintained in the changing threat environment. The defense-in-depth philosophy applies to measures against intentional acts as required by § 73.100(b), and the designs of physical security systems should employ defense in depth through systems diversity, independence, and separation under § 73.100(b)(2). The most common defense-in-depth measures apply concepts of redundancy, diversity, independence, and safety margin to ensure systems reliability and availability. The defense-in-depth philosophy applies to the design of a physical protection program, which integrates engineered controls and administrative controls, to provide protection against the DBT for radiological sabotage. Section 73.100(b)(3) would require the physical protection program to be designed and implemented to achieve and maintain the reliability and availability of SSCs required for demonstrating compliance with specified performance requirements. These physical protection performance requirements were informed by § 73.55(b) and the Commission's Advanced Reactor Policy Statement. The performance objective of protecting against the DBT of radiological sabotage is achieved by the design and implementation of the physical protection program, maintained at all times, with the following required performance capabilities proposed in the provisions in § 73.100(b)(3): intrusion detection, intrusion assessment, security communication, security response, protecting against land and waterborne vehicle bomb assaults, and access control portals. The physical protection program must maintain the reliability and availability of SSCs relied upon for demonstrating compliance with the performance requirements. The terms “reliability and availability” are intended to describe defense in depth in a performance-based manner and would be critical elements for demonstrating compliance with the proposed requirement for protection against the DBT of radiological sabotage as described in the proposed § 73.100(b)(2). The first element, “intrusion detection,” would be provided through the use of detection equipment, patrols, access controls, and other program elements and would provide notification to the licensee that a potential threat is present and where the threat is located. The second element, “intrusion assessment,” would provide a mechanism through which the licensee would identify the nature of the threat detected. This would be accomplished through the use of video equipment, patrols, and other program elements that would provide the licensee with timely information about the threat for use in determining how to respond. The third element, “security communication,” would provide a mechanism through which the licensee would communicate the necessary information to the response force to ensure effectiveness of the physical protection program. This would be accomplished through the redundant, independent, and diverse design of physical security and/or plant SSCs relied on for onsite and offsite security communications. The continuity and integrity of communications should account for the DBT's ability to affect the reliability and availability of communications. The fourth element, “security response,” would provide a mechanism through which the licensee would be capable of timely security response to interdict and neutralize threats up to and including the DBT of radiological sabotage. The security response may include the use of onsite armed responders, law enforcement responders (local, State, or Federal), or other offsite armed responders (
                        *e.g.,*
                         licensee proprietary or contract security personnel who are positioned offsite), or a combination thereof, as appropriate.[14]
                        
                         The licensee must provide protection against any element of the DBT, to include those that do not rise to the full capability of the DBT. Structures, systems, and components relied on to provide delay functions must be designed to provide for timely response to adversary attacks with adequate defense in depth. Delay would allow the licensee to take necessary actions to counter any attempt by the threat to advance towards the protected target or target set element. The overall response objective would be to place the threat in a condition from which the threat no longer has the potential for, or capability of, doing harm to the protected target. The fifth element, “protecting against land and waterborne vehicle bomb assaults,” would provide a mechanism through which the licensee would be capable of protecting the plant against the DBT vehicle bomb assault. The methods that are relied on to protect against a DBT land vehicle and waterborne vehicle bomb assault must be designed to protect the reactor building, structures containing safety or security related systems, and components from explosive effects. The sixth element, “access control portals,” would provide a mechanism through which the licensee would be capable of detecting and denying unauthorized access to persons and pass-through of contraband materials (
                        *e.g.,*
                         weapons, incendiaries, explosives) to protected areas. Integrity of the access control system is maintained through licensee oversight and ensures that attempts to circumvent or bypass the established process will be detected and access denied. The proposed performance requirements would permit the applicant or licensee to determine how to design the physical protection program to protect the plant against the DBT of radiological sabotage without 
                        ( print page 86978)
                        prescriptive requirements such as those currently found in § 73.55. DG-5076, “Guidance for Technology Inclusive Requirements for Physical Protection of Licensed Activities at Commercial Nuclear Plants,” has been developed by the NRC to describe one acceptable approach to demonstrate compliance with requirements proposed in § 73.100. Section 73.100(b)(4) would require the licensee to identify target sets in accordance with § 73.55(f). For non-LWR and SMRs, target sets would be defined in DG-5071, “Target Set Identification and Development for Nuclear Power Plants,” as the minimum combination of equipment, operator actions, and/or structures that, if all are prevented from performing their intended safety function or prevented from being accomplished, barring extraordinary actions by plant operations, would likely result in a significant release of radionuclides from any source (
                        *e.g.,*
                         a release to the environment exceeding that analyzed in the DBA licensing basis). Section 73.100(b)(5) would require that each licensee perform a site-specific analysis for the purpose of identifying and analyzing site-specific conditions that affect the design of the onsite physical protection program. Section 73.100(b)(6) would require licensees to implement a performance evaluation program, which would ensure that a licensee will periodically test and evaluate the effectiveness of the physical protection program to protect against the DBT. This program would ensure that licensees are able to demonstrate that the physical protection program satisfies the response requirements of § 73.100 and that the site's protective strategy effectively protects against the DBT. Licensee performance evaluations would include methods to assess, test, and challenge the integration of the physical protection programs functions and demonstrate the effectiveness of security plans, licensee protective strategy, and implementing procedures in accordance with § 73.100(g). Section 73.100(b)(7) would require licensees to implement an AA program in accordance with § 73.56. Section 73.100(b)(8) would require licensees to establish, maintain, and implement protection against a cyberattack based on either the proposed cybersecurity program described in § 73.110 or the program described in existing § 73.54. Section 73.100(b)(9) would require an IMP that monitors the initial and continuing trustworthiness and reliability of individuals granted or retaining unescorted access or unescorted AA to a protected or vital area. The IMP must also implement defense-in-depth methodologies to minimize the potential for an insider (active, passive, or both) to adversely affect the licensee's capability to protect against radiological sabotage. Because no one element of the AA program, FFD program, cybersecurity program, or physical protection program, would, by itself, provide the level of protection against the insider necessary to demonstrate compliance with the performance objective of the proposed § 73.100(b), the effective integration of these programs is a necessary requirement to achieve defense in depth against the potential insider. Section 73.100(b)(10) would require that the licensee have the capability to track, trend, correct, and prevent recurrence of failures and deficiencies in the implementation of the requirements of this section. Section 73.100(b)(11) would require the coordination of the security plans and associated procedures with other onsite plans to manage the safety and security interface during normal or emergency operations. Section 73.100(c) was developed from § 73.55(c)(7), “Security implementing procedures,” and § 73.55(d), “Security organization,” and would outline the requirements for the composition, equipping, and training of the security organization. The purpose of the security organization is to effectively implement the physical protection program. Individuals assigned to perform physical protection or contingency response duties must be trained, equipped, and qualified to perform assigned duties and responsibilities. Section 73.100(d) would establish a performance requirement for searches of personnel, vehicles, and materials for the protection against radiological sabotage. The requirement describes broad categories of material (explosives, firearms, incendiary devices, etc.) to be detected and prevented from entry into the protected area; specific items that will be prohibited would not be prescribed in the regulation but will be stated in the licensee security plans with detailed descriptions being identified in implementation procedures. Section 73.100(e) would require a training and qualification program, described in the training and qualification plan, that ensures personnel are able to effectively perform their assigned security-related job duties. This high-level requirement would allow flexibility in how the licensee chooses to train its security personnel. One method for accomplishing this requirement would be to provide a training and qualification program that would be equivalent to appendix B to part 73. Section 73.100(f) would require periodic security reviews of the physical protection program to ensure effective implementation of the program by independent individuals. The evaluation process would provide a systematized approach for assessing the physical protection program as a basis for further development and improvement. Program reviews should be designed to ensure that the physical protection program maintains effectiveness and demonstrates compliance with NRC requirements. Section 73.100(f)(1) was developed from § 73.55(m) and would require review of each element of the physical protection program. Section 73.100(f)(2) would require licensees to perform self-assessments of physical protection program functions to ensure that the capability to detect, assess, interdict, and neutralize the DBT of radiological sabotage is maintained. Section 73.100(f)(3) would require an audit of the effectiveness of the physical protection program; security plans; implementing procedures; cybersecurity programs; management of the safety/security interface activities; the testing, maintenance, and calibration program; and response commitments by local, State, and Federal law enforcement authorities. Section 73.100(f)(4) would require that results and recommendations, management findings, and any actions taken be documented and maintained to be available for inspection by the NRC. These reviews are independent of the ongoing performance evaluations described in § 73.100(b)(6) and (g). Section 73.100(g) would require that licensee performance evaluations, described in § 73.100(b)(6), include methods appropriate and necessary to assess, test, and challenge the integration of the physical protection program's functions to protect against the DBT. The performance evaluations must also address the licensee's measures to protect against cyberattacks, in accordance with the required cybersecurity plan, and engineered systems designed to protect against the DBT standalone ground vehicle bomb attack. Section 73.100(h) would establish performance requirements for maintaining security SSCs relied on to perform security functions to protect against the DBT. It would require that corrective actions and compensatory measures be taken by a licensee in response to a degradation of security 
                        ( print page 86979)
                        equipment or failure of the equipment to perform its intended functions. The licensee would be required to maintain the SSCs described in its design and licensing basis to ensure that they are reliable and available. Section 73.100(i) would establish requirements for the suspension of security measures in response to emergency and extraordinary conditions. The requirements of this paragraph, which were developed from § 73.55(p), would be intended to provide flexibility to a licensee for taking reasonable actions that depart from a security plan in an emergency when such actions are immediately needed to protect the public health and safety and no action consistent with license conditions and TS that can provide adequate or equivalent protection is immediately apparent in accordance with proposed § 53.740(h). Section 73.100(j) would establish requirements regarding the inspection, retention and maintenance of records required to be kept by the NRC regulations, orders, or license conditions. These proposed requirements are developed from § 73.55(q). B. Section 73.110: Technology-Inclusive Requirements for Protection of Digital Computer and Communication Systems and Networks Section 53.860 would require that a licensee establish, implement, and maintain a cybersecurity program in accordance with § 73.54 or § 73.110. Section 73.110 would establish requirements for the development and maintenance of a cybersecurity program for commercial nuclear plants licensed under part 53. This proposed section would implement a graded approach to determine the level of cybersecurity protection required for digital computers, communication systems, and networks. The proposed new section is informed by: (1) the operating experience from power reactors and fuel cycle facilities; and (2) the existing § 73.54 framework, which addresses some of the basic issues for cybersecurity regardless of the type of reactor. Differences between the § 73.54 requirements and those proposed in § 73.110 are primarily based on the implementation of a consequence-based approach to cybersecurity that provides flexibility to accommodate the wide range of reactor technologies to be assessed by the NRC. A graded approach based on consequences is intended to account for the differing risk levels among reactor technologies. Specifically, the proposed new section would require licensees to demonstrate protection against cyberattacks in a manner that is commensurate with the potential consequences from those attacks. Under proposed § 73.110(a), licensees would need to ensure that digital computer and communications systems are adequately protected against a potential cyberattack that would result in: (1) a scenario where the cyberattack leads to offsite radiation doses that would endanger public health and safety (
                        *i.e.,*
                         the resulting consequence exceeds the reference dose values in § 53.210); or (2) a scenario where the cyberattack adversely impacts the physical security digital assets used by the licensee to prevent unauthorized removal of material or radiological sabotage. Security digital assets would include those used for nuclear MC&A. The proposed § 73.110(b) would require licensees to protect the communication system and networks associated with the functions described in § 73.110(a)(1) and (a)(2) from cyberattacks. To accomplish this, the licensee would establish, implement, and maintain a cybersecurity program for protecting digital assets within the scope of § 73.110 that would make use of risk insights, including threat information, and would consider the resulting level of consequences of the threats. If the outcome of the assessment by the licensee under § 73.110(b)(1) revealed that a potential cyberattack would not compromise any digital assets that support safety and security functions, and thus would not result in the consequences listed in § 73.110(a) (
                        *e.g.,*
                         would not exceed the reference dose values), then only a narrow set of the cybersecurity program requirements in § 73.110(d) and (e) would apply. For example, the licensee would only need to develop a cybersecurity program that implements the requirements dealing with: Analyzing modifications of any asset before implementation to see if they demonstrate compliance with the potential consequences in § 73.110(a);
                    Ensuring employees and contractors are aware of cybersecurity requirements and have some level of cybersecurity training;
                    Evaluating and managing cybersecurity risks to the plant;
                    Reviewing the cybersecurity plan for any required changes; and,
                    Retaining records of the cybersecurity plan along with any plan changes. Section 73.110(c) through (e) were developed from § 73.54(a)(2), and (c) through (h), respectively. The proposed requirements would address the need for the licensee to develop a cybersecurity program that implements a defense-in-depth protective strategy as required by proposed section § 73.110(d)(2). A defense-in-depth protective strategy for cybersecurity is represented by collections of complementary and redundant security controls that establish multiple layers of protection to safeguard critical digital assets. Under a defense-in-depth protective strategy, the failure of a single protective strategy or security control should not result in the compromise of safety and security functions. C. Section 73.120: Access Authorization Program for Commercial Nuclear Plants Section 73.120 would address AA for certain commercial nuclear plants licensed under part 53. The proposed language in § 73.120 would provide an alternate approach to the existing framework for AA under §§ 73.55, 73.56, and 73.57, commensurate with risk and consequences to public health and safety. It would be available to part 53 applicants and licensees who demonstrate in an analysis that the offsite consequences of a DBE satisfy the criterion defined in § 53.860(a)(2)(i) (
                        *i.e.,*
                         would not exceed the offsite dose values in § 53.210(b)). The proposed requirements in § 73.120 would be similar to the existing AA program elements for those NRC licensed facilities issued additional security measures (ASMs) orders and for materials licensees under § 37.21. Applicants not satisfying the criterion would need to establish, implement, and maintain a full AA program, including an IMP, in accordance with § 73.56. Proposed § 73.120(a) would be based on an applicant satisfying the eligibility criterion in § 53.860(a)(2)(i). Section 73.120(b) would identify the categories of individuals who would be subject to an AA program in accordance with this section. The applicability statement in § 73.120(b)(1)(i) would encompass individuals whom the licensee intends to grant unescorted access to the facilities' most sensitive areas, consistent with § 73.56(b)(1)(i) for power reactors and the ASM orders and license conditions issued to any NRC licensed facility or material licensee. Sections 73.120(b)(1)(ii) through (iv) would be consistent with § 73.56(b)(1)(ii) through (iv), respectively. The program would include individuals who may be onsite or offsite (
                        *e.g.,*
                         remote operators or information technology staff) and have virtual access to important plant operational and communication systems based upon assigned duties and 
                        ( print page 86980)
                        responsibilities. An individual who has remote access to plant equipment and communication systems may have trusted privileges greater than the personnel at the plant site. Section 73.120(b)(1)(iii) would state that offsite law enforcement personnel on official duty would not be subject to the licensee AA program. Section 73.120(c) would provide general performance objectives and requirements largely consistent with the AA program requirements for nuclear power reactors under § 73.56 and would provide licensees and applicants the flexibility in establishing their AA program to demonstrate compliance with various performance objectives. Section 73.120(c)(1) would include background investigation requirements consistent with § 37.25, as well as ASMs and license conditions that are applied to non-power reactor licensees. Background investigations include important elements to establish the trustworthiness and reliability of an individual, such that they do not constitute an unreasonable risk to public health and safety or the common defense and security. These include the following: (1) personal history disclosure, (2) verification of true identity, (3) employment history evaluation, (4) unemployment/military service/education, (5) credit history evaluation, (6) character and reputation evaluation, and (7) Federal Bureau of Investigation criminal history record check. Section § 73.120(c)(2) would establish behavioral observation requirements, which are an awareness initiative for recognizing behaviors adverse to the safe operation and security of the facility through observing the behavior of others in the workplace and reporting aberrant behavior or changes in behavior that might reflect negatively on an individual's trustworthiness or reliability. Maintaining behavioral observation would assist and/or improve worker safety and reduce the risk of an insider threat. This proposed requirement in § 73.120(c)(2) would be a scaled version of the full BOP required under § 73.56(f). Section § 73.120(c)(2) would provide licensees greater flexibility to implement behavioral observation options for individuals granted unescorted access to the commercial nuclear plant's protected area. Such options on reporting questionable behavior may include a program similar to the Department of Homeland Security's program, “If you see something, say something,” or to a corporate behavioral awareness program. Commensurate with the potential lower safety and security risks of a commercial nuclear plant that meets the criterion in § 53.860(a)(2)(i), § 73.120(c)(2) would not require the establishment of a comprehensive training program for behavioral observation (
                        *i.e.,*
                         initial and refresher training including knowledge checks) as required for power reactors under § 73.56 and part 26. Under § 73.120(c)(2)(ii), behavioral observation would be able to be performed in-person or remotely by video, and identified behavior of concern would need to be reported to plant supervision. The remote access alternative to face-to-face interactions provides substantial flexibility for licensees and applicants. Any video conferencing or other acceptable electronic means promoting face-to-face interaction for those individuals working remotely would demonstrate compliance with this regulation. Section 73.120(c)(3) captures and maintains the self-reporting of legal actions as an essential performance element to enhance the licensee's behavioral observation initiative similar to the current requirements under § 73.56(g), assuring that personnel who are granted and who maintain unescorted access are trustworthy and reliable. Section 73.120(c)(4) would provide a scalable approach for granting and maintaining unescorted access. One component not included from § 73.56 is the need for a psychological assessment and reassessment under § 73.56(e) for granting unescorted access and § 73.56(i)(v)(B) for individuals who perform one or more of the job functions described in § 73.120(b)(1)(ii) for maintaining unescorted access. Moreover, the requirement would permit criminal history updates to be completed within 10 years of the last review, compared to the three- or five-year reinvestigation periodicity for personnel at an operating commercial nuclear plant. In addition, no credit check re-evaluation would be required for these individuals. The continued need to maintain unescorted access would be evaluated on an annual basis by the reviewing official. Guidance in DG-5074, “Access Authorization Program for Commercial Nuclear Plants,” would specify that this evaluation should be based on a compilation of personnel interactions as described in the licensee's or applicant's policy and procedures for behavioral observation and the maintenance of an approved AA list. Section 73.120(c)(5) would require licensees and applicants to determine when a person no longer requires the need for unescorted access or no longer satisfies the AA requirement found within this section. Guidance in DG-5074 would further explain that licensees have the flexibility to terminate unescorted access to specific areas of the site if individuals lack the continued need for that access to perform their duties and responsibilities. Section 73.120(c)(6) would be consistent with the purpose of § 37.23(e) and would include the individual's right to correct and complete information as required under § 37.23(g). The section would include a requirement for designating a reviewing official. The language would provide clarity regarding the roles and responsibility of a reviewing official, who would be the only individual authorized to make unescorted access determinations. Section 73.120(c)(7) would align with the corresponding requirements under § 37.23(f), and § 73.120(c)(8) would align with the corresponding requirements under § 37.31. These requirements would encompass the roles and responsibilities for licensees, applicants, and if applicable, the contractor/vendors to establish, implement, and maintain a system of files and records to ensure personal information is not disclosed to unauthorized persons. Section 73.120(c)(9) would align with the requirements of § 37.33. Section 73.120(c)(10) would require licensees, applicants, and contractors or vendors to maintain the records that are required by the regulations in this section and retain them for a period of 3 years after the record is superseded or no longer needed. The record retention period of three years would be consistent with § 37.23(h), contrasting with the five-year retention period under § 73.56(o). Records maintained in any database(s) would need to be available for NRC review, consistent with the requirements found under § 73.56(o)(6)(ii). VI. Specific Requests for Comments The NRC is seeking advice and recommendations from the public on this proposed rule. We are particularly interested in comments and supporting rationale from the public on the following: Part 26—Fitness for Duty Program 1. The proposed rule under § 26.603(c) would enable a licensee or other entity to implement an FFD program under proposed § 26.604, “FFD program requirements for facilities that satisfy the § 26.603(c) criterion,” if the 
                        ( print page 86981)
                        licensee or other entity performs a site-specific analysis to demonstrate that the facility and its operation satisfy the criterion in § 53.860(a)(2). Should the NRC consider replacing its proposed § 26.603(c) criterion referencing § 53.860(a)(2) with an alternative requirement that if the commercial nuclear plant is of the class described in § 53.800, “Facility licensees for self-reliant-mitigation facilities,” and either § 53.800(a)(1) or (2) is satisfied, then drug and alcohol testing would not be required? This proposal would align the § 26.603(c) criterion with that proposed in the NRC-licensed operator regulatory framework of part 53. Please provide your considerations and rationale for your recommendation. Should the NRC also consider making a conforming change to the proposed § 73.120 criterion used for the AA program? Please provide your considerations and rationale for your recommendation. Part 26—Technology-Inclusive Approaches to Fatigue Management Requirements Applicable to Unit Outages In establishing the outage minimum days off requirement of § 26.205(d)(4), the NRC's objective was to ensure that individuals performing the duties described in § 26.4(a)(1) through (a)(4) have sufficient periodic long-duration breaks to prevent cumulative fatigue from degrading their ability to safely and competently perform their duties. In addition to the science of fatigue management, the NRC considered several factors in establishing the existing requirements. These additional factors were practical and safety considerations associated with the management of refueling outages for large LWRs, including the following: (1) the typical duration and frequency of outages; (2) the availability of contract personnel to perform the work; (3) the risk presented by the outage work while the reactor is shut down; and (4) the controls applied to the work that may limit the potential for latent errors to challenge reactor safety when the reactor is returned to power. The details of such considerations may differ for new reactor technologies or designs. Such considerations may not be relevant for some reactor designs (
                        *e.g.,*
                         reactors capable of on-line refueling) and there may be additional, more pertinent factors to consider for other designs. The NRC is seeking stakeholder input on whether alternative fatigue management requirements applicable to outages should be adopted to support technology-inclusive approaches that would be appropriate to support the licensing and regulation of future commercial nuclear plants. Please provide your considerations and rationale for your recommendation. Part 26—Draft Regulatory Guidance Approach for Fatigue Management In support of this proposed rule, the NRC has issued DG-5078, “Fatigue Management for Nuclear Power Plant Personnel at Commercial Nuclear Plants Licensed Under 10 CFR part 53.” This DG describes methods the NRC staff considers acceptable for addressing certain aspects of FFD programs at commercial nuclear facilities licensed under part53. The NRC staff also intends to eventually transition this draft guide into an update to RG 5.73, “Fatigue Management for Nuclear Power Plant Personnel,” or the development of a new RG. At this point, NRC staff is considering four options for future RG development: *Option 1: Amend the existing RG.*
                         The NRC may develop an updated version of RG 5.73 that continues to endorse (with clarifications, additions, and exceptions) the guidance contained in NEI 06-11, “Managing Personnel Fatigue at Nuclear Power Reactor Sites,” Revision 1, and incorporates the topics discussed within DG-5078 as new NRC staff positions in section C of RG 5.73.
                    *Option 2: Issue a new RG specific to part 53 licensees.*
                         The NRC may develop an entirely new RG applicable specifically to facilities licensed under part 53. This new RG would capture the guidance contained in DG-5078 and incorporate existing guidance (
                        *e.g.,*
                         selected guidance in RG 5.73 and NEI 06-11) that is considered to be technology inclusive in nature. The existing guidance (
                        *i.e.,*
                         RG 5.73) would remain in place as the guidance for facilities licensed under parts 50 and 52.
                    *Option 3: Review and potentially endorse new or revised industry-developed guidance.*
                         The NRC may engage with the industry regarding a potential update to industry guidance document NEI 06-11 or the development of new, separate industry-developed guidance specific to facilities licensed under part 53. The NRC would then review the new or revised industry-developed guidance within the NRC's RG process, which includes opportunities for public participation. New or revised industry-developed guidance could incorporate DG-5078 or propose alternatives for the NRC to consider.
                    *Option 4: Develop a comprehensive revision of the existing RG.*
                         The NRC may develop a more comprehensive revision of RG 5.73 that would explicitly detail all NRC positions reflected in the existing RG (including those endorsed positions currently contained in NEI 06-11, Revision 1), along with the guidance of DG-5078. Such a revision would thereby be a “stand-alone” document, without reference to or explicit endorsement of separate, industry-developed guidance. The NRC is seeking stakeholder input regarding which of the four options listed above would be optimal (or whether there are other options that the NRC should consider). Please provide your considerations and rationale for your recommendation. Part 53—Overall Organization Part 53 is structured as one framework with subparts providing technical, licensing, and administrative requirements for the various stages of the life cycle of a commercial nuclear plant. The organization of part 53 in this manner puts a complete set of requirements for each stage of the life cycle in a separate subpart with additional subparts for licensing and administrative requirements. The NRC is seeking comment on the proposed organization of the requirements in part 53 and possible improvements to how specific requirements (
                        *e.g.,*
                         examples of which specific sections) could be consolidated or otherwise reorganized to make the rule clearer or more concise. There are numerous references in proposed part 53 to other NRC regulations. Examples of such references include those in proposed § 53.610 to NRC regulations related to radiation protection (part 20), FFD (part 26), physical security (part 73), and MC&A (10 CFR part 74, “Material Control and Accounting of Special Nuclear Material”) for facilities receiving byproduct or SNMs. The NRC is seeking comment on whether such references to other regulations in various sections in the proposed part 53 provide benefits to applicants and licensees, or to other stakeholders seeking to understand the regulatory framework under part 53, or whether such references could be removed to reduce the length of part 53. Part 53, Subpart B—Comprehensive Risk Metrics The NRC is proposing to require the use of comprehensive risk metrics and associated risk performance objectives as one of several performance standards in part 53. Comprehensive risk metrics could include a risk metric or set of risk metrics that approximate the total overall risk from the facility to the 
                        ( print page 86982)
                        extent practicable. Associated risk performance objectives are preestablished values indicative of the comprehensive risk metrics that are used during risk-informed decision-making to gauge plant safety. Specifically, comprehensive risk metrics and associated risk performance objectives would provide one element of the safety criteria for LBEs other than DBAs in the proposed § 53.220. Comprehensive risk metrics, in the form of the IEFR and the ILCFR, and associated risk performance objectives, in the form of the QHOs of 5×10−7
                         per year and 2×10−6
                         per year, respectively, were similarly used in the LMP methodology to ensure that other evaluation criteria were conservatively defined and as a tool for focusing attention on matters important to managing the risks posed by nuclear power plants. The use of such comprehensive risk metrics and associated risk performance objectives in an integrated risk-informed decision-making process is similar to that used in RG 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,” Revision 3. The NRC is seeking comment on the use of comprehensive risk metrics and associated risk performance objectives in part 53 as one of several performance standards. The IEFR and ILCFR and the QHOs represent comprehensive risk metrics and associated risk performance objectives that the NRC has used for decades in a variety of capacities. What other performance standards could be used to address the comprehensive risks posed by proposed commercial nuclear plants? Please provide your considerations and rationale for your recommendation. If an applicant proposes a novel approach to comprehensive plant risk and the NRC approves the approach, should the resulting NRC-approved comprehensive plant risk metrics and associated risk performance objectives be codified or otherwise memorialized over time and, if so, how? Part 53, Subpart B—Defense in Depth Proposed § 53.250 would establish requirements based on the longstanding NRC philosophy of providing defense in depth to address uncertainties concerning the design, operation, and performance of commercial nuclear plants during LBEs. The NRC is seeking comment on the inclusion of the proposed requirements to assess and provide defense in depth. The NRC is also seeking comment on whether to include specific provisions in § 53.250 and subpart B to more explicitly address the possible role of inherent characteristics of some SSCs in preventing or mitigating unplanned events. The proposed § 53.250 is worded to preclude relying on a single engineered design feature to address the range of LBEs other than DBAs, which could possibly allow crediting inherent characteristics without further lines of defense. How could possible inherent characteristics of SSCs be considered in the proposed requirements in § 53.250 or in any alternative requirements for defense in depth provided in response to this item? Please provide your considerations and rationale for your recommendation. Part 53, Subpart C—Probabilistic Risk Assessment Current consensus PRA standards provide processes for appropriately defining the scope of a PRA and determining applicability of supporting requirements to suit the specific needs of a given applicant under proposed part 53. In addition to assessing other aspects of PRA acceptability such as PRA peer reviews, NRC determinations of the acceptability of such PRAs would assess the appropriateness of the applicant-defined scope as part of determining the applicability of a consensus PRA standard supporting an application. This approach is consistent with the current state of practice and offers appropriate flexibility for PRAs to be developed and assessed based on the application they are used to support, which includes consideration of how PRA results and insights are relied upon, together with factors such as safety margin, simplicity of design, and treatment of uncertainty. The NRC is seeking comment on what additional guidance, if any, is needed regarding PRA acceptability for Part 53 applicants and licensees. Part 53, Subparts C and D—Earthquake Engineering Proposed § 53.480 would establish requirements related to seismic design considerations. This proposed section is intended to provide a clear connection between siting activities and seismic design activities and to support various approaches to presenting seismic hazards and addressing those hazards in designs. The proposed requirements are intended to provide sufficient flexibility to allow approaches like those currently in parts 50 and 100 or approaches that might be endorsed by the NRC in the future that could incorporate more risk insights from PRAs. The NRC is seeking comment on whether the proposed requirements for earthquake engineering provide appropriate flexibility in addressing seismic risks while also ensuring that the regulations continue to adequately address seismic hazards. Please provide your considerations and rationale for your recommendation. Part 53, Subpart E—Construction and Manufacturing 1. Proposed § 53.610(b)(1)(iii) would require procedures that describe how construction will be controlled so as not to impact other features important to the design (
                        *e.g.,*
                         dewatering, slope stability, backfill, compaction, and seepage). The NRC is seeking comment on whether such specific requirements are useful or whether these requirements could be met through other requirements proposed in part 53 or already present in other relevant regulations (
                        *e.g.,*
                         quality assurance requirements in appendix B to part 50). Part 53, Subparts E and H—Manufacturing Licenses 1. The proposed requirements governing manufacturing are set forth in subpart E, and the proposed requirements governing the licensing processes are contained in subpart H. Some of the proposed requirements, including provisions related to the loading of unirradiated fuel into a manufactured reactor, are intended to cover a factory-fabrication model that has been suggested for some micro-reactor designs. However, as written, the proposed provisions are not limited to any size or type of reactor. The NRC is seeking comment on whether the proposed regulations are sufficient to govern various scenarios for the possible manufacturing and deployment of manufactured reactors. If a comment indicates that the proposed regulations are not sufficient, please describe the reasons why, including, if applicable, any plausible scenario for which the commenter believes the proposed regulations are not sufficient. 2. The proposed regulations in subpart H allow holders of or applicants for a COL to reference an ML but do not include such a provision for the holder of or applicant for a CP or OL. This proposed change from the current relationship between subparts in part 52 and the part 50 licensing process was made to simplify the provisions in the proposed part 53 for licensing and deploying manufactured reactors. The NRC seeks comment on whether part 53 should include provisions for an applicant for or a holder of a CP or an OL to reference an ML and, if so, how this should be done.
                        ( print page 86983) 3. Proposed § 53.1295 states that the holder of an ML could not begin manufacture of a manufactured reactor less than 6 months before the expiration of the license. This limitation is similar to the current restriction in § 52.177, which states that the manufacture of a reactor cannot begin less than 3 years before the expiration of the license. The restriction was revised from 3 years in part 52 to 6 months in the proposed part 53 in recognition of the likely use of MLs for a factory-fabrication model for micro-reactors. The NRC seeks comment on whether it is necessary or appropriate to revise the 3-year restriction in part 52 on when manufacturing activities could begin in relation to license expiration and, if so, what that restriction should be. 4. Proposed § 53.1288 provides the finality provisions for MLs and includes, as does existing § 52.171, limitations on the NRC's imposition of new requirements on either the design or the requirements for the manufacture of a manufactured reactor. No MLs have been issued under part 52 and there is no practical experience with the proposed finality sections. While the implications of the finality provisions related to the design of a manufactured reactor can reasonably be inferred from experience with DCs and COLs, there is no experience or available guidance regarding finality for “requirements for the manufacture of the manufactured reactor.” The NRC is seeking comment on the proposed finality provisions for MLs and specifically if and how finality for manufacturing processes might be requested and used. 5. The NRC is seeking comment on the proposed regulations for the loading of fresh (unirradiated) fuel into a manufactured reactor for subsequent transport to a site for which the Commission has issued a COL that authorizes construction and operation of a commercial nuclear plant using the manufactured reactor. The proposed regulation includes provisions for loading of fuel into manufactured reactors at a manufacturing facility prior to transporting the fueled reactor to its deployment site, as suggested by some stakeholders. The NRC has historically viewed reactor operation as including fuel load, and existing NRC regulations reflect this view. While the Act authorizes the NRC to issue licenses to manufacture production or utilization facilities, it does not contain specific provisions on fueling or operating facilities licensed under an ML, and existing ML regulations under part 52 do not include provisions for fuel load. The proposed rule addresses this matter by allowing an applicant to combine an ML with a part 70 license, which would authorize possession of a manufactured reactor in which the licensee has loaded unirradiated fuel provided at least two independent criticality prevention mechanisms are in place, each of which is sufficient to prevent criticality assuming optimum neutron moderation and neutron reflection conditions. This requirement would limit the possibility of creating fission products and allow the control of SNM, so that the loading of the fuel into a manufactured reactor could be governed primarily via a part 70 license and associated regulations (including those in subpart H of part 70). A specific topic on which the NRC is seeking comment is on the potential benefits of and issues with including the requirements of subpart H of part 70 within the proposed regulations for loading fuel into manufactured reactors at the manufacturing facility. For example, should the NRC include a threshold for including the requirements of subpart H of part 70 and, if so, what factors and decision criteria should be considered in such a threshold? If a comment indicates that the proposed regulations are not sufficient, please describe the reasons why, including the plausible scenarios for which the proposed regulations would not work or could be made to work better. 6. Section 170, “Indemnification and Limitation of Liability,” of the Act states that each license under section 103 shall have as a condition of the license a requirement that the licensee have and maintain financial protection of such type and in such amounts as the NRC shall require. The NRC is seeking comment on whether the proposed regulations should include amounts of required financial protections for MLs for fueled manufactured reactors, and, if so, what would be appropriate amounts of required financial protection. 7. Some stakeholders have suggested that a fueled manufactured reactor with appropriate protections against criticality should not be categorized as a utilization facility under NRC regulations or Section 11cc. of the Act. The NRC is seeking comment on possible approaches where the NRC could find that a fueled manufactured reactor would not be a utilization facility, the basis for such a finding, and the potential benefits of and potential issues with such a finding. 8. Proposed requirement § 53.620(d)(2)(i) would require a security program, including a physical security plan, for any ML authorizing possession of a manufactured reactor into which fuel has been loaded at the manufacturing facility. Currently, requirements in § 73.67(c)(1) only require that a physical security plan be submitted for those licensees who possess, use, transport, or deliver to a carrier for transport SNM of moderate strategic significance, or 10 kg or more of SNM of low strategic significance. The NRC is seeking comment on whether the proposed requirement: (1) should be specific to the facility type (
                        *i.e.,*
                         manufacturing facility) or be specific to the category of material being used at the facility; (2) should apply to all manufacturing plants, including those at which licensees may only possess SNM of low strategic significance (
                        *i.e.,*
                         category III), or only those facilities for which an applicant must submit a physical security plan per § 73.67(c)(1); or (3) should include more specific requirements on the supplemental security measures that may be needed for licensees possessing SNM of moderate strategic significance (
                        *i.e.,*
                         category II)? 9. Proposed requirement § 53.620(d)(2)(i) would require a cybersecurity program. The proposed general cybersecurity performance requirements would be to provide reasonable assurance that a cyberattack could not adversely impact the functions performed by digital assets used by the licensee for implementing the physical security, radiation monitoring, and criticality requirements. The NRC is seeking comment on the following: (1) to what extent stakeholders envision physical security controls, radiation monitoring, and criticality controls at a manufacturing facility being digital; (2) to what extent should the ML holder be required to protect digital computer and communications systems that impact safety and security functions from a cyberattack at a manufacturing facility authorized to load fuel; and (3) whether the rule provides sufficient clarity on the cybersecurity measures needed for license issuance or if additional detail should be included either in the rule or in guidance? 10. Proposed requirement § 53.620(d)(2)(i)(B) would require that the physical security program be designed to prevent unintended and uncontrolled criticality events. This would include criticality events that are initiated maliciously. The NRC is seeking comment on whether the ML holder should be required to design its security program to protect against radiological sabotage 
                        ( print page 86984)
                        (
                        *i.e.,*
                         an unintended criticality event leading to unacceptable radiological consequences), in addition to theft and diversion. For example, should the NRC establish security requirements to prevent an adversary, including an insider, from tampering with the reactor at a manufacturing facility or during transport in such a way as to cause an inadvertent criticality event? If so, should the NRC consider factors such as the category of fuel and the number of reactors at a factory that can simultaneously be loaded with fuel in establishing the security requirements? 11. Proposed requirement § 53.620(d)(2)(i) would require an ML holder to meet the performance objectives in § 73.67. Requirements § 73.67(e) and § 73.67(g) include provisions for security of category II and category III quantities of SNM, respectively, during transportation. The NRC is seeking comment on the extent to which the ML should require ASMs (
                        *i.e.,*
                         security measures above those required by § 73.67(e) and § 73.67(g)) for transportation of a fueled reactor to its place of operation. What should those measures be? 12. Proposed requirement § 53.620(d)(2)(i) would require an ML holder to meet the performance objectives of § 73.67. For licensees utilizing a category II quantity of SNM, the requirement in § 73.67(d)(4) would have the ML holder conduct a screening to confirm the identity of an individual prior to granting unescorted access to the controlled access area where the material is used or stored. The purpose of this requirement is to both confirm the identity of the individual and support a determination that the individual is trustworthy and reliable. The NRC is seeking comment on whether the ML requirements should include ASMs (
                        *i.e.,*
                         measures beyond those required by § 73.67(d)(4)) in order to provide reasonable assurance of identity confirmation and trustworthiness and reliability. 13. The NRC is seeking comment on whether provisions regulating the testing of fueled manufactured reactors in the manufacturing facility should be included in part 53 and, if so, what would be practical for the holder of an ML while also providing adequate protection of public health and safety. One possibility could be COLs that would be issued to the holders of an ML to cover low power (
                        *e.g.,*
                         <5% rated thermal power) nuclear physics testing of fueled manufactured reactors within the manufacturing facility prior to the manufactured reactors being transported to and incorporated into a commercial nuclear plant for the purpose of energy production. The NRC recognizes configuration changes are needed to perform nuclear physics testing and is seeking comment on what requirements should apply to the manufactured reactors and the manufacturing facility during such testing (
                        *e.g.,*
                         limiting power levels). If a comment indicates that the regulations should address limited operations at manufacturing facilities, please describe the likely scenarios that would need to be addressed and suggest what would be appropriate requirements for such scenarios. While an ML holder could accomplish nuclear physics testing by applying for a COL under the proposed subpart H of part 53, stakeholders have indicated that many of the requirements would likely be unnecessary, given the reduced risk profile posed by such activities. Therefore, the NRC is seeking comment on what requirements in subpart H of part 53 should apply to applicants for a COL who would perform testing of fueled manufactured reactors at the manufacturing plant. Examples of proposed requirements that might be relaxed or modified for applications for low power testing at manufacturing plants include those related to selection of LBEs to reflect limited inventory of radionuclides and decay heat, aircraft impact assessments, and earthquake engineering. Additionally, the NRC is seeking comment on whether several other requirements in part 53 could be modified for applications for a low power testing COL at a manufacturing facility. For example, the NRC is seeking comment on how portions of the ML facility used to support testing should fall within the requirements for construction activities under § 53.610; whether §§ 53.710 and 53.715 (SSC configuration control) must be implemented to ensure portions of the ML facility relied on to limit potential radiological consequences from LBEs are available to perform their safety functions; and whether the requirements of § 53.730 could be modified to reflect the conditions of low power physics testing. If a comment indicates that some design and analysis requirements and related application requirements in subpart H of the proposed part 53 are not needed for the testing of fueled manufactured reactors, please provide a rationale supporting your comment and, if applicable, what alternate requirements would be appropriate. Moreover, the licensing mechanism for the facility could present unique challenges. One option could be to issue a low power testing COL for each fueled manufactured reactor to be tested. This would comport with the agency's practice of issuing one license per reactor but could prove prohibitive from a cost standpoint and may provide very little safety benefit if all manufactured reactors are the same. Alternatively, one low power testing COL could be issued for the portions of the ML facility used to test the fueled manufactured reactors and allow multiple fueled manufactured reactors to be completed and tested over the course of the ML. Under this approach, any ITAAC related to testing of the fueled manufactured reactors would need to be closed after they were manufactured but prior to testing, and the NRC would issue a notice of intended operation and provide the public an opportunity to request a hearing on whether each fueled manufactured reactor as constructed complies, or on completion will comply, with the acceptance criteria of the license. The NRC is seeking comment on the potential benefits and issues with having a COL for each fueled manufactured reactor to be tested versus having a COL cover the testing of multiple fueled manufactured reactors. If a comment indicates a preference for a particular approach, please provide a rationale supporting the comment and describe the specific scenarios that the regulations need to address. Part 53, Subpart F—Staffing and Generally Licensed Reactor Operators Under the Act Sections 106 and 107, the NRC is proposing to group commercial reactors into classes upon the basis of the similarity of operating and technical characteristics of the facilities, and then to prescribe uniform conditions for licensing individuals as operators of any of the various classes; determine the qualifications of such individuals; and, for certain classes of commercial reactors, issue general licenses (
                        *i.e.,*
                         licenses for which no application is needed) to such individuals allowing the individuals to operate the commercial reactor. 1. 
                        *Categories of Individuals Who May Manipulate Facility Controls:*
                         The NRC is proposing requirements that would allow the manipulation of the controls of certain facilities by GLROs in lieu of specifically licensed reactor operators and senior reactor operators. Reactor operators and senior reactor operators are the only categories of individuals currently allowed to be licensed to manipulate the controls of utilization facilities under part 55. The NRC is interested in public perspectives on this proposed addition of the GLRO category, particularly in light of new reactor technologies and concepts of operations.
                        ( print page 86985) 2. 
                        *Criteria for GLRO Staffing:*
                         The NRC is proposing criteria under which facilities would be staffed by GLROs in lieu of specifically licensed reactor operators and senior reactor operators. These criteria establish a new class of self-reliant-mitigation facilities, as defined in part 53, for which distinct GLRO licensing and staffing requirements would apply. The NRC is soliciting public feedback regarding whether these proposed criteria are appropriate and what, if any, alternative criteria should be considered. Please provide your considerations and rationale for your answer. 3. 
                        *Medical Requirements for GLROs:*
                         Based on the proposed criteria that a self-reliant-mitigation facility, as defined in part 53, must meet, the NRC is proposing not to subject GLROs to requirements for medical fitness and medical examination. This is in contrast with the proposed requirements associated with specifically licensed reactor operators and senior reactor operators, as well as the existing requirements for reactor operators and senior reactor operators under part 55. The NRC is soliciting public feedback regarding whether GLROs should be subject to medical fitness and/or medical examination requirements like reactor operators and senior reactor operators. Please provide your considerations and rationale for your answer. 4. 
                        *Onshift Engineering Expertise:*
                         The NRC is proposing to require that engineering expertise be accounted for within facility staffing plans. This proposed requirement would be in lieu of the traditional position of the Shift Technical Advisor. The NRC is further proposing that individuals providing such engineering expertise would need, among other things, to possess either a qualifying 4-year degree or licensure as a Professional Engineer. The NRC is interested in feedback from the public regarding the appropriateness of this requirement, including any alternatives that should be considered. Please provide your considerations and rationale for your answer. 5. 
                        *Use of Simulation Facilities as HFE Testbeds:*
                         The NRC is proposing to establish regulations pertaining to the use of simulation facilities within the context of the licensing programs both for specifically licensed reactor operators and senior reactor operators as well as for GLROs. However, these regulations, as currently proposed, do not address the use of simulation facilities within the context of serving as testbeds for HFE-related analyses and assessments. Rather, the NRC currently envisions that the use of simulation facilities as HFE testbeds is more appropriately addressed via guidance documents. The NRC is soliciting public feedback regarding whether simulation facility requirements should also address the use of simulation facilities as HFE testbeds. Please provide your considerations and rationale for your answer. Part 53, Subpart F—Emergency Preparedness and Security Programs 1. The proposed framework for part 53 would incorporate the changes to NRC regulations from the final rulemaking on “Emergency Preparedness for Small Modular Reactors and Other New Technologies” (the EP for SMR/ONT rule) by including references to § 50.160, “Emergency preparedness for small modular reactors, non-light-water reactors, and non-power production or utilization facilities,” and by making conforming changes within § 50.160. The proposed framework for part 53 would also introduce a graded approach to physical protection requirements that includes the criterion in § 53.860(a)(2)(i) to establish a class of licensees that would not be required to protect against the design-basis threat (DBT) of radiological sabotage. The NRC is soliciting public comment relating to these topics, which could include ways that graded approaches for both emergency preparedness and security programs might be assessed and considered during the licensing process. The NRC is seeking comment on the sufficiency and clarity of requirements in proposed part 53 related to the assessments needed to support graded emergency planning and security. If a comment indicates that there is an issue with the sufficiency or clarity of the proposed regulations, please describe the reasons why, including, if applicable, any scenario for which the proposed regulations are not sufficient and possible ways to clarify the requirements. The NRC is specifically seeking comment on possible challenges arising from the interactions between the proposed regulations and related assessments for grading the requirements for emergency planning and security. 2. The NRC is preparing various guidance documents to support this rulemaking and other ongoing or recently completed rulemakings related to emergency preparedness and security. DG-5076, “Guidance for Technology-Inclusive Requirements for Physical Protection of Licensed Activities at Commercial Nuclear Plants,” has been issued along with this proposed rulemaking and public comments are requested via this notice on that draft guidance. The NRC is also planning to issue a draft revision of RG 1.242, “Performance-Based Emergency Preparedness for Small Modular Reactors, Non-Light-Water Reactors, and Non-Power Production or Utilization Facilities,” for public comment. The planned revision to RG 1.242 would add guidance for part 53 applicants and licensees. In the staff requirements memorandum to SECY-23-0021, the Commission directed the NRC staff to address the consideration of security-related events for an advanced reactor that addresses security through design and engineered safety features when it harmonizes this rulemaking with the EP for SMR/ONT rule. In the EP for SMR/ONT rule, the NRC established an alternative performance-based and risk-informed approach for emergency planning, including determining the need for and size of an emergency planning zone (EPZ) to support predetermined, prompt protective actions. The NRC has incorporated the relevant rule language from the EP for SMR/ONT rule into this proposed rule and is seeking stakeholder feedback as to whether additional rule language changes or additional guidance would be beneficial. In light of the Commission direction and the above considerations, the NRC is assessing how best to address the treatment of security-related events in emergency planning, including in the determination of EPZ size, for reactors licensed under part 53. Part 53 is introducing an alternative approach to meeting security regulations that should be taken into consideration under § 50.160. Stakeholders are encouraged to take a holistic view of the various activities and opportunities to provide comments on this rulemaking and related guidance supporting this rulemaking (
                        *e.g.,*
                         DG-5076 on physical protection requirements, future revisions to RG 1.242). In developing comments, the NRC urges stakeholders to consider various scenarios that might arise when implementing graded approaches for security and emergency planning for various reactor designs. Scenarios could include the following: the potential consequences from security events up to and including the DBT of radiological sabotage are bounded by unlikely and very unlikely event sequences such that security events do not need separate analyses in the EPZ size determination;( print page 86986)
                    the potential consequences from security events up to and including the DBT are not bounded by unlikely and very unlikely event sequences but could otherwise support a reduced EPZ size consistent with considerations discussed in RG 1.242 and NUREG-0396, “Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants”; or
                    the potential consequences from security events up to and including the DBT are not bounded by unlikely and very unlikely event sequences and warrant consideration of increasing the size of the EPZ. The NRC is interested in comments on the need for additional rule language or guidance to address graded approaches for emergency planning and security programs under the scenarios described above for part 53 applicants and licensees. Please address within the comments any technical, policy, or legal issues that are associated with your suggestions. Part 53, Subpart F—Integrity Assessment Program Requirements Decades of operating experience with LWRs suggests that phenomena such as environmentally assisted fatigue and chemical interactions could impact certain SSCs during the life of a commercial nuclear plant. Under the existing regulatory framework, historically, some of these phenomena were not addressed during early licensing reviews but were identified and addressed later when significant safety issues arose (
                        *e.g.,*
                         see numerous generic letters, bulletins, orders, and development and implementation of vessel integrity and materials reliability programs) or a licensee voluntarily pursued renewal of an OL under part 54. The NRC is proposing to include a new set of programmatic requirements for an Integrity Assessment Program that would ensure these phenomena are addressed early in the life of a commercial nuclear plant licensed under part 53. The requirements would be provided in § 53.870. The NRC is seeking comment on whether the proposed requirements under the Integrity Assessment Program appropriately complement design requirements to address concerns regarding aging, cyclic or transient load limits, and degradation mechanisms related to chemical interactions, operating temperatures, effects of irradiation, and other environmental factors. In addition, the NRC is interested in views on whether, and if so how, degradation mechanisms are or could be addressed in other programs. Part 53, Subpart G—Decommissioning 1. On March 3, 2022, the NRC published the proposed rule entitled “Regulatory Improvements for Production and Utilization Facilities Transitioning to Decommissioning” (87 FR 12254). This rulemaking would amend the NRC's current regulations to provide an appropriate regulatory framework for nuclear power reactors transitioning from operations to decommissioning. The rulemaking would address lessons learned from licensees that have completed or are currently in the decommissioning process. The NRC staff sent a draft final rule to the Commission for its consideration on January 31, 2024, in SECY-24-0011, “Final Rule: Regulatory Improvements for Production and Utilization Facilities Transitioning to Decommissioning (3150-AJ59; NRC-2015-0070).” What aspects of this draft final rule, if any, should be incorporated in a part 53 final rule and why? 2. Proposed § 53.1060(b) in subpart G would require that, “No later than 30 days after the Commission publishes notice in the 
                        **Federal Register**
                         under § 53.1452(a), the licensee must submit a report containing a certification that financial assurance for decommissioning is being provided in an amount specified in the licensee's most recent updated certification, including a copy of the financial instrument obtained to satisfy § 53.1040.” This is similar to the current requirement in § 50.75(e)(3) for part 52 COL holders. The NRC is seeking comment on whether commercial nuclear plant COL holders under part 53 should have the same requirement as COL holders under part 52 to demonstrate that they have financial assurance in place no later than 30 days after the Commission issues the notice of intended operation under § 53.1452. Please provide your considerations and rationale for your answer. Part 53, Subpart H—Licenses To Construct and Operate Commercial Nuclear Plants of Identical Design at Multiple Sites In addition to including provisions in part 53, subpart H, for referencing ESPs, standard design approvals, and design certifications in applications for commercial nuclear plants, the proposed § 53.1470 provides optional requirements related to the submittal and NRC review of CP, OL, and COL applications to construct and operate commercial nuclear plants of identical design at multiple sites, similar to requirements found in appendix N in both 10 CFR parts 50 and 52. This section would set out the particular requirements and provisions applicable to situations in which applications for CPs and subsequent OLs, or COLs, under this part, are filed by one or more applicants for licenses to construct and operate nuclear power reactors of identical design (“common design”) to be located at multiple sites. Hearings for applications filed under appendix N in both parts 50 and 52 are governed by subpart D of part 2, as would be the case for future part 53 applications under proposed § 53.1470. Under the proposed requirements in this section, each application is to be treated as a separate application, with the exception of the common design, and so would require separate applications, separate determinations of sufficiency for docketing, separate notices of docketing, and so forth. Proposed § 53.1470 would also require that each application list all the applications that are to be treated together to ensure that the NRC is clearly informed of the intentions of all applicants. Ordinarily, the NRC would publish in the 
                        **Federal Register**
                         a separate notification of docketing for each application, so that delays in the docketing of one application would not delay the docketing and subsequent technical review of other applications. However, if circumstances allow (
                        *e.g.,*
                         sufficiency review for multiple applications are completed simultaneously), the NRC could publish a single notice of docketing for multiple applications. With regard to how the NRC would fulfill its obligations under the National Environmental Policy Act of 1969, as amended, the NRC staff would prepare a separate environmental document for each application, but the NRC could conduct joint scoping on environmental issues related to the common design. If the applications reference a standard design certification or the use of a manufactured reactor, then the environmental document would need to incorporate by reference the environmental assessment (EA) prepared for either the design certification or the ML, as applicable. In addition, § 53.1470 would require the ACRS to report on each of the applications, as would be required by provisions in subpart H of part 53. Each ACRS report would be limited to the safety matters which are not relevant to the common design. In addition, the ACRS would need to issue a report on the safety of the common design—except for those matters relevant to the 
                        ( print page 86987)
                        safety of a referenced design certification or manufactured reactor. Given this synopsis of how the requirements in proposed § 53.1470 would be implemented as currently written, the NRC is seeking comment on whether there are opportunities to allow added flexibility for applicants under these provisions. This could include consideration of whether applications for which the “common design” is not completely identical could be evaluated under this provision and, if so, what the process would be for determining the appropriateness of a common review. In addition, the NRC is interested in feedback about the pros and cons of requiring that applications under these proposed provisions be submitted at the same time versus allowing them to be submitted on a staggered basis. Part 53, Subparts H and I—Probabilistic Risk Assessment Information Proposed § 53.1239(a)(18) in subpart H and the related references to this proposed requirement for the holders of OLs and COLs would require a description of the PRA required by § 53.450(a), and its results to be included in FSARs. However, guidance documents may further clarify the division of PRA-related information needed to be in the FSAR, in other possible licensing basis documents, and controlled as plant records subject to inspections and audits. For example, a possible approach could be to include a summary of the PRA results in the FSAR and control that information under § 53.1545 and create a separate document related to the broader PRA analyses and related processes as a program document under § 53.1560. The program document would provide more detail than the summaries in the FSAR but still be a much-condensed source of information in comparison to the documentation of the PRA. This possible approach would reflect the role of the PRA in the licensing process under part 53 and in maintaining margins to the safety and evaluation criteria in subparts B and C but may allow a more appropriate evaluation process to address the particulars and complexities of the PRA-related documents. The NRC is seeking comment on the appropriate placement of PRA-related information among various licensing basis documents and plant records. In addition to the placement of PRA-related information, the NRC is seeking comment on the appropriate control of that information and on the routine submittal of updates to the NRC. Please provide your considerations and rationale for your answer. Part 53, Subparts H and I—Changes to Manufacturing Licenses Proposed § 53.1530 would not allow the holder of an ML or the holder of a COL using a manufactured reactor to make changes to the design of the manufactured reactor without requesting a license amendment from the NRC. The proposed requirements do not include a specific mention of the manufacturing processes for which the NRC could possibly provide finality under proposed § 53.1288. The NRC is seeking comment on the appropriate change control provisions for MLs, including whether criteria could be developed to determine when a license amendment request would not be required and whether those criteria should address changes in manufacturing processes as well as changes in the design. Please provide your considerations and rationale for your recommendation. Financial Qualifications Utility new reactor applicants are exempt under § 50.33(f) from financial qualification reviews because they are generically presumed to be financially qualified for operations. In contrast, merchant power plant new reactor applicants are required under § 50.33(f)(2) to submit information that demonstrates they possess or have reasonable assurance of obtaining the funds necessary to cover estimated construction and operating costs for the period of the license. A “merchant power plant new reactor applicant” is a non-rate-regulated entity (
                        *e.g.,*
                         a nonutility) that engages in the business of production, manufacturing, generating, buying, aggregating, marketing, or brokering electricity for sale at wholesale or for retail sale to the public. Over the past decade, the agency has heard some concerns about the challenges that merchant power plant applicants face in demonstrating compliance with the current financial qualification requirements. Does this standard continue to pose challenges for merchant power plant applicants? If so, please provide a detailed explanation of these challenges. Should part 53 have the same financial qualification requirements as parts 50 and 52? Why or why not? Are there categories of merchant new reactor applicants for which a part 70 “appears to be financially qualified” standard would be more appropriate? [15]
                        
                         If so, please explain what types of applicants should be able to use the part 70 financial qualification standard and what distinguishes these applicants from ones that should not be able to use this standard. If a part 70 financial qualification standard were to apply to a category of merchant new reactor applicants, should it also apply to pre-construction license transfer applications for these reactors? Why or why not? Is there another standard the agency should consider for financial qualification of merchant new reactor applicants? Commenters are encouraged to provide specific suggestions and the basis for those suggestions. Part 73, Section 73.100—Physical Security The proposed § 73.100 would identify the proposed performance-based physical security requirements with which future commercial power reactor applicants or licensees' physical protection programs would need to demonstrate compliance, without prescribing the specific methods that must be used to satisfy them. Applicants and licensees would have increased flexibility regarding the modern technologies and methods that they could use. Implementing guidance in DG-5076 (proposed RG 5.97), “Guidance for Technology Inclusive Requirements for Physical Protection of Licensed Activities at Commercial Nuclear Plants,” would be available to assist applicants and licensees. For example, DG-5076 provides detailed guidance, including performance standard recommendations, on the probability of detection and alternative sources of power for exterior intrusion detection systems (subsection 4.1.1.1.A), interior intrusion detection (subsection 4.1.1.1.B), intrusion assessment (subsection 4.1.1.2.A), security response/neutralization subsection (4.1.1.4.A), security communication (subsection 4.1.1.3.A), and security delay (subsection 4.1.1.4.C). Does the NRC's proposed approach in § 73.100 provide a sufficient level of detail to be readily understood and easily applied to the licensing and oversight of new and advanced power reactors, or should the NRC consider moving some objective and measurable security performance standard recommendations from the draft implementing guidance in DG-5076 into proposed § 73.100? If so, which objective and measurable security performance standard recommendations should be moved from DG-5076 to § 73.100? Please provide the basis for your response.
                        ( print page 86988) Part 73, Section 73.110—Cybersecurity The proposed § 73.110 would require licensees to demonstrate protection against cyberattacks in a manner that is commensurate with the potential consequences from those attacks, without prescribing the specific methods that must be used to demonstrate protection. Under proposed § 73.110(a), licensees would need to ensure that digital computer and communications systems are adequately protected against a potential cyberattack that would, for example, result in adverse impacts to the physical security digital assets used by the licensee to prevent unauthorized removal of material per § 53.860(a). Protecting against such a potential cyberattack would involve requiring cybersecurity for SNM at a commercial nuclear reactor licensed under part 53. Applicants and licensees would have increased flexibility regarding the modern technologies and methods that they could use for protecting against such a potential cyberattack. Detailed implementing guidance in DG-5075 (proposed RG 5.96), “Establishing Cybersecurity Programs for Commercial Nuclear Plants licensed under 10 CFR part 53,” would be available to assist applicants and licensees. For example, DG-5075 provides guidance on the implementation of security by design features (
                        *e.g.,*
                         facility design) for negating the potential consequences from such a potential cyberattack. If a cyberattack were to compromise the availability, integrity, or confidentiality of data or systems associated with security systems/measures for the protection of SNM at a commercial nuclear reactor licensed under part 53, do the potential consequences warrant requiring cybersecurity for such material? Please provide the basis for your response including a detailed explanation of challenges, if any, posed by requiring cybersecurity for SNM at a commercial nuclear reactor licensed under part 53. Recent Legislation On July 9, 2024, the President signed into law the Accelerating Deployment of Versatile, Advanced Nuclear for Clean Energy Act of 2024, also referred to as the ADVANCE Act. Section 203, “Licensing Considerations Relating to Use of Nuclear Energy for Nonelectric Applications,” and Section 208, “Regulatory Requirements for Micro-Reactors,” of the ADVANCE Act specifically mention the technology-inclusive regulatory framework to be established under section 103(a)(4) of NEIMA as a potential vehicle to be considered for the report to Congress required under section 203 and a potential vehicle to implement strategies and guidance for the licensing and regulation of micro-reactors required under section 208. This proposed rulemaking is, in part, how the NRC is implementing section 103(a)(4) of NEIMA. The NRC is seeking comment on how part 53 could be revised to better enable its potential use to implement the ADVANCE Act. Specifically, Section 208 of the ADVANCE Act requires the NRC to develop and implement “risk-informed and performance-based strategies and guidance” in several areas for the licensing and regulation of micro-reactors, including with respect to “licensing mobile deployment.” The ADVANCE Act requires the NRC to consider “the unique characteristics of micro-reactors,” including physical size, design simplicity, and source term; opportunities to incorporate specific improvements related to streamlining the review process; and other policy and licensing issues. With regard to implementation, the ADVANCE Act provides the NRC with three options. The NRC may implement the developed strategies and guidance, as appropriate, via (1) the existing regulatory framework, (2) the Part 53 rulemaking, or (3) a pending or new rulemaking. Given the language included in Section 208, the NRC is seeking comment on how part 53 could be revised to better address the ADVANCE Act's requirements related to strategies and guidance for micro-reactors. VII. Section-by-Section Analysis The following paragraphs describe the specific changes proposed by this rulemaking. § 1.43 Office of Nuclear Reactor Regulation This proposed rule would revise § 1.43(a)(2) to extend the authority of the Office of Nuclear Reactor Regulation to regulate source, byproduct, and SNM at facilities licensed under part 53. § 2.1 Scope This proposed rule would revise § 2.1(e) to apply to standard design approvals under part 53. § 2.4 Definitions This proposed rule would revise § 2.4 to update the definition of “
                        *Contested proceeding*
                        ” to include NRC enforcement actions against applicants for a standard DC under part 53. It would also update the definition of “
                        *Facility*
                        ” to encompass utilization facilities as defined in § 53.020 (there are no production facilities under part 53). § 2.100 Scope of Subpart This proposed rule would revise § 2.100 to extend the scope of subpart A to licenses and standard design approvals issued under §§ 53.1200 through 53.1221. § 2.101 Filing of Application This proposed rule would revise § 2.101 to be applicable to part 53 applicants in addition to part 50 and 52 applicants by adding references to part 53 in paragraphs (a)(3)(i), (a)(5), and (a)(9). § 2.104 Notice of Hearing This proposed rule would extend the hearing notice requirement in § 2.104(a) to applications concerning facilities covered under part 53. Footnote 1 to § 2.104 would be revised in a corresponding manner. § 2.105 Notice of Proposed Action This proposed rule would revise § 2.105 to extend the requirement in § 2.104 to publish a notice of intended operation or a notice of proposed action, as applicable, to part 53 applicants in addition to part 50 and 52 applicants by adding corresponding references to part 53 in paragraphs (a), (a)(4), (a)(10), (a)(12), (a)(13), and (b)(3). § 2.106 Notice of Issuance This proposed rule would revise § 2.106 to extend the issuance notice requirement to applications concerning facilities covered under part 53 through updated references in paragraphs (a)(2) and (3), and (b)(2). § 2.109 Effect of Timely Renewal Application This proposed rule would revise § 2.109 to add references to part 53 in paragraphs (b), (c), and (d) regarding the timing of license renewal applications. § 2.110 Filing and Administrative Action on Submittals for Standard Design Approval or Early Review of Site Suitability Issues This proposed rule would revise § 2.110 to include references to part 53 in paragraphs (a)(1) and (b). § 2.202 Orders This proposed rule would revise § 2.202(e) to add references to part 53 regarding the requirements to be followed for orders involving the modification of a license, COL, ESP, standard DC rule, standard design approval, or ML.
                        ( print page 86989) § 2.309 Hearing Requests, Petitions To Intervene, Requirements for Standing, and Contentions This proposed rule would revise § 2.309 to include references to part 53 in paragraphs (a), (f)(1)(i), (f)(1)(vi) and (vii), (g), (h)(2), (i)(2), and (j) regarding a request for hearing under § 53.1452. § 2.310 Selection of Hearing Procedures This proposed rule would revise § 2.310 by revising paragraph (a), the introductory text for paragraph (h), and paragraphs (i) and (j) to incorporate references to part 53 regarding hearing procedures. § 2.329 Prehearing Conference This proposed rule would revise § 2.329(a) to extend the timing requirements for prehearing conferences involving CPs and licenses under part 53. § 2.339 Expedited Decision-Making Procedure This proposed rule would revise § 2.339(d) to include references to part 53 regarding expedited decision-making procedures. § 2.340 Initial Decision in Certain Contested Proceedings; Immediate Effectiveness of Initial Decisions; Issuance of Authorizations, Permits and Licenses This proposed rule would revise § 2.340 regarding initial decisions of a presiding officer in certain contested proceedings, the effective date of those decisions, and the issuance of authorizations, permits, and licenses, by incorporating references to part 53 in paragraphs (b), (c), (d), (f), (i), and (j). § 2.341 Review of Decisions and Actions of a Presiding Officer This proposed rule would revise § 2.341(a)(1) to include an updated reference to part 53 regarding the allowance of a period of interim operation. § 2.400 Scope of Subpart This proposed rule would revise § 2.400 to extend the scope of subpart D of part 2 to include part 53 applicants for licenses to construct or operate nuclear power reactors of identical design at multiple sites. § 2.401 Notice of Hearing on Construction Permit or Combined License Applications Pursuant to Appendix N of 10 CFR Parts 50, 52, or 53 This proposed rule would revise the section heading and § 2.401 to extend the hearing notice requirement to applications concerning facilities covered under part 53. § 2.402 Separate Hearings on Separate Issues; Consolidation of Proceedings This proposed rule would revise § 2.402(a) to apply provisions regarding separate hearings and the consolidation of proceedings to part 53 applicants. § 2.403 Notice of Proposed Action on Applications for Operating Licenses Pursuant To Appendix N of 10 CFR Part 50 This proposed rule would revise § 2.403 to require the Commission to publish a notification of proposed action in the 
                        **Federal Register**
                         after applications under part 53 are docketed. § 2.404 Hearings on Applications for Operating Licenses Pursuant to Appendix N of 10 CFR Part 50 This proposed rule would revise § 2.404 to apply to applications for an OL under part 53. § 2.405 Initial Decisions in Consolidated Hearings This proposed rule would revise § 2.405 to be applicable to CPs, full-power OLs, and COLs under part 53. § 2.406 Finality of Decisions on Separate Issues This proposed rule would revise § 2.406 to be applicable to proceedings conducted pursuant to part 53. § 2.500 Scope of Subpart This proposed rule would revise § 2.500 to extend the provisions of subpart E of part 2 to include applications for a license to manufacture nuclear power reactors under part 53. § 2.501 Notice of Hearing on Application Under Subpart F of 10 CFR Part 52 or 53 for a License To Manufacture Nuclear Power Reactors This proposed rule would revise the section heading and § 2.501(a) by extending its provisions to applications for a license to manufacture nuclear power reactors under part 53. § 2.643 Acceptance and Docketing of Application for Limited Work Authorization This proposed rule would revise § 2.643(b) regarding the acceptance and docketing of an application for a CP for a utilization facility of the type specified in part 53. § 2.645 Notice of Hearing This proposed rule would revise § 2.645(a) to incorporate a reference to part 53. § 2.649 Partial Decisions on Limited Work Authorization This proposed rule would revise § 2.649 to extend its provisions to LWAs issued under part 53. § 2.800 Scope and Applicability This proposed rule would revise § 2.800 by revising paragraphs (c) and (d) to incorporate references to part 53 regarding the scope and applicability of the rulemaking procedures contained in this subpart. § 2.801 Initiation of Rulemaking This proposed rule would revise § 2.801 to include a reference to part 53. § 2.813 Written Communications This proposed rule would revise § 2.813(a) to apply general requirements for correspondence with the Commission to communications concerning part 53, in addition to parts 50, 52, and 100. § 2.1103 Scope of Subpart K This proposed rule would revise the first sentence of § 2.1103 to extend the provisions of subpart K of part 2 to licenses under part 53 to expand the spent fuel capacity at the site of a civilian nuclear power plant. § 2.1202 Authority and Role of NRC Staff This proposed rule would amend § 2.1202 by revising paragraphs (a)(1) through (3), and (a)(6) to include references to part 53. § 2.1301 Public Notice of Receipt of a License Transfer Application This proposed rule would revise § 2.1301(b) to include a corresponding reference to license transfers under part 53 in addition to parts 50 and 52. § 2.1403 Authority and Role of the NRC Staff This proposed role would update § 2.1403 to specify that “significant hazards considerations” has the same meaning as defined in part 53. § 2.1500 Purpose and Scope This proposed rule would revise § 2.1500 to extend the scope of subpart O of part 2 to DC rulemaking hearings under part 53. § 2.1502 Commission Decision To Hold Legislative Hearing This proposed rule would revise § 2.1502, paragraphs (a) and (b)(1) to 
                        ( print page 86990)
                        incorporate references to part 53 regarding the Commission's decision to hold a DC rulemaking. § 10.1 Purpose This proposed rule would revise § 10.1(a)(3) to include a reference to part 53. § 10.2 Scope This proposed rule would revise § 10.2(b) to extend the scope of subpart A to applicants and holders of licenses, certificates, and standard design approvals under part 53 in addition to part 52. § 11.7 Definitions This proposed rule would revise § 11.7 such that terms defined in part 53 have the same meaning when used in part 11. § 19.2 Scope This proposed rule would revise § 19.2(a) to include references to part 53. § 19.3 Definitions This proposed rule would revise the definitions of “
                        *License*
                        ” and “
                        *Regulated entities*
                        ” in § 19.3 to incorporate references to part 53. § 19.11 Posting of Notices to Workers This proposed rule would amend § 19.11 by revising paragraphs (a), (b), and (e)(1) to apply to applicants and holders of licenses, permits, standard design approvals, and standard DCs under part 53 in addition to part 52. § 19.14 Presence of Representatives of Licensees and Regulated Entities, and Workers During Inspections This proposed rule would revise § 19.14(a) to apply to applicants and holders of a license, standard design approval, ESP, or standard DC under part 53 in addition to part 52. § 19.20 Employee Protection This proposed rule would revise § 19.20 to include a reference to protected activities under part 53. § 20.1002 Scope This proposed rule would revise the first sentence of 10 CFR part 20, “Standards for Protection Against Radiation,” § 20.1002 to extend the scope of part 20 to apply to persons licensed by the Commission to receive, use, transfer, or dispose of byproduct, source, or SNM or to operate a production or utilization facility under part 53. § 20.1003 Definitions This proposed rule would revise § 20.1003 to update the definition of “
                        *License*
                        ” to include those issued under part 53. § 20.1101 Radiation Protection Programs This proposed rule would revise § 20.1101(d) to exclude licensees subject to § 53.260 from its requirements. § 20.1401 General Provisions and Scope This proposed rule would revise § 20.1401, paragraphs (a) and (c) to extend the scope of subpart E of part 20 to apply to the decommissioning of facilities licensed under part 53 and the release of part of a facility or site for unrestricted use in accordance with § 53.1080. § 20.1403 Criteria for License Termination Under Restricted Conditions This proposed rule would revise § 20.1403(d) to include decommissioning plans under part 53. § 20.1404 Alternate Criteria for License Termination This proposed rule would revise § 20.1404(a)(4) to include a reference to part 53 regarding alternate criteria for license termination. § 20.1406 Minimization of Contamination This proposed rule would revise § 20.1406(a) to include references to applicants for licenses other than ESPs or MLs under part 53. It would also revise § 20.1406(b) to include references to standard DCs and standard design approvals under part 53 in addition to part 52. § 20.1501 General This proposed rule would revise § 20.1501(b) regarding the requirement for retention of records from surveys describing the location and amount of subsurface residual radioactivity at a site to include a reference to the retention requirements under part 53. § 20.1905 Exemptions to Labeling Requirements This proposed rule would revise § 20.1905(g) to apply to facilities licensed under part 53 in addition to parts 50 and 52 regarding exemptions to labeling requirements. § 20.2004 Treatment or Disposal by Incineration This proposed rule would revise § 20.2004(b)(1) to include references to part 53 regarding the treatment or disposal of waste oil by incineration. § 20.2201 Reports of Theft or Loss of Licensed Material This proposed rule would revise § 20.2201 to include references to part 53 in paragraphs (a)(2)(i), (b)(2)(i) and (c) regarding requirements for reports of theft or loss of licensed material. § 20.2202 Notification of Incidents This proposed rule would revise § 20.2202(d)(1) to add references to part 53 regarding reports to the NRC Operations Center. § 20.2203 Reports of Exposures, Radiation Levels, and Concentrations of Radioactive Material Exceeding the Constraints or Limits This proposed rule would revise § 20.2203(c) to refer to procedures under part 53 for reporting occurrences of exposures, radiation levels, and concentrations of radioactive material exceeding the constraints or limits. § 20.2206 Reports of Individual Monitoring This proposed rule would revise § 20.2206(a)(1) to include a reference to part 53. § 21.2 Scope This proposed rule would revise § 21.2, paragraphs (a), (b), and (c) to include references to part 53 regarding the scope and applicability of part 21 requirements. § 21.3 Definitions This proposed rule, in § 21.3 would revise the definitions of “
                        *Basic component,*
                        ” “
                        *Commercial grade item,*
                        ” “
                        *Critical characteristics,*
                        ” “
                        *Dedicating entity,*
                        ” “
                        *Dedication,*
                        ” “
                        *Defect,*
                        ” and “
                        *Substantial safety hazard*
                        ” with references to part 53. § 21.21 Notification of Failure To Comply or Existence of a Defect and Its Evaluation This proposed rule would revise § 21.21, by incorporating references to part 53, to update the requirements for notifying the Commission of a failure to comply or defect in paragraphs (a)(3) and (d)(1). § 21.51 Maintenance and Inspection of Records This proposed rule would revise § 21.51(a)(4) and (5) to apply to applicants for standard DC and applicants or holders of a standard design approval under part 53, in addition to part 52, regarding the retention of records.
                        ( print page 86991) § 21.61 Failure To Notify This proposed rule would revise § 21.61(b) to include references to part 53 licensees and applicants regarding failure to provide the notice required in § 21.21. § 25.5 Definitions This proposed rule would update the definition of “
                        *License*
                        ” to include those issued under part 53. § 25.17 Approval for Processing Applicants for Access Authorization This proposed rule would revise § 25.17(a) to add a reference to part 53 regarding AAs for individuals who need access to classified information in connection with activities under part 53. § 25.35 Classified Visits This proposed rule would update § 25.35(a) to apply the requirements for classified visits to licensees, certificate holders, and applicants under part 53 in addition to part 52. § 26.3 Scope This proposed rule would amend § 26.3 by revising paragraph (d) and adding new paragraph (f) which would establish the phase of construction or operation by which applicants and licensees under part 53 would be required to comply with subpart M of part 26, or all of the requirements of part 26 except subparts K and M. § 26.4 FFD Program Applicability to Categories of Individuals This proposed rule would revise paragraphs (a), (b), (c), (e), (f), (g), and (h) of § 26.4 to include references to part 53 and provisions for implementing an FFD program under subpart M. § 26.5 Definitions This proposed rule would amend § 26.5 by adding definitions for “
                        *Biological marker,*
                        ” “
                        *Change,*
                        ” “
                        *Illicit substance,*
                        ” “
                        *Reduction in FFD program effectiveness,*
                        ” and “
                        *Special Nuclear Material.*
                        ” It would also revise definitions of “
                        *Constructing or construction activities,*
                        ” “
                        *Contractor/vendor (C/V),*
                        ” “
                        *Other entity,*
                        ” “
                        *Questionable validity,*
                        ” “
                        *Reviewing official,*
                        ” “
                        *Safety-related structures, systems, and components (SSCs),*
                        ” “
                        *Security-related SSCs,*
                        ” and “
                        *Unit outage*
                        ” within this section. § 26.8 Information Collection Requirements: OMB Approval This proposed rule would revise § 26.8(b) with the new information collection requirements contained in proposed §§ 26.202, 26.603, 26.604, 26.605, 26.606, 26.607, 26.608, 26.609, 26.611, 26.613, 26.617, and 26.619. § 26.21 Fitness-for-Duty Program This proposed rule would revise § 26.21 to include a reference to § 26.3(f). § 26.51 Applicability This proposed rule would revise § 26.51 to extend the requirements of subpart C of part 26 to licensees and other entities identified in § 26.3(f) that do not implement the requirements of subpart M of part 26, as well as licensees and other entities that implement the requirements of § 26.605. § 26.53 General Provisions This proposed rule would revise § 26.53 paragraphs (e), (g), (h), and (i) to include references to § 26.3(f). § 26.63 Suitable Inquiry This proposed rule would revise § 26.63(d) with a reference to § 26.3(f). § 26.73 Applicability This proposed rule would revise § 26.73 to extend the requirements of subpart D of part 26 to licensees and other entities identified in § 26.3(f) that do not implement the requirements of subpart M of part 26, as well as licensees and other entities that implement the requirements of § 26.605(b). § 26.81 Purpose and Applicability This proposed rule would revise § 26.81 to extend the requirements of subpart E of part 26 to licensees and other entities identified in § 26.3(f) that do not implement the requirements of subpart M of part 26, as well as licensees and other entities that implement the requirements of § 26.605. § 26.201 Applicability This proposed rule would revise § 26.201 to include references to the proposed provisions in §§ 26.3(f) and 26.202, as well as revise the applicability of requirements in subpart I of part 26. § 26.202 General Provisions for Facilities Licensed Under Part 53 This proposed rule would add new § 26.202, which would require applicable licensees under part 53 to incorporate a policy for fatigue management into their FFD program in accordance with the provisions of this section. § 26.205 Work Hours This proposed rule would revise paragraphs (d)(7)(iii) and (d)(8) of § 26.205 to incorporate references to §§ 26.606 and 26.202(a) and (b). § 26.207 Waivers and Exceptions This proposed rule would revise § 26.207(a)(1)(ii) to include references to §§ 26.608 and 26.202(c) and to include provisions for implementing certain face-to-face supervisor assessments using electronic communications. § 26.211 Fatigue Assessments This proposed rule would revise § 26.211, paragraphs (a)(1), (a)(3), and (b) to incorporate references to §§ 26.202(c), 26.607(b), 26.608, and 26.619 and to include provisions for implementing certain face-to-face assessments using electronic communications. Subpart M—Fitness for Duty Programs for Facilities Licensed Under Part 53 This proposed rule would add new Subpart M of part 26 containing §§ 26.601, 26.603, 26.604 through 26.611, 26.613, 26.615, 26.617, and 26.619, which adds an optional technology-inclusive, risk-informed, and performance-based approach for the application of drug and alcohol testing and fatigue management requirements for facilities licensed under part 53. § 26.601 Applicability This proposed rule would add § 26.601, which would allow a licensee or other entity in § 26.3(f) to establish an FFD program in accordance with the requirements of subpart M of part 26. § 26.603 General Provisions This proposed rule would add § 26.603, which would establish the general requirements for implementing an FFD program under subpart M of part 26. § 26.604 FFD Program Requirements for Facilities That Satisfy the § 26.603(c) Criterion This proposed rule would add § 26.604, which would establish the FFD program elements for a licensee or other entity whose facilities and operations demonstrate compliance with the criterion in § 26.603(c). § 26.605 FFD Program Requirements for Facilities That Do Not Implement § 26.604 This proposed rule would add § 26.605, which would establish the FFD program elements for a licensee or other entity that does not demonstrate compliance with the criterion in § 26.603(c), or otherwise chooses to maintain an FFD program under this section.
                        ( print page 86992) § 26.606 Written Policies and Procedures This proposed rule would add § 26.606, which would require licensees and other entities that implement an FFD program under subpart M of part 26 to develop a written FFD policy statement and provide it to all individuals subject to the FFD program, and to establish, implement, and maintain written procedures addressing the topics outlined in this section. § 26.607 Drug and Alcohol Testing This proposed rule would add § 26.607, which would establish requirements for licensees and other entities performing drug and alcohol testing as part of an FFD program under subpart M of part 26. § 26.608 FFD Program Training This proposed rule would add § 26.608, which would require individuals who are subject to the FFD program under subpart M of part 26 to receive periodic training on FFD policies and procedures, including their duties and responsibilities under the BOP. § 26.609 Behavioral Observation This proposed rule would add § 26.609, which would establish the requirements for a BOP under subpart M of part 26. § 26.610 Sanctions This proposed rule would add § 26.610, which would require licensees and other entities implementing an FFD program under subpart M of part 26 to establish sanctions for FFD policy violations. § 26.611 Protection of Information This proposed rule would add § 26.611, which would require licensees and other entities implementing an FFD program under subpart M of part 26 to establish a system to protect personal information against unauthorized disclosure. § 26.613 Appeals Process This proposed rule would add § 26.613, which would require licensees and other entities that implement an FFD program under subpart M of part 26 to establish procedures for an individual to appeal a policy violation determination. § 26.615 Audits This proposed rule would add § 26.615, which would establish provisions for licensees and other entities that implement an FFD program under subpart M of part 26 to conduct audits to monitor the effectiveness of FFD program elements. § 26.617 Recordkeeping and Reporting This proposed rule would add § 26.617, which would require licensees or other entities implementing an FFD program under subpart M of part 26 to retain records pertaining to the administration of the program and to make reports in accordance with the requirements of this section. § 26.619 Suitability and Fitness Determinations This proposed rule would add § 26.619, which would require licensees and other entities that implement FFD programs to develop, implement, and maintain procedures to assess whether individuals are fit to perform the duties that make them subject to the FFD program. § 26.709 Applicability This proposed rule would designate the current paragraph as new paragraph (a), and it would be revised to reference paragraphs (a) through (d) of § 26.3. It would also add paragraph (b) to § 26.709, which would extend the requirements of subpart N of part 26 to licensees and other entities identified in § 26.3(f) that do not implement the requirements of subpart M of part 26, as well as licensees and other entities that implement the requirements of § 26.605(b). § 26.711 General Provisions This proposed rule would revise § 26.711(c) and (d) to incorporate a reference to § 26.3(f). § 26.825 Criminal Penalties This proposed rule would revise § 26.825(b) to include a reference to the proposed § 26.601. § 30.4 Definitions This proposed rule would revise the definition for “
                        *Utilization facility*
                        ” in § 30.4 to include utilization facilities defined in the regulations under part 53 in addition to part 50. § 30.50 Reporting Requirements This proposed rule would revise § 30.50(c)(3) to include references to part 53 in addition to part 50. § 40.60 Reporting Requirements This proposed rule would revise § 40.60(c)(3) to include references to part 53 in addition to part 50 regarding reporting requirements. § 50.47 Emergency Plans This proposed rule would revise § 50.47(a)(1) and (e) with appropriate references to part 53. § 50.54 Conditions of Licenses This proposed rule would revise § 50.54(q)(2), (q)(4), and (gg)(1) with appropriate references to part 53. § 50.160 Emergency Preparedness for Small Modular Reactors, Non-Light-Water Reactors, and Non-Power Production or Utilization Facilities This proposed rule would revise § 50.160(b)(3) and (c)(2) with the appropriate references to part 53. Appendix B to 10 CFR Part 50—Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants This proposed rule would revise appendix B to part 50 by revising the introduction and specific criteria to incorporate the appropriate references and terminology for part 53. § 51.20 Criteria for and Identification of Licensing and Regulatory Actions Requiring Environmental Impact Statements This proposed rule would revise § 51.20(b)(1) and (2) to require an EIS prior to the issuance of a CP, LWA, or ESP under part 53, or the issuance to renewal of a full power or design capacity license to operate a nuclear power reactor, testing facility, or fuel reprocessing plant under part 53. § 51.22 Criterion for Categorical Exclusion; Identification of Licensing and Regulatory Actions Eligible for Categorical Exclusion or Otherwise Not Requiring Environmental Review This proposed rule would revise § 51.22 to include corresponding references to part 53 in paragraphs (c)(3), (c)(9), (c)(12), (c)(17), (c)(22) and (23). § 51.26 Requirement To Publish Notice of Intent and Conduct Scoping Process This proposed rule would revise § 51.26(d) to add a reference to part 53. § 51.30 Environmental Assessment This proposed rule would revise the introductory text to paragraph (a) and revise paragraphs (d) and (e) of § 51.30 to incorporate the appropriate references to part 53 regarding EAs. § 51.31 Determinations Based on Environmental Assessment This proposed rule would revise § 51.31(a) to include a reference to part 53.
                        ( print page 86993) § 51.32 Finding of No Significant Impact This proposed rule would revise § 51.32(b)(1) and (3), finding there is no significant environmental impact associated with the issuance of standard DCs and MLs under part 53. § 51.49 Environmental Report-Limited Work Authorization This proposed rule would revise the introductory text of § 51.49(c) to require applicants for an ESP under part 53 requesting a LWA to include the environmental report required by § 51.50(b). § 51.50 Environmental Report—Construction Permit, Early Site Permit, or Combined License Stage This proposed rule would revise § 51.50, paragraphs (a), (b)(4), and the introductory text for paragraph (c) to incorporate the appropriate references to part 53. § 51.53 Postconstruction Environmental Reports This proposed rule would revise § 51.53(d) to include the appropriate references to part 53 regarding a license termination plan or decommissioning plan and related requirements for postconstruction environmental reports. § 51.54 Environmental Report—Manufacturing License This proposed rule would update § 51.54(a) to require applicants for MLs under part 53 to submit an environmental report with the application. § 51.55 Environmental Report—Standard Design Certification This proposed rule would update § 51.55(a) to require applicants for a standard DC under part 53 to submit an environmental report with the application. § 51.58 Environmental Report—Number of Copies; Distribution This proposed rule would revise § 51.58(b) to incorporate the appropriate references to part 53. § 51.77 Distribution of Draft Environmental Impact Statement This proposed rule would revise the introductory text for § 51.77(a) to add a reference to part 53. § 51.92 Supplement to the Final Environmental Impact Statement This proposed rule would revise § 51.92(b) to apply to COL applications referencing an ESP under part 53. § 51.95 Postconstruction Environmental Impact Statements This proposed rule would revise the introductory text for § 51.95(c) to include a reference to part 53 regarding the Commission's obligations to prepare an EIS following the renewal of an operating or COL for a nuclear plant under part 53. § 51.101 Limitations on Actions This proposed rule would revise § 51.101(a)(2) to include the corresponding references to part 53 where appropriate. § 51.103 Record of Decision—General This proposed rule would update § 51.103(a)(6) to apply to the issuance of a LWA in connection with a CP or COL under part 53. § 51.105 Public Hearings in Proceedings for Issuance of Construction Permits or Early Site Permits; Limited Work Authorizations This proposed rule would revise § 51.105(c)(1) to include the appropriate reference to LWAs under part 53 for CPs or ESPs. § 51.107 Public Hearings in Proceedings for Issuance of Combined Licenses; Limited Work Authorizations This proposed rule would amend § 51.107 by revising the introductory text for paragraphs (a) and (b) and updating paragraph (d)(1) to include the appropriate corresponding references to part 53. § 51.108 Public Hearings on Commission Findings That Inspections, Tests, Analyses, and Acceptance Criteria of Combined Licenses Are Met This proposed rule would revise § 51.108 to incorporate the appropriate references to part 53. 10 CFR part 53—Risk-Informed, Technology-Inclusive Regulatory Framework for Commercial Nuclear Plants This proposed rule would add a new part to 10 CFR Chapter I, designated as Part 53 including §§ 53.000 through 53.9010. § 53.000 Purpose This proposed rule would add § 53.000 which provides an optional technology-inclusive, performance-based framework for the issuance, amendment, renewal, and termination of licenses, permits, certifications, and approvals for commercial nuclear plants licensed under section 103 of the Atomic Energy Act of 1954, as amended. Subpart A—General Provisions This proposed rule would add subpart A, to establish a set of general provisions, which apply to all applicants and licensees under part 53. § 53.015 Scope This proposed rule would add § 53.015, which would extend the provisions of subpart A to all applicants and licensees under part 53. § 53.020 Definitions This proposed rule would add § 53.020, which would define key terms in part 53. § 53.040 Written Communications This proposed rule would add § 53.040, which would govern how applicants and licensees submit written communications to the NRC, including applications, submissions related to the security plans, emergency plan, and quality assurance, certifications of permanent cessation of operations and permanent fuel removal, and other submittals required under part 53. § 53.050 Deliberate Misconduct This proposed rule would add § 53.050, which would prohibit licensees or applicants, contractors and subcontractors, or employees of those entities from deliberately violating NRC rules, regulations, or orders, or the terms, conditions, and limitations of a part 53 license. This proposed rule would also prohibit deliberate submissions of incomplete or inaccurate information. Violations would be subject to enforcement actions under subpart B of part 2. § 53.060 Employee Protection This proposed rule would add § 53.060, which would prohibit applicants and licensees from discriminating against employees for engaging in the protected activities listed in this section and provide remedial procedures for employees who believe they are the subjects of discrimination. § 53.070 Completeness and Accuracy of Information This proposed rule would add § 53.070, which would require licensees and applicants under part 53 to provide complete and accurate information in accordance with all applicable laws, Commission regulations, and the terms and conditions of their license. This proposed rule would also require licensees to notify the Commission within two days of identifying information with material implications 
                        ( print page 86994)
                        for public health and safety or common defense and security. § 53.080 Specific Exemptions This proposed rule would add § 53.080, which would establish the special circumstances under which the Commission could grant exemptions to part 53 licensees and the Commission's criteria for making such a determination. § 53.090 Standards for Review This proposed rule would add § 53.090 to establish the standards that the Commission would consider when determining whether to issue a permit or license under part 53. § 53.100 Jurisdictional Limits This proposed rule would add § 53.100, which would provide that permits, licenses, standard design approvals, and standard DCs are solely issued for activities within the jurisdiction of the United States. § 53.110 Attacks and Destructive Acts This proposed rule would add § 53.110, which would exempt licensees or applicants under part 53 from providing design features to protect against attacks or destructive acts directed at the facility by United States adversaries. § 53.115 Rights Related to Special Nuclear Material This proposed rule would add § 53.115, which would establish provisions regarding the rights to SNM under a part 53 license. § 53.117 License Suspension and Rights of Recapture This proposed rule would add § 53.117, which would provide that the Commission may suspend licenses and recapture material or control of a facility in a state of war or national emergency declared by Congress. § 53.120 Information Collection Requirements: OMB Approval This proposed rule would add § 53.120, which would establish requirements for information collection requirements and Office of Budget and Management approval. Subpart B—Technology-Inclusive Safety Requirements This proposed rule would add subpart B, to establish a set of technology-inclusive performance standards that would be used throughout part 53 to determine appropriate regulatory controls for SSCs, human actions, and programs. § 53.210 Safety Criteria for Design-Basis Accidents This proposed rule would add § 53.210 to set dose values to ensure that plants are designed to limit the public's radiation exposure in the event of a DBA. § 53.220 Safety Criteria for Licensing-Basis Events Other Than Design-Basis Accidents This proposed rule would add § 53.220 to require plants to implement a combination of design features and programmatic controls to control risks to the public in the event of a LBE other than a DBA. § 53.230 Safety Functions This proposed rule would add § 53.230, which specifies that limiting the release of radioactive materials from the facility is the primary safety function of a commercial nuclear plant, and that additional safety functions must be defined to support the retention of radioactive materials during LBEs. § 53.240 Licensing-Basis Events This proposed rule would add § 53.240 to require commercial nuclear plants to conduct an analysis of LBEs to confirm that design features and programmatic controls satisfy the safety criteria under §§ 53.210 and 53.220, or alternatively, under § 53.470. § 53.250 Defense in Depth This proposed rule would add § 53.250 to establish a performance-based, defense-in-depth approach to address uncertainties about the effectiveness and reliability of plant SSCs, personnel, and programmatic controls. § 53.260 Normal Operations This proposed rule would add § 53.260, requiring holders of licenses to operate commercial nuclear plants to control public doses and dose rates in unrestricted areas to meet the requirements in part 20, during normal plant operation. § 53.270 Protection of Plant Workers This proposed rule would add § 53.270, requiring holders of licenses to operate commercial nuclear plants to control occupational doses to meet the requirements in part 20. Subpart C—Design and Analysis Requirements This proposed rule would add subpart C, which requires the implementation of certain design features and the performance of risk assessments and analyses to demonstrate compliance with the safety criteria and safety functions in subpart B. § 53.400 Design Features for Licensing-Basis Events This proposed rule would add § 53.400, which would require design features that satisfy the safety criteria defined in § 53.210 and § 53.220 or § 53.470 and fulfill the safety functions identified in § 53.230 during LBEs. § 53.410 Functional Design Criteria for Design-Basis Accidents This proposed rule would add § 53.410, which would stipulate that functional design criteria must be defined for each design feature required by § 53.400 to demonstrate compliance with the safety criteria defined in § 53.210 for DBAs. § 53.415 Protection Against External Hazards This proposed rule would add § 53.415, which would require SR SSCs to be designed to withstand the effects of natural phenomena and constructed hazards while performing the intended safety functions. § 53.420 Functional Design Criteria for Licensing-Basis Events Other Than Design-Basis Accidents This proposed rule would add § 53.420, which would require functional design criteria to be defined for each design feature required by § 53.400 to demonstrate compliance with the safety criteria defined in § 53.220 for LBEs other than DBAs. § 53.425 Design Features and Functional Design Criteria for Normal Operations This proposed rule would add § 53.425, which would require commercial nuclear plants to implement design features and define functional design criteria sufficient to demonstrate compliance with § 53.850 and show through functional design criteria that design features and corresponding programmatic controls control wastes, as required under part 20. § 53.430 Design Features and Functional Design Criteria for Protection of Plant Workers This proposed rule would add § 53.430, which would require commercial nuclear plants to implement design features and define functional design criteria sufficient to demonstrate compliance with § 53.270. § 53.440 Design Requirements This proposed rule would add § 53.440, which would establish various 
                        ( print page 86995)
                        design feature requirements, including protection against fires and explosions, criticality accidents, and the impact of a large commercial aircraft. § 53.450 Analysis Requirements This proposed rule would add § 53.450, which would require commercial nuclear plants to perform PRAs in combination with other analytical methods to identify and assess risks and determine compliance with the safety criteria in subpart B. In addition, § 53.450 would require analysis of DBAs and other analyses to assess the adequacy of protections against fire, aircraft impact, and the release of effluents. § 53.460 Safety Categorization and Special Treatments This proposed rule would add § 53.460 to address the safety classification of SSCs and determine appropriate special treatments. § 53.470 Maintaining Analytical Safety Margins Used To Justify Operational Flexibilities This proposed rule would add § 53.470 to permit applicants and licensees to implement more restrictive criteria than that defined in §§ 53.220 and 53.450(e) to support operational flexibilities. § 53.480 Earthquake Engineering This proposed rule would add § 53.480 to provide overall seismic design considerations based on the safety criteria in subpart B and siting requirements in subpart D to ensure that SSCs are able to withstand the effects of earthquakes without loss of capability to fulfill safety functions. Subpart D—Siting Requirements This proposed rule would add subpart D, which would address requirements associated with the siting of commercial nuclear facilities under part 53, including considerations of external hazards and potential adverse impacts on the surrounding population. § 53.500 General Siting and Siting Assessment This proposed rule would add § 53.500, which would require a siting assessment for each commercial nuclear plant to ensure that design features and programmatic controls are sufficient to address LBEs and mitigate potential adverse impacts of the plant on the surrounding environs. § 53.510 External Hazards This proposed rule would add § 53.510, which would require site-specific assessments, including an evaluation of geological and seismic siting factors, to identify and characterize the external hazard level for a range of natural and constructed hazards. § 53.520 Site Characteristics This proposed rule would add § 53.520, which would require the design and analyses conducted under subpart C to consider how site characteristics may contribute to LBEs. § 53.530 Population-Related Considerations This proposed rule would add § 53.530, which would establish requirements related to the facility's exclusion area, low-population zone, and population center distance. § 53.540 Siting Interfaces This proposed rule would add § 53.540, which would require that external hazards and site characteristics must be accounted for in the design features, programmatic controls, and supporting analyses used to demonstrate compliance with the safety criteria in §§ 53.210 and 53.220. Subpart E—Construction and Manufacturing Requirements This proposed rule would add subpart E, which would establish requirements for the construction and manufacture of commercial nuclear plants. § 53.600 Construction and Manufacturing—Scope and Purpose This proposed rule would add § 53.600, which would indicate that this subpart applies to construction and manufacturing activities authorized by a CP, COL, ML, or LWA issued under this part. § 53.605 Reporting of Defects and Noncompliance This proposed rule would add § 53.605, which would describe the procedures, notification requirements, and records retention requirements that each CP, ML, and COL is subject to with respect to reporting of defects and noncompliance. § 53.610 Construction This proposed rule adds § 53.610 to address the management and control of the construction of a commercial nuclear plant, including specific requirements for procedures and quality assurance, control of radioactive materials, and post construction inspections. § 53.620 Manufacturing This proposed rule would add § 53.620, which would ensure that the holders of an ML under part 53 develop plans, programs, and organizational units to manage and control manufacturing activities, and would establish requirements for the loading of fuel into a manufactured reactor for subsequent transport to a commercial nuclear plant and operation pursuant to a COL. Subpart F—Requirements for Operation This proposed rule would add subpart F, which would establish regulatory requirements to ensure that the safety criteria in subpart B are satisfied whenever a commercial nuclear plant licensed under part 53 is operational. This includes periods of normal operation and unplanned events. § 53.700 Operational Objectives This proposed rule would add § 53.700, which would establish general operational objectives to ensure that licensees under part 53 have implemented and maintained the SSCs necessary to demonstrate compliance with the safety functions identified in subpart B for addressing normal operations and responding to LBEs. § 53.710 Maintaining Capabilities and Availability of Structures, Systems, and Components This proposed rule would add § 53.710, which would require licensees under part 53 to demonstrate compliance with the safety criteria in subpart B by establishing TS for all SR SSCs and developing documents and procedures for all NSRSS SSCs. § 53.715 Maintenance, Repair, and Inspection Programs This proposed rule would add § 53.715, which would require licensees to develop, implement, and maintain programs to assess and manage any risks posed by maintenance activities and to evaluate the efficacy of performance, condition monitoring, and maintenance activities. § 53.720 Response to Seismic Events This proposed rule would add § 53.720, which would establish requirements for licensees to respond to a seismic event during the operating phase of the life cycle of a commercial nuclear plant. § 53.725 General Staffing, Training, Personnel Qualifications, and Human Factors Requirements This proposed rule would add § 53.725, which would provide an 
                        ( print page 86996)
                        overview of the staffing, training, personnel qualifications, and human factors requirements established in §§ 53.725 through 53.830 and would provide definitions of “
                        *Automation,*
                        ” “
                        *Auxiliary operator,*
                        ” “
                        *Controls,*
                        ” “
                        *Generally licensed reactor operator,*
                        ” “
                        *Load following,*
                        ” “
                        *Operator,*
                        ” “
                        *Performance testing,*
                        ” “
                        *Reference plant,*
                        ” “
                        *Self-reliant mitigation facility,*
                        ” “
                        *Senior operator,*
                        ” “
                        *Simulation facility,*
                        ” and “
                        *Systems approach to training.*
                        ” Proposed §§ 53.725 through 53.830 would apply to applicants for or holders of OLs or COLs under part 53. § 53.726 Communications This proposed rule would add § 53.726, which would contain communications requirements applicable to sections §§ 53.725 through 53.830. It also contains requirements to notify the Commission within 30 days should a specifically licensed operator or senior operator be reassigned, terminated, or suffer permanent disability or illness. § 53.728 Completeness and Accuracy of Information This proposed rule would add § 53.728, which would require submitted information to be complete and accurate in all material respects. § 53.730 Defining, Fulfilling, and Maintaining the Role of Personnel in Ensuring Safe Operations This proposed rule would add § 53.730, which would establish technical requirements for applicants or holders of OLs or COLs within the areas of HFE, human-system interface design, concept of operations, functional requirements analysis, function allocation, operating experience, procedures, staffing, operator training, operator examinations, and operator proficiency. § 53.735 General Exemptions This proposed rule would add § 53.735, which would establish general exemptions for licensed operators. § 53.740 Facility Licensee Requirements—General This proposed rule would add § 53.740, which would establish staffing requirements for interaction-dependent-mitigation facilities and self-reliant mitigation facilities. § 53.745 Operator License Requirements This proposed rule would add § 53.745, which would require individuals to be licensed to perform certain functions. § 53.760 Operator Licensing This proposed rule would add § 53.760, which would address the applicability of the requirements of §§ 53.760 through 53.795 for specifically licensed operators and senior operators. § 53.765 Medical Requirements This proposed rule would add § 53.765, which would establish medical requirements for specifically licensed operators and senior operators. § 53.770 Incapacitation Because of Disability or Illness This proposed rule would add § 53.770, which would establish requirements to address permanent medical conditions for specifically licensed operators and senior operators. § 53.775 Applications for Operators and Senior Operators This proposed rule would add § 53.775, which would establish the application process and requirements for individuals applying for specific operator and senior operator licenses. § 53.780 Training, Examination, and Proficiency Program This proposed rule would add § 53.780, which would contain the requirements associated with specifically licensed operator and senior operator initial training, initial examinations, requalification training, requalification examinations, examination integrity, simulation facilities, waivers, and proficiency. § 53.785 Conditions of Operator and Senior Operator Licenses This proposed rule would add § 53.785, which would establish conditions for specific operator and senior operator licenses. § 53.790 Issuance, Modification, and Revocation of Operator and Senior Operator Licenses This proposed rule would add § 53.790, which would contain requirements associated with the issuance, modification, or revocation of specific operator and senior operator licenses. § 53.795 Expiration and Renewal of Operator and Senior Operator Licenses This proposed rule would add § 53.795, which would contain requirements associated with the expiration and renewal of specific operator and senior operator licenses. § 53.800 Facility Licensees for Self-Reliant-Mitigation Facilities This proposed rule would add § 53.800, which would establish the technical criteria by which commercial nuclear plants under part 53 are determined to be of the self-reliant mitigation class of facilities that would be staffed by GLROs in lieu of specifically licensed operators and senior operators. § 53.805 Facility Licensee Requirements Related to Generally Licensed Reactor Operators This proposed rule would add § 53.805, which would establish requirements that apply to the facility licensee at those facilities staffed by GLROs. § 53.810 Generally Licensed Reactor Operators This proposed rule would add § 53.810, which would issue and describe the general license for GLROs that manipulate the controls of a self-reliant mitigation facility. § 53.815 Generally Licensed Reactor Operator Training, Examination, and Proficiency Programs This proposed rule would add § 53.815, which would contain the requirements for GLRO initial training, initial examinations, continuing training, requalification examinations, examination integrity, simulation facilities, examination waivers, and proficiency. § 53.820 Cessation of Individual Applicability This proposed rule would add § 53.820, which would address the requirements by which the general license for GLROs would cease to be applicable on an individual basis. § 53.830 Training and Qualification of Commercial Nuclear Plant Personnel This proposed rule would add § 53.830, which would address training and qualification requirements for supervisors, technicians, and other appropriate operating personnel at commercial nuclear plants. § 53.845 Programs This proposed rule would add § 53.845, which would require licensees under part 53 to establish programs that include, but are not limited to, radiation protection, emergency preparedness, security, quality assurance, integrity assessment, fire protection, ISI and IST, and facility safety, to ensure that the safety criteria and functions in subpart 
                        ( print page 86997)
                        B are maintained during normal operations and LBEs. § 53.850 Radiation Protection This proposed rule would add § 53.850, which would require licensees under part 53 to implement and maintain programs and processes to limit and monitor radioactive plant effluents and limit the exposure of plant personnel and the public. § 53.855 Emergency Preparedness This proposed rule would add § 53.855, which would require licensees under this part to have an emergency response plan for radiological emergencies. § 53.860 Security Programs This proposed rule would add § 53.860, which would require licensees under part 53 to develop, implement, and maintain programs for physical security, FFD, AA, cybersecurity, and information security. § 53.865 Quality Assurance This proposed rule would add § 53.865, which would require licensees under part 53 to establish a quality assurance program that includes a written manual to ensure activities are conducted in accordance with codes and standards found acceptable by the NRC. § 53.870 Integrity Assessment Programs This proposed rule would add § 53.870, which would require licensees under part 53 to establish an integrity assessment program to ensure that the plant continues to fulfill safety criteria and functional design criteria as it ages. § 53.875 Fire Protection This proposed rule would add § 53.875, which would require licensees under part 53 to establish a fire protection plan and describe the necessary elements that the plan must incorporate. § 53.880 Inservice Inspection and Inservice Testing This proposed rule would add § 53.880, which would require licensees under part 53 to develop and implement a program for ISI and IST in accordance with the requirements of this section. § 53.910 Procedures and Guidelines This proposed rule would add § 53.910, which would require licensees under part 53 to develop, maintain, and implement procedures and guidelines that address normal plant operations and responses to unplanned events. Subpart G—Decommissioning Requirements This proposed rule would add subpart G, to establish decommissioning requirements for applicants for or holders of an OL or COL under part 53. § 53.1000 Scope and Purpose This proposed rule would add § 53.1000, which would establish the scope of the decommissioning requirements for applicants and licensees under part 53 and describe the contents of subpart G of part 53. § 53.1010 Financial Assurance for Decommissioning This proposed rule would add § 53.1010, which would establish the requirement that applicants for an OL or COL under part 53 provide reasonable assurance that funds will be available for the decommissioning process. This section would describe the requirements associated with the required plan and an associated decommissioning report that ensures and documents that adequate funding for decommissioning will be available. § 53.1020 Cost Estimates for Decommissioning This proposed rule would add § 53.1020, which would require site-specific cost estimates for decommissioning and establish the aspects that must be included in the estimate. § 53.1030 Annual Adjustments to Cost Estimates for Decommissioning This proposed rule would add § 53.1030, which would require that holders of an OL or COL under part 53 annually adjust their cost estimate for decommissioning to account for escalation in labor, energy, and waste burial costs. This section would allow licensees to elect either a site-specific adjustment factor or a generic adjustment factor. § 53.1040 Methods for Providing Financial Assurance for Decommissioning This proposed rule would add § 53.1040, which would establish suitable methods that holders of an OL or COL under part 53 may use to provide financial assurance for decommissioning to the NRC. § 53.1045 Limitations on the Use of Decommissioning Trust Funds This proposed rule would add § 53.1045, which would establish requirements for decommissioning trust funds under part 53, including criteria for using decommissioning trust funds and required terms. § 53.1050 NRC Oversight This proposed rule would add § 53.1050, which would outline the steps the NRC may take to ensure adequate accumulation of decommissioning funds. § 53.1060 Reporting and Recordkeeping Requirements This proposed rule would add § 53.1060, which would contain reporting and recordkeeping requirements related to decommissioning for each holder of an OL or COL under part 53. This section would outline requirements for documents such as: certification of decommissioning funding, decommissioning cost estimates and copies of financial instruments, licensee records of information important to safe and effective decommissioning, post-shutdown decommissioning activities report, financial assurance reports, and reports on the status of funding for managing irradiated fuel. § 53.1070 Termination of License This proposed rule would add § 53.1070, which would establish procedures for decommissioning and license termination applicable to licensees under part 53 that have determined to permanently cease operations. § 53.1075 Program Requirements During Decommissioning This proposed rule would add § 53.1075, which would require licensees under part 53 to establish and maintain a decommissioning fire protection program to prevent, detect, and control fires, and ensure that the risk of fire induced radiological hazards are minimized through the various stages of facility decommissioning. § 53.1080 Release of Part of a Commercial Nuclear Plant or Site for Unrestricted Use This proposed rule would add § 53.1080, which would establish licensee procedures for requesting and NRC procedures for approving partial release of a commercial nuclear plant or site for unrestricted use prior to receiving approval of a license termination plan from the Commission under part 53. Subpart H—Licenses, Certifications, and Approvals This proposed rule would add subpart H, which would govern the process of applying for, amending, renewing, or 
                        ( print page 86998)
                        terminating a LWA, ESP, standard design approval, standard DC, ML, CP, OL, or COL under part 53. § 53.1100 Filling of Application for Licenses, Certifications, or Approvals; Oath or Affirmation This proposed rule would add § 53.1100, which would establish requirements for applicants seeking a standard design approval, standard DC, license, or permit under part 53 to submit an application. § 53.1101 Requirement for License This proposed rule would add § 53.1101, which would prohibit any use of a utilization facility except as authorized by a license issued by the NRC or by an exception as described in § 53.1120. § 53.1103 Combining Applications and Licenses This proposed rule would add § 53.1103, which would permit applicants under part 53 seeking multiple licenses to submit a single application, and the Commission to issue a single license for activities that would otherwise be licensed separately. § 53.1106 Elimination of Repetition This proposed rule would add § 53.1106, which would allow applicants under part 53 to reference information contained in previous documents filed with the Commission so long as those references are clear and specific. § 53.1109 Contents of Applications; General Information This proposed rule would add § 53.1109, which would establish the general content to be included in applications made under part 53, including but not limited to the identifying information of the applicant and the radiological emergency response plans of government entities within the plume exposure pathway EPZ. § 53.1112 Environmental Conditions This proposed rule would add § 53.1112, which would allow the Commission to attach conditions to CPs, ESPs, and licenses issued under part 53 to address environmental issues during construction, operation, or decommissioning. These conditions will be derived from the information contained in the environmental report submitted as part of the application for a permit or license. § 53.1115 Agreement Limiting Access to Classified Information This proposed rule would add § 53.1115, which would require applicants to agree in writing, prior to receiving a license or standard design approval under part 53, to restrict any facilities, or any individuals with access to plant facilities, from possessing Restricted Data or classified National Security Information until they have received the appropriate authorization. § 53.1118 Ineligibility of Certain Applicants This proposed rule would add § 53.1118, which would prevent citizens, nationals, or agents of a foreign country or corporations owned, controlled, or dominated by a foreign entity from applying for or obtaining a license under part 53. § 53.1120 Exceptions and Exemptions From Licensing Requirements This proposed rule would add § 53.1120, which would establish the activities that are exempt from licensing requirements. § 53.1121 Public Inspection of Applications This proposed rule would add § 53.1121, which would allow applicant submissions to be made publicly available under the provisions of part 2. § 53.1124 Relationship Between Sections This proposed rule would add § 53.1124, which would outline the relationship between LWAs, ESPs, standard design approvals, standard DCs, MLs, CPs, OLs, and COLs under part 53. § 53.1130 Limited Work Authorizations This proposed rule would add § 53.1130, which would establish requirements for requesting an LWA and grounds for the Commission to issue an LWA. It would also contain details about the effect of an LWA and the implementation of a redress plan. § 53.1140 Early Site Permits This proposed rule would add § 53.1140, which would provide an overview of the requirements regarding applications for and the issuance of ESPs under part 53. § 53.1143 Filing of Applications This proposed rule would add § 53.1143, which would enable an applicant under part 53 to apply for an ESP, regardless of whether they have filed an application for a CP or COL for that site. § 53.1144 Contents of Applications for Early Site Permits; General Information This proposed rule would add § 53.1144, which would require applications for ESPs to include the information required by § 53.1109(a) through (d) and (j). § 53.1146 Contents of Applications for Early Site Permits; Technical Information This proposed rule would add § 53.1146, which would require applicants for ESPs to submit technical information, including but not limited to a Site Safety Analysis Report and emergency plans. § 53.1149 Review of Applications This proposed rule would add § 53.1149, which would establish standards for review of applications for ESPs under part 53, including requirements for the Commission to prepare an EIS and assess the adequacy of protective actions in the event of a radiological emergency. It would also require the administrative review of applications and hearings to follow the procedural requirements of part 2. § 53.1155 Referral to the Advisory Committee on Reactor Safeguards This proposed rule would add § 53.1115, which would require the ACRS to review SR content in the application for an ESP under part 53. § 53.1158 Issuance of Early Site Permit This proposed rule would add § 53.1158, which would establish the conditions under which the Commission may issue an ESP under part 53, as well as the information, terms, and conditions to be included in the permit. § 53.1161 Extent of Activities Permitted This proposed rule would add § 53.1161, which would require that a valid ESP only be used for the purpose of site redress, unless the site is referenced in an application for a CP or COL under part 53. § 53.1164 Duration of Permit This proposed rule would add § 53.1164, which would govern the conditions under which an ESP remains valid following the date of issuance. § 53.1167 Limited Work Authorization After Issuance of Early Site Permit This proposed rule would add § 53.1167, which would permit the 
                        ( print page 86999)
                        holder of an ESP to request a LWA under § 53.1130. § 53.1170 Transfer of Early Site Permit This proposed rule would add § 53.1170, which would govern the transfer of an ESP in accordance with § 53.1570. § 53.1173 Application for Renewal This proposed rule would add § 53.1173, which would establish the conditions and procedures for renewing an ESP under part 53. § 53.1176 Criteria for Renewal This proposed rule would add § 53.1176, which would establish the criteria that the Commission may use to grant a renewal of an ESP under part 53. § 53.1179 Duration of Renewal This proposed rule would add § 53.1179, which would govern the duration of a renewed ESP under part 53. § 53.1182 Use of Site for Other Purposes This proposed rule would add § 53.1182, which would govern acceptable uses of the site for purposes other than those described in the permit. § 53.1188 Finality of Early Site Permit Determinations This proposed rule would add § 53.1188, which would address the finality of ESP determinations under part 53. § 53.1200 Standard Design Approvals This proposed rule would add § 53.1200, which would address the procedures for filing an application for a standard design approval under part 53, the process of review by NRC staff, and referral to the ACRS of standard designs. § 53.1203 Filing of Applications This proposed rule would add § 53.1203, which would enable applicants to submit a final design for the entire facility, or major portions, to the NRC staff for review. § 53.1206 Contents of Applications for Standard Design Approvals; General Information This proposed rule would add § 53.1206, which would require applications for a standard design approval under part 53 to contain the information required by § 53.1109(a) through (c) and (j). § 53.1209 Contents of Applications for Standard Design Approvals; Technical Information This proposed rule would add § 53.1209, which would require the inclusion of certain technical information, including a FSAR, site parameters, and design information, when an applicant seeks review of major portions of a standard design. § 53.1210 Contents of Applications for Standard Design Approvals; Other Application Content This proposed rule would add § 53.1210, which would require applications for standard design approvals under part 53 to include a description of the availability controls used to satisfy the safety criteria of § 53.220, the program to protect Safeguards Information against unauthorized disclosure, evidence that safety questions associated with SSCs have been resolved, and a description of how design features fulfill design criteria. § 53.1212 Standards for Review of Applications This proposed rule would add § 53.1212, which would require applications for standard design approval to be reviewed under the standards in parts 20, 53, and 73. § 53.1215 Referral to the Advisory Committee on Reactor Safeguards This proposed rule would add § 53.1215, which would require the ACRS to report on any portions of the application for a standard design approval under part 53 concerning safety. § 53.1218 Staff Approval of Design This proposed rule would add § 53.1218, which would require the NRC staff to make a determination on the acceptability of the design, publish its decision in the 
                        **Federal Register**
                        , and issue a report analyzing the design that is available at 
                        *http://nrc.gov.*
                         Additionally, the rule would establish the conditions under which a design approval under part 53 remains valid. § 53.1221 Finality of Standard Design Approvals; Information Requests This proposed rule would add § 53.1221, which would require NRC staff and the ACRS to rely upon an approved design in their review of any standard DC, ML, or individual facility license application under part 53 that references the standard design approval. The proposed rule would also govern requirements for issuing information requests. § 53.1230 Standard Design Certifications This proposed rule would add § 53.1230, which would provide an overview of the requirements and procedures that govern the issuance of standard DCs under part 53. § 53.1233 Filing of Applications This proposed rule would add § 53.1233, which would enable an application for DC to be filed, regardless of whether an application for a CP, COL, or ML has been filed, provided it complies with the filing requirements in § 53.040 and §§ 2.811 through 2.819. § 53.1236 Contents of Applications for Standard Design Certifications; General Information This proposed rule would add § 53.1236, which would require an application for a standard DC under part 53 to contain all of the information required by § 53.1109(a) through (c) and (j). § 53.1239 Contents of Applications for Standard Design Certifications; Technical Information This proposed rule would add § 53.1239, which would require applicants for a standard DC under part 53 to submit a FSAR that includes technical design information at a level of detail sufficient to enable the Commission to make a safety determination. § 53.1241 Contents of Applications for Standard Design Certifications; Other Application Content This proposed rule would add § 53.1241, which would require applications for standard DCs under part 53 to include an environmental report, as well as a description of the availability controls used to satisfy the safety criteria of § 53.220, proposed ITAAC, the program to protect Safeguards Information against unauthorized disclosure, evidence that safety questions associated with SSCs have been resolved, and a description of how design features fulfill design criteria. § 53.1242 Review of Applications This proposed rule would add § 53.1242, which would require applications for standard DCs to be reviewed for compliance with the standards in parts 20, 51, 53, and 73. It would also establish procedural requirements for reviewing applications and holding hearings in accordance with subpart H of part 2.
                        ( print page 87000) § 53.1245 Referral to the Advisory Committee on Reactor Safeguards This proposed rule would add § 53.1245, which would require the ACRS to report on any portions of the application for a standard DC under part 53 concerning safety. § 53.1248 Issuance of Standard Design Certification This proposed rule would add § 53.1248, which would establish the conditions under which the Commission may issue a DC rule that specifies the site parameters, design characteristics, and any additional terms and conditions of the DC rule. § 53.1251 Duration of Certification This proposed rule would add § 53.1251, which would set the conditions under which a standard DC remains valid. § 53.1254 Application for Renewal This proposed rule would add § 53.1254, which would establish the conditions and procedures for renewing a standard DC under part 53. § 53.1257 Criteria for Renewal This proposed rule would add § 53.1257, which would enable the Commission to issue a rule granting the renewal of a standard DC under part 53, impose additional requirements, and grant amendment requests. § 53.1260 Duration of Renewal This proposed rule would add § 53.1260, which would provide that a renewal of a standard DC under part 53 is valid for not less than 10 years, nor more than 15 years. § 53.1263 Finality of Standard Design Certifications This proposed rule would add § 53.1263, which would establish limited conditions under which the Commission may initiate a rulemaking to modify, rescind, or impose new requirements on a standard DC rule under part 53. It would also address requests for an exemption from elements of the certification information, and require that applicants for a CP, COL, or ML that references a DC rule make information normally contained in engineering documents available for audit. § 53.1270 Manufacturing Licenses This proposed rule would add § 53.1270, which would provide an overview of the requirements and procedures for applying for and issuing an ML under part 53. § 53.1273 Filing of Applications This proposed rule would add § 53.1273, which would establish the requirements to apply for an ML under part 53. § 53.1276 Contents of Applications for Manufacturing Licenses; General Information This proposed rule would add § 53.1276, which would require applicants for an ML under part 53 to include the information contained in § 53.1109(a) through (e) and (j). § 53.1279 Contents of Applications for Manufacturing Licenses; Technical Information This proposed rule would add § 53.1279, which would require an applicant for an ML under part 53 to include certain technical information in a FSAR, including but not limited to information about site parameters, design information, manufacturing information, and information related to the potential fueling and ultimate deployment of a completed manufactured reactor. § 53.1282 Contents of Applications for Manufacturing Licenses; Other Application Content This proposed rule would add § 53.1282, which would require applicants for an ML under part 53 to include in their application the proposed ITAAC, an environmental report, a description of the program to protect Safeguards Information against unauthorized disclosure, and a description of how design features fulfill design criteria. It would also include content requirements for the ITAAC and environmental reports in applications that reference a standard DC. § 53.1285 Review of Applications This proposed rule would add § 53.1285, which would require applications for MLs under part 53 to be reviewed for compliance with applicable standards and establish procedural requirements for reviewing applicants and holding hearings in accordance with part 2. § 53.1286 Referral to the Advisory Committee on Reactor Safeguards This proposed rule would add § 53.1286, which would require the ACRS to report on any portions of the application for an ML under part 53 concerning safety. § 53.1287 Issuance of Manufacturing Licenses This proposed rule would add § 53.1287, which would establish the conditions under which the Commission may issue an ML under part 53. § 53.1288 Finality of Manufacturing Licenses This proposed rule would add § 53.1288, which would address the limited circumstances in which the Commission may modify, rescind, or impose new requirements following the issuance of an ML under part 53. It would also address requests for a departure from the specifications of the license. § 53.1291 Duration of Manufacturing Licenses This proposed rule would add § 53.1291, which would govern the expiration of an ML, which is valid for no less than 5, nor more than 15 years from the date of issuance. § 53.1293 Transfer of Manufacturing Licenses This proposed rule would add § 53.1293, which would provide that an ML under part 53 may be transferred in accordance with § 53.1570. § 53.1295 Renewal of Manufacturing Licenses This proposed rule would add § 53.1295, which would establish the procedures for applicants to apply for and the Commission to grant a renewal of an ML under part 53. § 53.1300 Construction Permits This proposed rule would add § 53.1300, which would provide an overview of the requirements and procedures for applicants to apply for and the Commission to grant a CP under part 53. § 53.1306 Contents of Applications for Construction Permits; General Information This proposed rule would add § 53.1306, which would require applicants for a CP under part 53 to submit the general information required by § 53.1109, as well as financial information. § 53.1309 Contents of Applications for Construction Permits; Technical Information This proposed rule would add § 53.1309, which would require applicants for a CP under part 53 to submit a PSAR and a description of the program to protect Safeguards 
                        ( print page 87001)
                        Information from unauthorized disclosure. § 53.1312 Contents of Applications for Construction Permits; Other Application Content This proposed rule would add § 53.1312, which would require applicants for a CP under part 53 to submit an environmental report and to provide additional details in the PSAR if the application references an ESP, standard design approval, or standard DC. § 53.1315 Review of Applications This proposed rule would add § 53.1315, which would require applications for CPs under part 53 to be reviewed for compliance with applicable standards and establish procedural requirements for reviewing applications and holding hearings in accordance with part 2. § 53.1318 Finality of Referenced NRC Approvals, Permits, and Certifications This proposed rule would add § 53.1318, which would address the finality of ESPs, standard design approvals, and standard DCs referenced in the CP application. § 53.1324 Referral to the Advisory Committee on Reactor Safeguards This proposed rule would add § 53.1324, which would require the ACRS to report on any portions of the application for a CP under part 53 concerning safety. § 53.1327 Authorization To Conduct Limited Work Authorization Activities This proposed rule would add § 53.1327, which would govern authorization to conduct LWA activities. § 53.1330 Exemptions, Departures, and Variances This proposed rule would add § 53.1330, which would govern requests for and issuance of exemptions from the Commission's regulations and exemptions, departures, and variances from NRC approvals, permits, and certifications. § 53.1333 Issuance of Construction Permits This proposed rule would add § 53.1333, which would establish the conditions under which the Commission may issue CPs and accompanying terms and conditions under part 53. § 53.1336 Finality of Construction Permits This proposed rule would add § 53.1336, which would address the finality of CPs. § 53.1342 Duration of Construction Permits This proposed rule would add § 53.1342, which would establish requirements for the expiration of a CP. § 53.1345 Transfer of Construction Permits This proposed rule would add § 53.1345, which would govern the transfer of CPs under part 53. § 53.1348 Termination of Construction Permits This proposed rule would add § 53.1348, which would require the holder of a permit under part 53 to provide written certification to the Commission within 30 days of determining to permanently cease construction. § 53.1360 Operating Licenses This proposed rule would add § 53.1360, which would provide an overview of the requirements and procedures for applicants to apply for and the Commission to issue an OL under part 53. § 53.1366 Contents of Applications for Operating Licenses; General Information This proposed rule would add § 53.1366, which would require an application for an OL under part 53 to include the information required by § 53.1109 as well as financial information. § 53.1369 Contents of Applications for Operating Licenses; Technical Information This proposed rule would add § 53.1369, which would require an application for an OL under part 53 to include certain technical information in an FSAR at a level of detail sufficient for the Commission to reach a final conclusion on all safety matters. § 53.1372 Contents of Applications for Operating Licenses; Other Application Content This proposed rule would add § 53.1372, which would require an application for an OL under part 53 to include an environmental report and a description of availability controls. § 53.1375 Review of Applications This proposed rule would add § 53.1375, which would establish the standards and procedures for reviewing applications and holding hearings on OLs under part 53. § 53.1381 Referral to the Advisory Committee on Reactor Safeguards This proposed rule would add § 53.1381, which would require the ACRS to report on any portions of the application for a CP under part 53 concerning safety. § 53.1384 Exemptions, Departures, and Variances This proposed rule would add § 53.1384, which would govern requests for and the issuance of exemptions from the Commission's regulations and exemptions, departures, and variances from NRC approvals, permits, and certifications. § 53.1387 Issuance of Operating Licenses This proposed rule would add § 53.1387, which would establish the conditions under which the Commission may issue OLs and accompanying conditions and limitations, including TS, under part 53. § 53.1390 Backfitting of Operating Licenses This proposed rule would add § 53.1390, which would prevent the Commission from modifying, adding, or deleting any terms or conditions of the OL, except in accordance with § 53.1590. § 53.1396 Duration of Operating Licenses This proposed rule would add § 53.1396, which would provide that an OL under part 53 may be valid for up to 40 years. § 53.1399 Transfer of an Operating License This proposed rule would add § 53.1399, which would provide that an OL under part 53 may be transferred under § 53.1570. § 53.1402 Application for Renewal This proposed rule would add § 53.1402, which would provide that an application for a renewed OL under part 53 must be filed in accordance with § 53.1595. § 53.1405 Continuation of an Operating License This proposed rule would add § 53.1405, which would govern the continuing obligations of the holder of an OL under part 53 following the permanent cessation of operations.
                        ( print page 87002) § 53.1410 Combined Licenses This proposed rule would add § 53.1410, which would provide an overview of the requirements and procedures for applicants to apply for and the Commission to issue a COL under part 53. § 53.1413 Contents of Applications for Combined Licenses; General Information This proposed rule would add § 53.1413, which would require an application for a COL under part 53 to include the information required by § 53.1109 as well as financial information. § 53.1416 Contents of Applications for Combined Licenses; Technical Information This proposed rule would add § 53.1416, which would require applicants for a COL under part 53 to submit an FSAR with a level of technical information sufficient to reach a final conclusion on all safety matters. § 53.1419 Contents of Applications for Combined Licenses; Other Application Content This proposed rule would add § 53.1419, which would require applicants for a COL under part 53 to submit an environmental report, a description of availability controls, the ITAAC that the licensee must perform. It would also include ITAAC requirements for applications that reference an ESP, standard DC, ML, or combination thereof. § 53.1422 Review of Applications This proposed rule would add § 53.1422, which would require applications for COLs under part 53 to be reviewed for compliance with applicable standards and establish procedural requirements for reviewing applications and holding hearings in accordance with part 2. § 53.1425 Finality of Referenced NRC Approvals This proposed rule would add § 53.1425 which would address the finality of ESPs, standard DC rules, standard design approvals, or MLs referenced in the application for a COL under part 53. § 53.1431 Referral to the Advisory Committee on Reactor Safeguards This proposed rule would add § 53.1431, which would require the ACRS to report on any portions of the application for a COL under part 53 concerning safety. § 53.1434 Authorization To Conduct Limited Work Authorization Activities This proposed rule would add § 53.1434, which would address authorization to conduct LWA activities. § 53.1437 Exemptions, Departures, and Variances This proposed rule would add § 53.1437, which would govern the conditions in which the Commission may grant an exemption for one or more of its regulations, or an exemption, variance, or departure from a permit, design approval, or license. § 53.1440 Issuance of Combined Licenses This proposed rule would add § 53.1440, which would establish the conditions under which the Commission may issue COLs and accompanying conditions and limitations, including TS, under part 53. § 53.1443 Finality of Combined Licenses This proposed rule would add § 53.1443, which would govern permissible modifications or amendments that the Commission may make to a COL, as well as permissible changes that a licensee may make to facilities and procedures as described in the FSAR. § 53.1449 Inspection During Construction This proposed rule would add § 53.1449, which would establish requirements related to inspections, tests, or analyses for the holder of a COL under part 53. § 53.1452 Operation Under a Combined License This proposed rule would add § 53.1452, which would establish requirements describing the notifications, hearings, and findings to be made prior to commencing facility operations. § 53.1455 Duration of a Combined License This proposed rule would add § 53.1455, which would govern the duration of a COL under part 53. § 53.1456 Transfer of a Combined License This proposed rule would add § 53.1456, which would permit the transfer of a COL under part 53 in accordance with § 53.1570. § 53.1458 Application for Renewal This proposed rule would add § 53.1458, which would provide that an application for renewal of a COL must be filed in accordance with § 53.1595. § 53.1461 Continuation of Combined License This proposed rule would add § 53.1461, which would govern the continuing obligations of the holder of a COL under part 53 following the permanent cessation of operations. § 53.1470 Standardization of Commercial Nuclear Plant Designs: Licenses To Construct and Operate Nuclear Power Reactors of Identical Design at Multiple Sites This proposed rule would add § 53.1470, which would govern the requirements and procedures for filing and issuing applications for a CP, OL, or COL under part 53 in which the applicant seeks approval of the same design for multiple sites. Subpart I—Maintaining and Revising Licensing-Basis Information This proposed rule would add subpart I, which would address the maintenance of licensing-basis information for part 53. § 53.1500 Licensing-Basis Information This proposed rule would add § 53.1500, describing the purpose of subpart I, which would be to provide the requirements for the maintenance of licensing-basis information for commercial nuclear plants licensed under part 53. § 53.1502 Specific Terms and Conditions of Licenses This proposed rule would add § 53.1502, which would outline the specific terms and conditions for obtaining a license under part 53. § 53.1505 Changes to Licensing-Basis Information Requiring Prior NRC Approval This proposed rule would add § 53.1505, which would provide an overview of the process for licensees to request, and the Commission to issue, amendments to licensing-basis information under part 53. § 53.1510 Application for Amendment of License This proposed rule would add § 53.1510, which would require licensees under part 53 to file an application to request an amendment to the license. Applicants must assess how their requested changes would impact the safety criteria and analysis 
                        ( print page 87003)
                        requirements in subpart B and C, as applicable, whether the amendment involves no significant hazards consideration using the standards in § 53.1520 and consider potential impacts on environmental factors. § 53.1515 Public Notices; State Consultation This proposed rule would add § 53.1515, which would outline the Commission's procedures for issuing a notification in the 
                        **Federal Register**
                         and consulting with the State in which the commercial nuclear facility is located in connection with its consideration of applications for an amendment to an OL or COL under part 53. § 53.1520 Issuance of Amendment This proposed rule would add § 53.1520, which would outline criteria for the Commission to consider in issuing license amendments under part 53. § 53.1525 Revising Certification Information Within a Design Certification Rule This proposed rule would add § 53.1525, which would address the requirements for applicants to request, and the Commission to grant, an exemption to a DC rule under part 53. § 53.1530 Revising Design Information Within a Manufacturing License This proposed rule would add § 53.1530, which would require the holder of an ML to request an amendment under § 53.1510 and, as applicable, § 53.1520 to make changes to the design of a manufactured reactor. It would also outline the requirements for holders of a COL under part 53 to request amendments for changes to the design information of a manufactured reactor. § 53.1535 Amendments During Construction This proposed rule would add § 53.1535, which would outline the process for licensees under part 53 to request amendments to CPs or LWAs during construction. § 53.1540 Updating Licensing-Basis Information and Determining the Need for NRC Approval This proposed rule would add § 53.1540, which would provide an overview of the regulations in subpart I for holders of an OL or COL under part 53 to modify licensing-basis information and definitions relevant to §§ 53.1545 through 53.1565. § 53.1545 Updating Final Safety Analysis Reports This proposed rule would add § 53.1545, which would require licensees under part 53 to regularly update FSARs in accordance with the requirements of this section to reflect changes to licensing-basis information. § 53.1550 Evaluating Changes to Facility as Described in Final Safety Analysis Reports This proposed rule would add § 53.1550, which would require licensees under part 53 to follow the guidelines outlined in this section in determining whether changes to licensing-basis information described in the FSAR (as updated) require them to obtain a license amendment. § 53.1560 Updating Program Documents Included in Licensing-Basis Information This proposed rule would add § 53.1560, which would require the holders of an OL or COL under part 53 to regularly update the program documents that they submitted in their application for a license. § 53.1565 Evaluating Changes to Programs Included in Licensing-Basis Information This proposed rule would add § 53.1565, which would enable licensees under part 53 to make changes to the facility, procedures, or organization, or address changes to site environs as described in program documents without NRC approval if these changes satisfy the criteria outlined in this section. § 53.1570 Transfer of Licenses This proposed rule would add § 53.1570, which would outline the requirements for an application for transfer of a license issued under part 53. § 53.1575 Termination of Licenses This proposed rule would add § 53.1575, which would outline the process for terminating an OL or COL issued under part 53. § 53.1580 Information Requests This proposed rule would add § 53.1580, which would address the process and circumstances under which the NRC may send information requests to the various types of licensees within part 53. § 53.1585 Revocation, Suspension, Modification of Licenses and Approvals for Cause This proposed rule would add § 53.1585, which would address grounds for the revocation, suspension, or modification of a license or standard design approval issued under part 53. § 53.1590 Backfitting This proposed rule would add § 53.1590, which would define backfitting and establish requirements to be met by the NRC when it takes backfitting actions under part 53. § 53.1595 Renewal This proposed rule would add § 53.1595, which would provide for the renewal of a license under part 53 upon expiration. Subpart J—Reporting and Other Administrative Requirements This proposed rule would add subpart J, to establish various reporting and other administrative requirements for licensees under part 53. § 53.1600 General Information This proposed rule would add § 53.1600, which provides an overview of the sections that would require applicants and licensees under part 53 to provide NRC inspectors with unfettered access to sites and facilities, maintain records and make reports, demonstrate compliance with financial qualification and reporting requirements, and maintain required financial protection for accidents. § 53.1610 Unfettered Access for Inspections This proposed rule would add § 53.1610, which would require applicants and licensees under part 53 to provide unfettered access to NRC inspectors, including access to records, premises, activities, and licensed materials, in addition to providing office space to accommodate temporary or resident inspectors. § 53.1620 Maintenance of Records, Making of Reports This proposed rule would add § 53.1620, which would require part 53 licensees to maintain all records and make reports as required by the conditions of the license or by the regulations in part 53. § 53.1630 Immediate Notification Requirements for Operating Commercial Nuclear Plants This proposed rule would add § 53.1630, which would impose immediate notification requirements on 
                        ( print page 87004)
                        part 53 licensees following the declaration of an Emergency Class or the discovery of certain non-emergency events. § 53.1640 Licensee Event Report System This proposed rule would add § 53.1640, which would require any commercial plant licensee holding an OL under part 53 to submit a Licensee Event Report in accordance with the specifications outlined in this section. § 53.1645 Reports of Radiation Exposure to Members of the Public The proposed rule would add § 53.1645, which would require annual reports to the Commission, including radiological reports as required by part 20, an Annual Radioactive Effluent Release Report, and an Annual Environmental Operating Report. § 53.1650 Facility Information and Verification The proposed rule would add § 53.1650, which would include a reporting requirement for applicants and holders of a CP or license under part 53 to support safeguards agreements between the United States and the IAEA. § 53.1660 Financial Requirements This proposed rule would add § 53.1660, which would introduce requirements and procedures related to financial qualifications and reporting requirements for applicants, licensees, and CP holders under part 53. § 53.1670 Financial Qualifications This proposed rule would add § 53.1670, which would require an applicant for a CP, OL, or COL under part 53 to must demonstrate possession or ability to obtain funds necessary for the activities for which the permit or license is sought. § 53.1680 Annual Financial Reports This proposed rule would add § 53.1680, which would require licensees and holders of a CP under part 53 to submit annual financial reports to the Commission, with exceptions for those that submit financial forms to the Securities and Exchange Commission or the Federal Energy Regulatory Commission. § 53.1690 Licensee's Change of Status; Financial Qualifications This proposed rule would add § 53.1690, which would require electric utility licensees that hold an OL or COL for a commercial nuclear plant under part 53 to provide the NRC with the financial qualifications information outlined in this section within seventy-five days of ceasing to be an electric utility. § 53.1700 Creditor Regulations This proposed rule would add § 53.1700, which would establish regulations with respect to the creditors of any facility under part 53. § 53.1710 Financial Protection This proposed rule would add § 53.1710, which would establish requirements for licenses under part 53 to obtain and maintain insurance to cover the costs of an accident. § 53.1720 Insurance Required To Stabilize and Decontaminate Plant Following an Accident This proposed rule would add § 53.1720, which would require commercial nuclear plant licensees under part 53 to obtain insurance sufficient to cover the costs of stabilizing and decontaminating the plant in the event of an accident. § 53.1730 Financial Protection Requirements This proposed rule would add § 53.1730, which would require commercial nuclear plant licensees under part 53 to satisfy the provisions of part 140. Subpart M—Enforcement This proposed rule would add subpart M, which would address certain violations and penalties associated with violations of part 53 regulations. § 53.9000 Violations This proposed rule would add § 53.9000, providing notice of the Commission's authority to obtain injunctions or other court orders for the violations enumerated in this section. § 53.9010 Criminal Penalties This proposed rule would add § 53.9010, providing notice to all persons and entities subject to part 53 that they are subject to criminal sanctions for willful violations, attempted violations, or conspiracy to violate certain regulations under part 53. § 70.20a General License to Possess Special Nuclear Material for Transport This proposed rule would revise § 70.20a(b) to include a reference to part 53. § 70.22 Contents of Applications This proposed rule would revise § 70.22, paragraphs (b), (h)(1), (j)(1), and (k) to include the appropriate references to part 53. § 70.24 Criticality Accident Requirements This proposed rule would revise § 70.24(d) to include the appropriate references to part 53. § 70.32 Conditions of Licenses This proposed rule would revise § 70.32(c)(1) and (d) to incorporate the appropriate references to part 53. § 70.50 Reporting Requirements This proposed rule would revise § 70.50(d) to clarify the applicability of the reporting requirements of this section to part 53 licensees. § 72.3 Definitions This proposed rule would revise the definition of “
                        *Independent spent fuel storage installation or ISFSI*
                        ” in § 72.3 to include a reference to facilities licensed under part 53. § 72.30 Financial Assurance and Recordkeeping for Decommissioning This proposed rule would revise § 72.30(e)(5) to include the appropriate references to part 53. § 72.32 Emergency Plan This proposed rule would revise § 72.32(c)(2) to include a reference to the exclusion area as defined in part 53. § 72.40 Issuance of License This proposed rule would revise § 72.40(c) regarding the issuance of a license under part 72 to include a reference to previous licensing actions, including the issuance of a CP under part 53. § 72.75 Reporting Requirements for Specific Events and Conditions This proposed rule would revise § 72.75(i)(1)(ii) regarding reporting requirements for specific events and conditions with references to reactors licensed under part 53. § 72.184 Safeguards Contingency Plan This proposed rule would revise § 72.184(a) regarding the requirements of a licensee's safeguarding contingency plan with a reference to nuclear facilities licensed under part 53. § 72.210 General License Issued This proposed rule would revise § 72.210 to issue a general license for the storage of spent fuel in an independent spent storage installation at power to persons authorized to possess or operate nuclear power reactors under part 53.
                        ( print page 87005) § 72.212 Conditions of General License Issued Under § 72.210 This proposed rule would revise § 72.212(b)(8) regarding the conditions of a general license issued under § 72.210 to include a reference to license amendments for a facility made pursuant to part 53. § 72.218 Termination of Licenses This proposed rule would revise § 72.218(a) to include a reference to the notification required under part 53 regarding the plan for managing spent fuel prior to decommissioning. It would also extend the provisions of § 72.218(b) to a reactor operating or COL under part 53. § 73.1 Purpose and Scope This proposed rule would revise § 73.1(b)(1)(i) to extend the scope of part 73 to production and utilization facilities licensed under part 53, in addition to parts 50 and 52. § 73.2 Definitions This proposed rule would revise § 73.2 introductory text and paragraph (a) such that terms defined in part 53 have the same meaning in part 73. § 73.8 Information Collection Requirements: OMB Approval This proposed rule would revise § 73.8(b) with the new information collection requirements contained in proposed §§ 73.77, 73.100, 73.110, and 73.120. § 73.50 Requirements for Physical Protection of Licensed Activities This proposed rule would revise § 73.50 to exempt nuclear reactor facilities licensed under part 53, in addition to parts 50 and 52, from the requirements of this section. § 73.55 Requirements for Physical Protection of Licensed Activities in Nuclear Power Reactors Against Radiological Sabotage This proposed rule would revise § 73.55, paragraphs (a)(4) and (6), (i)(4)(iii), (l)(1), (l)(7)(ii), (p)(1)(i), (r)(2), and (r)(4)(iii), to incorporate the appropriate references to part 53 regarding requirements for physical protection of licensed activities in nuclear power reactors against radiological sabotage. § 73.56 Personnel Access Authorization Requirements for Nuclear Power Plants This proposed rule would revise § 73.56(a)(3) to apply this section's personnel AA requirements to applicants for an OL or holders of a COL under part 53 who do not demonstrate compliance with certain requirements under part 53. § 73.57 Requirements for Criminal History Records Checks of Individuals Granted Unescorted Access to a Nuclear Power Facility, a Non-power Reactor, or Access to Safeguards Information This proposed rule would revise § 73.57(a)(3) to incorporate the appropriate references to OLs granted under part 53 and Commission findings under § 53.1452(g) regarding the requirement for license applicants to submit fingerprints for all personnel with unescorted access. § 73.58 Safety/Security Interface Requirements for Nuclear Power Reactors This proposed rule would revise § 73.58(a) to extend the requirements of this section to part 53 licensees. § 73.67 Licensee Fixed Site and In-Transit Requirements for the Physical Protection of Special Nuclear Material of Moderate and Low Strategic Significance This proposed rule would revise § 73.67(d) and (f) to include a reference to licensees authorized to operate a nuclear power plant under part 53. § 73.77 Cybersecurity Event Notifications This proposed rule would revise § 73.77, paragraphs (a), (b), (c)(6) and (7) regarding the notification process for cybersecurity events to include notifications for the declaration of an emergency class made under part 53. Subpart J—Security Requirements at Commercial Nuclear Plants This proposed rule would add new Subpart J of part 73 containing §§ 73.100, 73.110, and 73.120, to establish security requirements for commercial nuclear plants licensed under part 53. § 73.100 Technology-Inclusive Requirements for Physical Protection of Licensed Activities at Commercial Nuclear Plants Against Radiological Sabotage This proposed rule would add § 73.100, which would establish a performance-based regulatory framework for physical protection as an alternative to the prescriptive requirements of § 73.55, which also governs physical protection programs for part 50 and 52 licensees. § 73.110 Technology-Inclusive Requirements for Protection of Digital Computer and Communication Systems and Networks This proposed rule would add § 73.110, which would establish a consequence-based approach to cybersecurity and would require that part 53 licensees demonstrate reasonable assurance that digital computer and communication systems and networks are adequately protected against cyberattacks in a manner that is commensurate with the potential consequences of those attacks. § 73.120 Access Authorization Program for Commercial Nuclear Plants This proposed rule would add § 73.120, which would establish performance objectives as an alternative to compliance with the AA provisions of §§ 73.55, 73.56, and 73.57. This proposed rule would afford part 53 licensees additional flexibility in establishing an AA program that demonstrates compliance with the performance objectives and requirements of this section. § 73.1200 Notification of Physical Security Events This proposed rule would revise § 73.1200, paragraphs (a), (c)(1), (e)(1), (e)(3), (e)(4), (g)(1), (o)(5)(i), (o)(6)(i), (r), and (s) to extend the requirements of this section to part 53 licensees. § 73.1205 Written Follow-Up Reports of Physical Security Events This proposed rule would revise § 73.1205(b)(2) to extend the requirements of this section to part 53 licensees. § 73.1210 Recordkeeping of Physical Security Events This proposed rule would revise § 73.1210(a)(1) and (b)(3)(i) to extend the requirements of this section to part 53 licensees. § 73.1215 Suspicious Activity Reports This proposed rule would revise § 73.1215(d)(1) to include a reference to § 73.100. Appendix B to part 73—General Criteria for Security Personnel This proposed rule would revise appendix B to part 73 to state that terms defined in part 53 have the same meaning when used in this appendix. § 74.31 Nuclear Material Control and Accounting for Special Nuclear Material of Low Strategic Significance This proposed rule would revise § 74.31(a) to include a reference to 
                        ( print page 87006)
                        production or utilization facilities licensed under part 53, in addition to parts 50 and 70. § 74.41 Nuclear Material Control and Accounting for Special Nuclear Material of Moderate Strategic Significance This proposed rule would revise § 74.41(a) to include a reference to nuclear reactors licensed under part 53. § 74.51 Nuclear Material Control and Accounting for Strategic Special Nuclear Material This proposed rule would revise § 74.51(a) to include a reference to nuclear reactors licensed under part 53. § 75.4 Definitions This proposed rule would revise § 75.4 such that terms defined in § 53.020 have the same meaning when used in this part. The definition of “
                        *Facility*
                        ” would also be revised to include any plant or location where more than 1 effective kilogram of nuclear material is licensed pursuant to part 53. § 95.5 Definitions This proposed rule would revise the definition of “
                        *License*
                        ” in § 95.5 to include those issued under part 53. § 95.39 External Transmission of Documents and Material This proposed rule would revise § 95.39(a) to apply restrictions to the external transmission of documents and material containing classified information in connection with NRC licenses, certificates, standard design approvals, or standard DCs issued under part 53. § 140.2 Scope This proposed rule would revise § 140.2(a)(1) and (2) to include part 53 applicants and licensees within the scope of part 140 regulations. § 140.10 Scope This proposed rule would revise § 140.10 to apply the provisions of subpart B to applicants or holders of a license to operate a nuclear reactor under part 53, as well as applicants and holders of a COL under part 53. § 140.11 Amounts of Financial Protection for Certain Reactors This proposed rule would revise § 140.11(b) to require the licensee's primary financial protection to cover all reactors in any case where a person is authorized under part 53 to operate two or more nuclear reactors at the same location. § 140.12 Amount of Financial Protection Required for Other Reactors This proposed rule would revise § 140.12(c) to require the licensee's primary financial protection to cover all reactors in any case where a person is authorized under part 53 to operate two or more nuclear reactors at the same location. § 140.13 Amount of Financial Protection Required of Certain Holders of Construction Permits and Combined Licenses Under 10 CFR Part 52 This proposed rule would revise § 140.13 with the appropriate references to part 53 regarding the requirement for holders of a CP or COL under part 53 to obtain financial protection. § 140.20 Indemnity Agreements and Liens This proposed rule would revise § 140.20(a)(1)(i) and (ii) with appropriate references to part 53. § 150.15 Persons Not Exempt The proposed rule would revise § 150.15, paragraphs (a)(7)(iii) and (a)(8) to add a reference to facilities licensed under parts 53 and 52. § 170.3 Definitions The proposed rule would revise § 170.3 to incorporate references to part 53 into the definitions of “
                        *Manufacturing license,*
                        ” “
                        *Part 55 Reviews,*
                        ” “
                        *Power reactor,*
                        ” and “
                        *Special projects.*
                        ” § 170.12 Payment of Fees The proposed rule would revise § 170.12(d)(1)(v) regarding special project fees in connection with FSARs to include part 53. § 170.21 Schedule of Fees for Production and Utilization Facilities, Review of Standard Referenced Design Approvals, Special Projects, Inspections, And import and Export Licenses The proposed rule would revise § 170.21, footnote 1 to include fees charged for approvals issued under the exemption provision in § 53.080. § 170.41 Failure by Applicant or Licensee to Pay Prescribed Fees The proposed rule would revise § 170.41 to include a general reference to part 53 in connection with remedial actions that the Commission might take when an applicant or licensee fails to pay a prescribed fee required by this part. § 171.3 Scope The proposed rule would revise § 171.3 to apply the provisions of this part to any person holding an OL for a power reactor licensed under part 53 or a COL issued under part 53. § 171.5 Definitions This proposed rule would revise the definitions of “
                        *Operating license*
                        ” and “
                        *Power reactor*
                        ” in § 171.5 to incorporate the appropriate references to part 53. § 171.15 Annual fees: Non-Power Production or Utilization Licenses, Reactor Licenses, and Independent Spent Fuel Storage Licenses This proposed rule would revise § 171.15, paragraphs (a), (b)(2)(iii), (c)(1), and (d)(1) regarding annual fees that are applicable to part 53 licensees. § 171.17 Proration This proposed rule would revise § 171.17, paragraphs (a), (a)(1)(ii) and (a)(2) with references to part 53 licenses. VIII. Regulatory Flexibility Certification The Regulatory Flexibility Act of 1980, as amended at 5 U.S.C. 601 
                        *et seq,*
                         requires that agencies consider the impact of their rulemakings on small entities and, consistent with applicable statutes, consider alternatives to minimize these impacts on the businesses, organizations, and government jurisdictions to which they apply. In accordance with the Small Business Administration's (SBA's) regulation at 13 CFR 121.903(c), the NRC has developed its own size standards for performing an RFA analysis and has verified with the SBA Office of Advocacy that its size standards are appropriate for NRC analyses. The NRC size standards at § 2.810, “NRC size standards,” are used to determine whether an applicant or licensee qualifies as a small entity in the NRC's regulatory programs. Section 2.810 defines the following types of small entities: Small business is a for-profit concern and is a—(1) Concern that provides a service or a concern not engaged in manufacturing with average gross receipts of $8.0 million or less over its last 5 completed fiscal years; or (2) Manufacturing concern with an average number of 500 or fewer employees based upon employment during each pay period for the preceding 12 calendar months. Small organization is a not-for-profit organization which is independently 
                        ( print page 87007)
                        owned and operated and has annual gross receipts of $8.0 million or less. Small governmental jurisdiction is a government of a city, county, town, township, village, school district, or special district with a population of less than 50,000. Small educational institution is one that is—(1) Supported by a qualifying small governmental jurisdiction; or (2) Not State or publicly supported and has 500 or fewer employees. Number of Small Entities Affected The NRC is currently not aware of any known small entities as defined in § 2.810 that are planning to apply for a commercial nuclear plant ESP, CP, OL, ML, or COL under part 53 that would be impacted by this proposed rule. Based on this finding, the NRC has preliminarily determined that the proposed rule would not have a significant economic impact on a substantial number of small entities. Economic Impact on Small Entities Depending on how the ownership and/or operating responsibilities for such an enterprise were structured, applicants for a commercial nuclear plant rated 8 Megawatts electric (MWe) or less could conceivably qualify as small entities as defined by § 2.810. Owners that operate power reactors rated greater than 8 MWe could generate sufficient electricity revenue that exceeds the gross annual receipts limit of $8 million, assuming a 90 percent capacity factor and the June 2021 DOE's Energy Information Administration U.S. average price of electricity to the ultimate customer for all sectors of 11.3 cents per kilowatt-hour. Although the NRC is not aware of any small entities that would be affected by the proposed rule, there is a possibility that future applications for a commercial nuclear plant permit or license could be submitted by small entities who plan to own and operate a commercial nuclear plant rated 8 MWe or less. Commercial nuclear plants that are rated 8 MWe or less would most likely be used to support electrical demand for military bases or small remote towns and would provide process heat, so they would not directly compete with a larger commercial nuclear plant that would typically produce electricity for the grid. As a result of these differing purposes, the NRC would expect that small and large entities would not be in direct competition with each other. Therefore, the NRC preliminarily concludes that this proposed rule would not have a significant economic impact on a substantial number of small entities. Request for Comments The NRC is seeking comment on both its initial RFA analysis and on its preliminary conclusion that this proposed rule would not have a significant economic impact on a substantial number of small entities because of the likelihood that most expected applicants would not qualify as a small entity. Additionally, the NRC is seeking comment on its preliminary conclusion that if a small entity were to submit a commercial nuclear plant application, the small entity would not incur a significant economic impact as it would most likely not be in competition with a large entity. Any small entity that could be subject to this regulation that determines, because of its size, it is likely to bear a disproportionate adverse economic impact should notify the Commission of this opinion in a comment that indicates— 1. The applicant's size and how the proposed regulation would impose a significant economic burden on the applicant as compared to the economic burden on a larger applicant; 2. How the proposed regulations could be modified to take into account the applicant's differing needs or capabilities; 3. The benefits that would accrue or the detriments that would be avoided if the proposed regulations were modified as suggested by the applicant; 4. How the proposed regulation, as modified, would more closely equalize the impact of NRC regulations or create more equal access to the benefits of Federal programs as opposed to providing special advantages to any individual or group; and 5. How the proposed regulation, as modified, would still adequately demonstrate compliance with the NRC's obligations under the Act. IX. Regulatory Analysis The NRC has prepared a draft regulatory analysis for this proposed rule. The analysis examines the costs and benefits of the alternatives considered by the NRC. The conclusion from the analysis is that this proposed rule and associated guidance would result in net averted costs to the industry and the NRC of $28.1 million using a 7-percent discount rate and $34.5 million using a 3-percent discount rate due to reductions in exemption requests. The analysis also assumes one applicant under part 53. As the number of applicants increases, so do the estimated averted costs. The NRC requests public comment on the draft regulatory analysis, which is available as indicated in the “Availability of Documents” section of this document. Comments on the draft regulatory analysis may be submitted to the NRC as indicated under the 
                        **ADDRESSES**
                         caption of this document. X. Backfitting and Issue Finality This section describes the backfitting and issue finality implications of this proposed rule and the draft guidance documents described in section XVIII, “Availability of Guidance,” in this document, as applied to pertinent NRC approvals and certain applicants that reference NRC approvals in their applications. The NRC's current backfitting provisions associated with nuclear power plants appear in § 50.109, “Backfitting,” and apply to CPs and OLs under part 50. Issue finality provisions (analogous to the backfitting provisions in § 50.109) for approvals under part 52 are located in various provisions of part 52. The NRC Management Directive 8.4, “Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests,” describes the Commission's policies on backfitting and issue finality. This proposed rule would provide a regulatory scheme for entities to apply for approvals under part 53. The part 50 backfitting provisions and part 52 issue finality provisions apply to actions taken by the NRC under part 50 or part 52, respectively, or actions taken by the NRC under other parts of 10 CFR chapter I that, for holders of certain approvals under part 50 or part 52, inextricably affect their activities regulated under part 50 or part 52. Issuance and implementation of proposed part 53 would not constitute actions taken under part 50 or part 52. Also, proposed part 53 would not allow an applicant to reference approvals issued under part 50 or part 52. Therefore, the issuance and implementation of proposed part 53 would not affect part 50 or part 52 entities' activities regulated under part 50 or part 52. Therefore, the addition of part 53 through this proposed rule would not be within the scope of the part 50 backfitting and part 52 issue finality provisions. The NRC also proposes conforming changes to parts 1, 2, 10, 11, 19, 20, 21, 25, 26, 30, 40, 50, 51, 70, 72, 73, 74, 75, 95, 140, 150, 170, and 171 to reflect the addition of part 53. These changes would not meet the definition of “backfitting” in § 50.109 or § 70.76, “Backfitting,” because the proposed changes would not modify or add to the systems, structures, components, or 
                        ( print page 87008)
                        design of a facility or to the procedures or organization required to operate a facility under part 50 or 70. These changes would not meet the definition of “backfitting” in § 72.62, “Backfitting,” because the proposed changes would not add, eliminate, or modify the SSCs of an independent spent fuel storage installation (ISFSI) or the procedures or organization required to operate an ISFSI. These proposed changes would not inextricably affect activities regulated under parts 50, 52, 70, or 72. Therefore, the proposed changes to parts 1, 2, 10, 11, 19, 20, 21, 25, 26, 30, 40, 50, 51, 70, 72, 73, 74, 75, 95, 140, 150, 170, and 171 would not constitute backfitting under parts 50, 70, or 72 or affect the issue finality of an approval under part 52. The NRC is issuing 10 draft guidance documents that, if issued as final guidance documents, would provide guidance on the methods acceptable to the NRC for complying with aspects of this proposed rule. These documents would not apply to holders of approvals issued under part 50 or part 52. Further, as discussed in the guidance documents, applicants and licensees would not be required to comply with the positions set forth in the guidance. Therefore, issuance of the guidance documents as final guidance would not constitute backfitting under part 50 or affect the issue finality of any approval issued under part 52. XI. Cumulative Effects of Regulation The NRC seeks to minimize any potential negative consequences resulting from the cumulative effects of regulation (CER). The CER describes the challenges that licensees, or other impacted entities such as State partners, may face while implementing new regulatory positions, programs, or requirements (
                        *e.g.,*
                         rules, generic letters, backfits, inspections). The CER is an organizational effectiveness challenge that may result from a licensee or impacted entity implementing a number of complex regulatory actions, programs, or requirements within limited available resources. The NRC's CER process involved engaging with external stakeholders throughout this proposed rule and related regulatory activities. Public involvement has included numerous public meetings to examine the part 53 risk-informed, technology-inclusive requirements for commercial nuclear plants and the publication of numerous versions of preliminary proposed rule language. The NRC is considering holding additional public meetings during the remainder of the rulemaking process. In parallel with this proposed rule, the NRC is issuing 10 draft implementing guidance documents for comment to support informed external stakeholder feedback. Section XVII, “Availability of Guidance,” of this document describes how the public can access the draft implementing guidance. In addition to the questions in the “Specific Requests for Comments” section of this document, the NRC is requesting CER feedback on the following questions: 1. In light of any current or projected CER challenges, does the proposed rule's effective date provide sufficient time to implement the new proposed requirements, including changes to programs, procedures, and the facility? 2. If CER challenges currently exist or are expected, what should be done to address them? For example, if more time is required for implementation of the new requirements, what period of time is sufficient? 3. Do other (NRC or other agency) regulatory actions (
                        *e.g.,*
                         orders, generic communications, license amendment requests, inspection findings of a generic nature) influence the implementation of the proposed rule's requirements? 4. Are there unintended consequences? Does the proposed rule create conditions that would be contrary to the proposed rule's purpose and objectives? If so, what are the unintended consequences, and how should they be addressed? 5. Please comment on the NRC's cost and benefit estimates in the regulatory analysis that supports this proposed rule. The draft regulatory analysis is available as indicated under the “Availability of Documents” section of this document. XII. Plain Writing The Plain Writing Act of 2010 (Pub. L. 111-274) requires Federal agencies to write documents in a clear, concise, and well-organized manner. The NRC has written this document to be consistent with the Plain Writing Act as well as the Presidential Memorandum, “Plain Language in Government Writing,” published June 10, 1998 (63 FR 31885). The NRC requests comment on this document with respect to the clarity and effectiveness of the language used. XIII. Environmental Assessment and Proposed Finding of No Significant Environmental Impact The Commission has preliminarily determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in subpart A of part 51, that this rule, if adopted, would not be a major Federal action significantly affecting the quality of the human environment, and an EIS is not required. The implementation of the proposed rule requirements does not have a significant impact on the environment. The proposed rulemaking would either have requirements that are administrative in application, matters of procedure, or provide an equivalent level of safety as existing requirements; therefore, there would be similar environmental impacts from the implementation of the part 53 regulations as there are for existing requirements. The preliminary determination of this EA is that there will be no significant effect on the quality of the human environment from this action. Public stakeholders should note, however, that comments on any aspect of this EA may be submitted to the NRC as indicated under the 
                        **ADDRESSES**
                         section of this document. The EA is available as indicated under the “Availability of Documents” section of this document. The NRC has sent a copy of the EA, and this proposed rule to every State Liaison Officer and has requested comments. XIV. Paperwork Reduction Act This proposed rule contains new collections of information contained in parts 26, 50, 53, and 73 and NRC Forms 361S, 366, 366A, 366B, 893, and 894 that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq). The collections of information have been submitted to the OMB for review and approval. The proposed changes to parts 2, 10, 11, 19, 20, 21, 25, 30, 40, 51, 70, 72, 74, 75, 95, 140, 150, 170, and 171 do not contain any new or amended collections of information subject to the Paperwork Reduction Act of 1995. Existing collections of information were approved by the OMB, approval numbers 3150-0062 (part 11), 3150-0044 (part 19), 3150-0014 (part 20), 3150-0035 (part 21), 3150-0046 (part 25), 3150-0017 (part 30), 3150-0020 (part 40), 3150-0021 (part 51), 3150-0024 (NRC Form 396), 3150-0090 (NRC Form 398), 3150-0009 (part 70), 3150-0132 (part 72), 3150-0123 (part 74), 3150-0055 (part 75), 3150-0047 (part 95), 3150-0039 (part 140), and 3150-0032 (part 150). *Type of submission, new or revision:*
                         Revision and new. *The title of the information collection:*
                         Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors.
                        ( print page 87009) *The form number if applicable:*
                         NRC Forms 361S, 366, 366A, 366B, 893, and 894. *How often the collection is required or requested:*
                         Once, on occasion, every 30 days, biannually, annually, biennially, every four years, every five years, every ten years. *Who will be required or asked to respond:*
                         Part 53 commercial nuclear plant licensees and license applicants for commercial nuclear plants to be licensed under part 53. *An estimate of the number of annual responses:*
                         15 (2 responses for Part 26, 11 responses for Part 53, 2 responses for Part 50 and 0 responses for Part 73 and NRC Forms 361S, 366, 366A, 366B, 893, and 894) *The estimated number of annual respondents:*
                         2 (2 respondents for Part 26, 2 respondents for Part 53, 2 respondents for Part 50 and 0 respondents for Part 73 and NRC Forms 361S, 366, 366A, 366B, 893, and 894) *An estimate of the total number of hours needed annually to comply with the information collection requirement or request:*
                         230,244 hours. (656 hours for Part 26, 220,801 hours for Part 53, 8,767 hours for Part 50 and 0 hours for Part 73 and NRC Forms 361S, 366, 366A, 366B, 893, and 894) *Abstract:*
                         The NRC is proposing to establish an optional technology-inclusive regulatory framework for use by applicants for new commercial nuclear plant designs. The regulatory requirements developed in this rulemaking would use methods of evaluation, including risk-informed and performance-based methods, that are flexible and practicable for application to a variety of new reactor technologies. The NRC's goals in amending these regulations are to continue to provide reasonable assurance of adequate protection of public health and safety and the common defense and security at reactor sites at which new nuclear reactor designs are deployed to at least the same degree of protection as required for current-generation LWRs; protect health and minimize danger to life or property to at least the same degree of protection as required for current-generation LWRs; provide greater operational flexibilities where supported by enhanced margins of safety that may be provided in new nuclear designs; and promote regulatory stability, predictability, and clarity. The proposed rule covers diverse topics, which result in recordkeeping and reporting requirements related to contents of applications, plant design and analysis, siting, construction and manufacturing, licensing-basis information, facility operations, programs, staffing, FFD, physical security, cyber-security, AA, decommissioning, and quality assurance. In addition to the new information collections in the proposed regulations, part 53 would result in new collections via NRC Forms 361S, 366, 366A, 366B, 893, and 894. NRC Forms 366, 366A, and 366B would be modified to include part 53 reportable events covering an equivalent scope as the requirements in 10 CFR 50.73, but without LWR-specific terminology to ensure technology inclusiveness. The proposed rule also would require part 53 licensees to use NRC Forms 893 and 894 to report on positive drug and alcohol test results (NRC Form 893) and annual fitness-for-duty program performance (NRC Form 894). Finally, a new version of NRC Form 361 (NRC Form 361S) would be created for use by part 53 licensees, covering an equivalent scope as the requirements in 10 CFR 50.72, but without LWR-specific terminology to ensure technology inclusiveness. The NRC is seeking public comment on the potential impact of the information collections contained in this proposed rule and on the following issues: 1. Is the proposed information collection necessary for the proper performance of the functions of the NRC, including whether the information will have practical utility? Please explain your response. 2. Is the estimate of the burden of the proposed information collection accurate? Please explain your response. 3. Is there a way to enhance the quality, utility, and clarity of the information to be collected? Please explain your response. 4. How can the burden of the proposed information collection on respondents be minimized, including the use of automated collection techniques or other forms of information technology? Please explain your response. The OMB clearance documents and proposed rule is available as indicated under the “Availability of Documents” section in this document or may be viewed free of charge by contacting the NRC's PDR reference staff at 1-800-397-4209, at 301-415-4737, or by email to 
                        *PDR.resource@nrc.gov.*
                         You may obtain information and comment submissions related to the OMB clearance package by searching on 
                        *http://www.regulations.gov*
                         under Docket ID NRC-2019-0062. You may submit comments on any aspect of these proposed information collections, including suggestions for reducing the burden and on the above issues, by the following methods: *Federal Rulemaking Website:*
                         Go to 
                        *http://www.regulations.gov*
                         and search for Docket ID NRC-2019-0062.
                    *Mail comments to:*
                         FOIA, Library, and Information Collections Branch, Office of the Chief Information Officer, Mail Stop: T6-A10M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 or by email to 
                        *Infocollects.Resource@nrc.gov*
                         or to the OMB reviewer at: OMB Office of Information and Regulatory Affairs (3150-XXXX, 3150-0002, -0104, -0146, -0238), Attn: Desk Officer for the Nuclear Regulatory Commission, 725 17th Street NW, Washington, DC 20503. Submit comments by December 2, 2024. Comments received after this date will be considered if it is practical to do so, but the NRC staff is able to ensure consideration only for comments received on or before this date. Public Protection Notification The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless the document requesting or requiring the collection displays a currently valid OMB control number. XV. Criminal Penalties For the purposes of Section 223 of the Act, the NRC is issuing this proposed rule that would add a new part 53 and amend parts 26 and 73 under one or more of Sections 161b, 161i, or 161o of the Act, except as noted in proposed § 53.9010(b) and § 26.825(b). Willful violations of the part 53 and part 26 regulations not listed in proposed § 53.9010(b) and § 26.825(b) would be subject to criminal enforcement. Criminal penalties as they apply to regulations in part 53 would be discussed in § 53.9010. XVI. Voluntary Consensus Standards The NTTAA requires that Federal agencies use technical standards that are developed or adopted by voluntary consensus standards bodies unless the use of such a standard is inconsistent with applicable law or otherwise impractical. In this proposed rule, the NRC would revise regulations by adding a risk-informed, technology-inclusive regulatory framework for commercial advanced nuclear reactors. This action does not constitute the establishment of a standard that contains generally applicable requirements. XVII. Availability of Guidance As discussed in section II, Background, of this document, the NRC's development of proposed part 53 
                        ( print page 87010)
                        built upon recent and ongoing activities such as those described in SECY-19-0117. Because a number of those activities are ongoing to support new reactor applications under the existing regulatory framework of 10 CFR parts 50 and 52, the NRC staff identified in its response to SRM-SECY-20-0032 that the timing of guidance document development to support the part 53 rulemaking was a key risk and uncertainty to publishing the final part 53 rule. To mitigate this risk, the NRC engaged external stakeholders to ensure a common prioritization of the development of these guidance documents and to work diligently on those that would be needed to support this rulemaking, forthcoming applications, or broader efforts such as the Advanced Reactor Demonstration Program being sponsored by the DOE. The NRC also recognizes that guidance development to support part 53 and advanced reactors will continue as the industry and NRC learn lessons from licensing reviews and operating experience. Therefore, the NRC categorized guidance supporting the part 53 rulemaking into three categories: (1) guidance issued or under development to support applications under the existing regulatory framework; (2) implementing guidance for part 53-specific proposed rule language; and (3) future guidance activities that would need to be completed after the part 53 proposed rule is published for public comment. (1) Hundreds of guidance documents exist for the current fleet of operating reactors. While some of the guidance is specific to LWR technologies, other guidance is technology inclusive in nature and should be considered, as appropriate, in the development of all licensing applications and NRC reviews. In addition, the NRC has undertaken efforts to incorporate or reference the most relevant guidance in its efforts to develop additional guidance for future advanced reactors. The NRC has issued the following guidance to support licensing reviews of advanced reactors under the existing regulatory framework that will continue to inform applicant development and NRC reviews under parts 50 and 52. Conforming changes to these guidance documents would be needed to ensure they are applicable under part 53. The NRC will issue revisions or part 53-related companions to these guidance documents for public comment after the publication of this proposed rule and then finalize and issue the guidance documents with or after the final part 53 rule. RG 1.233, “Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors”
                    RG 1.247 for trial use, “Acceptability of Probabilistic Risk Assessment Results for Non-Light-Water Reactor Risk-Informed Activities”
                    NUREG-2246, “Fuel Qualification for Advanced Reactors”
                    RG 1.87, Revision 2, “Acceptability of ASME Code, Section III, Division 5, “High Temperature Reactors”
                    RG 1.246, “Acceptability of ASME Code, Section XI, Division 2, `Requirements for Reliability And Integrity Management (RIM) Programs for Nuclear Power Plants,' for Non-Light Water Reactors” Also, the NRC continues to develop additional guidance to support licensing reviews of advanced reactors under the existing regulatory framework. Some of these guidance documents have been issued and others will be issued before the finalization of part 53 to support near-term applicants and NRC reviews. For example, the NRC has been and continues to be engaged with the DOE and industry to develop content of application guidance and other regulatory guidance for advanced reactors to support applications and subsequent operations under the existing regulatory framework. These guidance documents, such as the industry-led Technology-Inclusive Content of Application Project guidance found in NEI 21-07, Revision 1, and the NRC-led Advanced Reactor Content of Application Project (ARCAP) interim staff guidance (ISG) documents and NRC regulatory guidance endorsing NEI 21-07, Revision 1, will support developers in preparing advanced reactor applications. These guidance documents provide an overview of the information that should be included in an advanced reactor application, a review roadmap for the NRC with the principal purpose of ensuring consistency, quality, and uniformity of NRC reviews, and a well-defined base from which the NRC can evaluate proposed changes in the scope and requirements of reviews. While specific sections of the information are primarily aligned with the LMP methodology, as endorsed in RG 1.233, as one acceptable process for applicants to use when developing portions of an application, the concepts and general information may be used to inform the review of an application submitted using other traditional licensing approach methodologies (as applicable). Other sections of the information are generally applicable and independent of the methodology used to develop an advanced reactor application. The ARCAP ISGs provide references to numerous regulatory guidance documents that should be considered by both applicants and the NRC in developing and reviewing, respectively, advanced reactor applications. The NRC has issued the following documents separately from this proposed rule. The NRC may issue other, related guidance documents with or after the final part 53 rule. RG 1.253, “Guidance for a Technology Inclusive Content of Application Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors”
                    DANU-ISG-2022-01, “Advanced Reactor Content of Application Project, `Review of Risk-Informed, Technology-Inclusive Advanced Reactor Applications—Roadmap' ”
                    DANU-ISG-2022-02, “Advanced Reactor Content of Application Project Chapter 2, `Site Information' ”
                    DANU-ISG-2022-03, “Advanced Reactor Content of Application Project Chapter 9, `Control of Routine Plant Radioactive Effluents, Plant Contamination and Solid Waste' ”
                    DANU-ISG-2022-04, “Advanced Reactor Content of Application Project Chapter 10, `Control of Occupational Dose' ”
                    DANU-ISG-2022-05, “Advanced Reactor Content of Application Project Chapter 11, `Organization and Human-System Considerations' ”
                    DANU-ISG-2022-06, “Advanced Reactor Content of Application Project Chapter 12, `Post-Construction Inspection, Testing, and Analysis Program' ”
                    DANU-ISG-2022-07, “Advanced Reactor Content of Application Project, `Risk-Informed Inservice Inspection/Inservice Testing' ”
                    DANU-ISG-2022-08, “Advanced Reactor Content of Application Project, `Risk-Informed Technical Specifications' ”
                    DANU-ISG-2022-09, “Advanced Reactor Content of Application Project, `Risk-Informed, Performance-Based Fire Protection Program (for Operations)' ”
                    RG 1.242, “Performance-Based Emergency Preparedness for Small Modular Reactors, Non-Light-Water Reactors, and Non-Power Production or Utilization Facilities”
                    RG 4.7, “General Site Suitability Criteria for Nuclear Power Stations” 
 
 (2) The NRC is issuing for comment nine draft guidance documents for the implementation of the proposed requirements in this rulemaking. The guidance is available in ADAMS under the Accession Numbers as indicated under the “Availability of Documents” section in this document. Comments on this draft regulatory guidance may be submitted by the methods outlined in the 
                        **ADDRESSES**
                         section of this document. Interested persons may obtain information and comment submissions related to the draft guidance by searching on 
                        *http://www.regulations.gov*
                         under Docket ID NRC-2019-0062. DG-1413, “Technology-Inclusive Identification of Licensing Events for Commercial Nuclear Plants” This DG describes an acceptable approach for identifying licensing events that can be used to inform the design basis, licensing basis, and content of applications for commercial nuclear plants, including large LWRs and non-LWRs. It applies to nuclear power reactor designers, applicants, and licensees of commercial nuclear plants applying for permits, licenses, certifications, and approvals under parts 50, 52, and 53. In this DG, the term “licensing events” is used in a generic sense to refer to collections of designated event categories such as, but not limited to AOOs, DBAs, DBEs, and postulated accidents. Specifically, this DG provides an acceptable approach for: (1) conducting a comprehensive and systematic search for initiating events; (2) using a systematic process to delineate a comprehensive set of event sequences; (3) grouping initiating events and event sequences into designated licensing event categories; and (4) providing assurance that the set of licensing events is complete. DG-5073, “Fitness For Duty Programs for Commercial Nuclear Plants And Manufacturing Facilities Licensed Under10 CFR part 53” This DG describes guidance for applicants under part 53 and licensees and other entities described in § 26.3(f) who would elect to or be required to implement FFD programs for facilities licensed under part 53. The FFD program requirements would be detailed in subpart M of part 26 and involve, in part, policies, procedures, drug and alcohol testing, laboratory requirements, behavioral observation, MRO responsibilities, fitness determinations, reporting, and recordkeeping. The FFD program for facilities licensed under part 53 subject to part 26 would also include requirements for a PMRP and FFD program change control that licensees or other entities must implement to maintain an effective FFD program. DG-5074, “Access Authorization Program for Commercial Nuclear Plants” This DG describes a method that the staff considers acceptable to comply with requirements in proposed § 73.120, “Access authorization program for commercial nuclear plants,” related to an AA program. This document provides guidance and would be one NRC-approved method (not the only method) for meeting regulatory requirements for part 53. The proposed language in § 73.120 would provide flexibility through availability of the use of an alternate approach, commensurate with risk and consequence to public health and safety, for part 53 applicants who demonstrate in an analysis that the offsite consequences satisfy the criterion defined in proposed § 53.860(a)(2)(i). DG-5075, “Establishing Cybersecurity Programs for Commercial Nuclear Plants Licensed Under10 CFR part 53” This DG describes an approach the NRC staff deems acceptable for complying with the Commission's proposed regulations for establishing, implementing, and maintaining a cybersecurity program at commercial nuclear plants that would be licensed under part 53. This guidance provides an approach for meeting the requirements of proposed § 73.110, “Technology-inclusive requirements for protection of digital computer and communication systems and networks.” DG-5076, “Guidance for Technology Inclusive Requirements for Physical Protection of Licensed Activities at Commercial Nuclear Plants” This DG describes methods and approaches that the NRC staff considers acceptable for meeting the proposed physical security requirements of part 53 and § 73.100. The guidance is intended to provide methods and considerations for complying with § 53.440(f) safety and security design process considerations, determining eligibility for meeting the performance criterion in § 53.860 to relieve the applicant from the applicable requirements to defend against radiological sabotage outlined in § 73.55 or § 73.100, and (if the required analysis for eligibility is not satisfied) applying the physical security requirements of § 73.100 as an alternative pathway from § 73.55 for protection against radiological sabotage. DG-5078, “Fatigue Management for Nuclear Power Plant Personnel at Commercial Nuclear Plants Licensed Under10 CFR part 53” This DG describes proposed methods that the NRC staff considers acceptable for addressing certain aspects of FFD programs that would be established at commercial nuclear facilities licensed under part 53. This guidance, in conjunction with the existing RG 5.73, “Fatigue Management for Nuclear Plant Personnel,” would provide comprehensive guidance regarding acceptable methods for the development and implementation of licensee fatigue-management programs. The NRC is issuing for public comment the following draft ISG documents for the implementation of NRC staff review of applications under the proposed requirements in this rulemaking: DRO-ISG-2023-01, “Operator Licensing Programs” This draft ISG provides guidance for the review of tailored operator licensing programs that are submitted for review consistent with the technical requirements of proposed § 53.730(g). This guidance primarily addresses the review of operator licensing examination processes to facilitate the ability of reviewers to assess whether a proposed approach to the testing of licensed operators and trainees reflects sound assessment testing practices that are suitable for the screening of competent licensed operators. Additionally, this ISG provides further review guidance in other areas such as licensed operator continuing training and proficiency programs. DRO-ISG-2023-02, “Interim Staff Guidance Augmenting NUREG-1791, `Guidance for Assessing Exemption Requests from the Nuclear Power Plant Licensed Operator Staffing Requirements Specified in10 CFR 50.54(m),' for Licensing Commercial Nuclear Plants under 10 CFR part 53” This draft ISG provides guidance for the review of customized facility operator staffing plans that are submitted for review consistent with the technical requirements of proposed § 53.730(f). This ISG is structured as a companion document to the existing NUREG-1791 and adapts the existing HFE-based methodologies of that document for use in the evaluation of staffing plans that would be submitted within the context of part 53 facilities. Additionally, this ISG provides further guidance to address other staffing-related considerations, such as provisions for engineering expertise. DRO-ISG-2023-03, “Development of Scalable Human Factors Engineering Review Plans” 
 
 This draft ISG applies to the HFE review of applications for OLs, COLs, DCs, and standard design approvals for commercial nuclear plants submitted under proposed part 53. The purpose of this ISG is to facilitate NRC understanding of an acceptable method for developing a scalable (
                        *i.e.,*
                         application-specific) plan for the review of these applications for compliance with applicable HFE requirements. The ISG describes a process and provides implementation guidance for the NRC to tailor HFE review plans to each application to achieve an effective and efficient review. (3) The NRC has identified future guidance activities that would need to be completed after the part 53 proposed rule is published for public comment to support advanced reactor applications and NRC reviews. For example, the NRC recognizes that new guidance would be needed for the implementation of provisions in proposed § 53.620(d) and the associated licensing provisions in proposed subpart H that would allow and establish requirements for the loading of fuel into a manufactured reactor for subsequent transport to and use at a commercial nuclear plant that will operate the facility pursuant to a COL. The NRC has not yet initiated the development of guidance documents in this category but will engage stakeholders during the development of these documents to ensure common prioritization. In addition, the NRC works with standards development organizations, advanced reactor developers, DOE, and other stakeholders to identify and facilitate new consensus codes and standards needed for advanced reactor development. The NRC will continue its membership and participation on standards development committees and working groups to support standards for advanced reactor technologies, where appropriate. XVIII. Public Meeting The NRC will conduct a public meeting on this proposed rule for the purpose of describing the proposed rule and implementation guidance to the public and answering questions from the public on the proposed rule and implementation guidance. The NRC will publish a notice of the public meeting's location, time, and agenda on the NRC's public meeting website at least 10 calendar days before the meeting. Stakeholders should monitor the NRC's public meeting website for information about the public meeting at: 
                        *https://www.nrc.gov/​public-involve/​public-meetings/​index.cfm.* XIX. Availability of Documents The documents identified in the following table are available to interested persons through one or more of the following methods, as indicated. Document
ADAMS accession No./Web link/
                                 
                                    **Federal Register**
                                     Citation**Proposed Rule Documents****Federal Register**
                                 Notification, “Proposed Rule: Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors,” October, 2024
ML24095A161.“Draft Environmental Assessment for the Proposed Rule—Risk Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors,” October, 2024
ML24095A163.“Draft Regulatory Analysis for the Proposed Rule: Risk Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors,” October, 2024
ML24095A166.**Information Collection Documents**Draft Supporting Statement for Information Collection Analysis—10 CFR Part 53
ML21162A109.Draft Supporting Statement for Information Collection Analysis—10 CFR Part 26
ML23030A400.Draft Supporting Statement for Information Collection Analysis—10 CFR Part 50
ML24220A036.Draft Supporting Statement for Information Collection Analysis—10 CFR Part 73
ML23030A576.Draft Supporting Statement for Information Collection Analysis—NRC Form 361S
ML24220A034.Draft Supporting Statement for Information Collection Analysis—NRC Form 366
ML24220A035.Draft Supporting Statement for Information Collection Analysis—NRC Form 893 and 894
ML24220A033.Proposed Rule—Part 26 Burden Tables for Risk Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors
ML24240A008.Proposed Rule—Part 50 Burden Tables for Risk Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors
ML24220A061.Proposed Rule—Part 53 Burden Tables for Risk Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors
ML24220A060.Proposed Rule—Part 73 Burden Tables for Risk Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors
ML24240A009.Draft NRC Form 361S, “Part 53 Plant Event Notification Worksheet”
ML23032A443.Draft NRC Form 366, “Licensee Event Report (LER)”
ML23032A445.Draft NRC Form 366A, “Licensee Event Report (LER) Continuation Sheet”
ML23032A447.Draft NRC Form 366B, “Licensee Event Report (LER) (Failure Continuation)”
ML23032A454.Draft NRC Form 893, “10 CFR Part 26, Subpart M, Single FFD Policy Violation Form”
ML23032A435.Draft NRC Form 894, “10 CFR Part 26, Subpart M, Annual Reporting Form for FFD Performance Information”
ML23032A439.**Draft Regulatory Guidance Documents**DG-1413, “Technology-Inclusive Identification Of Licensing Events For Commercial Nuclear Plants,” October, 2024
ML22257A173.DG-5073, “Fitness-For-Duty Programs For Commercial Nuclear Plants And Manufacturing Facilities Licensed Under 10 CFR Part 53,” October, 2024
ML22200A037.DG-5074, “Access Authorization Program for Commercial Nuclear Plants,” October, 2024
ML22199A246.DG-5075, “Establishing Cybersecurity Programs For Commercial Nuclear Plants Licensed Under 10 CFR Part 53,” October, 2024
ML22199A257.DG-5076, “Guidance for Technology Inclusive Requirements for Physical Protection of Licensed Activities at Commercial Nuclear Plants,” October, 2024
ML22203A131.( print page 87013)DG-5078, “Fatigue Management For Nuclear Power Plant Personnel At Commercial Nuclear Plants Licensed Under 10 CFR Part 53,” October, 2024
ML22264A109.**Draft ISG Documents**Draft ISG DRO-ISG-2023-01, “Operator Licensing Programs,” October, 2024
ML22266A066.Draft ISG DRO-ISG-2023-02, “Interim Staff Guidance Augmenting NUREG-1791, `Guidance for Assessing Exemption Requests from the Nuclear Power Plant Licensed Operator Staffing Requirements Specified in 10 CFR 50.54(m),' for Licensing Commercial Nuclear Plants under 10 CFR Part 53,” October, 2024
ML22266A068.Draft ISG DRO-ISG-2023-03, “Development of Scalable Human Factors Engineering Review Plans,” October, 2024
ML22266A072.**Other References**American National Standards Institute/ANS-3.4-2013, “Medical Certification And Monitoring Of Personnel Requiring Operator Licenses For Nuclear Power Plants”
*https://webstore.ansi.org/​Standards/​ANSI/​ansians2013.*ASME/ANS RA-S-1.4-2021, “Probabilistic Risk Assessment Standard for Advanced Non-Light Water Reactor Nuclear Power Plants”
*https://www.asme.org/​codes-standards/​find-codes-standards/​probabilistic-risk-assessment-standard-for-advanced-non-light-water-reactor-nuclear-power-plants/​2021/​pdf.*ASCE/SEI 43-19, “Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities”
*https://doi.org/​10.1061/​9780784415405.***Federal Register**
                                 notification—Final policy statement, “Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities; Final Policy Statement,” dated August 16, 1995
60 FR 42622.**Federal Register**
                                 notification—Final rule, “Fitness-for-Duty Programs,” dated June 7, 1989
54 FR 24468.**Federal Register**
                                 notification—Final rule, “Fitness for Duty Programs,” dated March 31, 2008
73 FR 16966.**Federal Register**
                                 notification—Final rule, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” dated August 28, 2007
72 FR 49352.**Federal Register**
                                 notification—Final rule, “Station Blackout,” dated June 21, 1988
53 FR 23203.**Federal Register**
                                 notification—Final rule, “Technical Specifications,” dated July 19, 1995
60 FR 36953, 36955.**Federal Register**
                                 notification—Guidance, “Mandatory Guidelines for Federal Workplace Drug Testing Programs,” dated January 23, 2017
82 FR 7920.**Federal Register**
                                 notification—Guidance, “Mandatory Guidelines for Federal Workplace Drug Testing Programs—Oral/Fluid,” dated October 25, 2019
84 FR 57554.**Federal Register**
                                 notification—Policy Statement, “Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants,” dated August 8, 1985
50 FR 32138.**Federal Register**
                                 notification—Policy Statement, “Safety Goals for the Operation of Nuclear Power Plants; Policy Statement; Correction and Republication,” dated August 21, 1986
51 FR 30028.**Federal Register**
                                 notification—Policy Statement, “Tribal Policy Statement,” dated January 9, 2017
82 FR 2402.**Federal Register**
                                 notification—Policy Statement, “Policy Statement on the Regulation of Advanced Reactors,” dated October 14, 2008
73 FR 60612.**Federal Register**
                                 notification—Policy Statement, “Final Safety Culture Policy Statement,” dated June 14, 2011
76 FR 34773.**Federal Register**
                                 notification—Proposed rule, “Emergency Preparedness for Small Modular Reactors and Other New Technologies,” dated May 12, 2020
85 FR 28436.**Federal Register**
                                 notification—Proposed rule, “Regulatory Improvements for Production and Utilization Facilities Transitioning to Decommissioning,” dated March 3, 2022
87 FR 12254.**Federal Register**
                                 notification—Public meeting, “Reporting Requirements for Nonemergency Events at Nuclear Power Plants,” dated November 29, 2021
86 FR 67669.ICRP, Publication 2 “Permissible dose for internal radiation,” dated 1960
*https://www.icrp.org/​publication.asp?​id=​icrp%20publication%202.*ICRP, Publication 26 “Recommendations of the ICRP,” dated 1977
*https://www.icrp.org/​publication.asp?​id=​ICRP%20Publication%2026.*ICRP, Publication 30 “Limits for Intakes of Radionuclides by Workers,” dated 1979
*https://www.icrp.org/​publication.asp?​id=​ICRP%20Publication%2030%20(Index).*Letter to Chairman Hanson, NRC, “Final Letter on Draft 10 CFR Part 53 Rulemaking Language,” dated November 22, 2022
ML22319A104.Letter to Chairman Hanson, NRC, “Fourth Interim Letter on 10 CFR Part 53 Rulemaking Language,” dated August 2, 2022
ML22196A292.Letter to Chairman Hanson, NRC, “Preliminary Proposed Rule Language For 10 CFR Part 53, Regulation of Advanced Nuclear Reactors, Interim Report,” dated May 30, 2021
ML21140A354.Letter to Chairman Hanson, NRC, “Preliminary Rule Language For 10 CFR Part 53, Subpart F, `Requirements for Operations,' Interim Report,” dated February 17, 2022
ML22040A361.Letter to Chairman Rempe, ACRS, “Response to the Advisory Committee on Reactor Safeguards, `Fourth Interim Letter on 10 CFR Part 53 Rulemaking Language,'” dated September 30, 2022
ML22249A073.Letter to Chairman Rempe, ACRS, “Response to the Advisory Committee on Reactor Safeguards Letter on Preliminary Rule Language for 10 CFR Part 53, Subpart F, `Requirements for Operations,' Interim Report,” dated March 30, 2022
ML22063A012.Letter to Chairman Sunseri, ACRS, “Part 53, Licensing and Regulation of Advanced Nuclear Reactors,” dated November 24, 2020
ML20311A006.( print page 87014)Letter to Chairman Svinicki, NRC, “10 CFR Part 53, Licensing and Regulation of Advanced Nuclear Reactors,” dated October 21, 2020
ML20295A647.*Michigan*
                                 v. 
                                *EPA,*
                                 135 S. Ct. 2699 (2015)National Library of Medicine, National Institutes of Health, Workshop Summary, “The Evolution of Telehealth: Where Have We Been and Where Are We Going?,” dated November 2012
*https://www.ncbi.nlm.nih.gov/​books/​NBK207141/​.*NEI 18-04, Rev. 1, “Risk-Informed Performance-Based Technology-Inclusive Guidance for Non-Light Water Reactors,” dated August 2019
ML19241A472.NIA, “Clarifying `Major Portions' of a Reactor Design in Support of a Standard Design Approval,” dated April 2017
*https://www.nuclearinnovationalliance.org/​clarifying-major-portions-reactor-design-support-standard-design-approval.*NRC, “A Regulatory Review Roadmap for Non-Light Water Reactors,” dated December 2017
ML17312B567.NRC, “Manufacturing License ML-1 for Production of Up to Eight Floating Nuclear Plants,” dated September 30, 1982
ML20070J215.NRC, “Risk-Informed and Performance-Based Human-System Considerations for Advanced Reactors,” dated March 2021
ML21069A003.NRC Form 890, “Single Positive Test Form”
ML22013B187.NRC Form 891, “Annual Reporting for Drug and Alcohol Tests”
ML22013B240.NRC From 892, “Annual Fatigue Reporting Form”
ML22013B250.NUREG-0654/FEMA-REP-1, Revision 2, “Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants,” dated December 2019
ML19347D139.NUREG-0880, “Safety Goals for Nuclear Power Plant Operation,” dated May 1983
ML071770230.NUREG-1530, Revision 1, “Reassessment of NRC's Dollar Per Person-Rem Conversion Factor Policy, Final Report,” dated February 2022
ML22053A025.NUREG/BR-0058, Revision 5, “Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commission,” dated April 2017
ML17100A480.NUREG/CR-5884, “Revised Analyses of Decommissioning for the Reference Pressurized Water Reactor Power Station,” dated November 1995
ML14008A187.NUREG/CR-6187, Volume 1, “Revised Analyses of Decommissioning for the Reference Boiling Water Reactor Power Station,” dated July 1996
ML14008A186.OMB Circular No. A-119, “Federal Participation in the Development and Use of Voluntary Consensus Standards and in Conformity Assessment Activities,” dated February 19, 1998
*https://obamawhitehouse.archives.gov/​omb/​circulars_​a119_​a119fr.*PNNL, Technical Letter Report, “The Use of Electronic Communications to Perform Determinations of Fitness,” dated August 2017
ML18081A607.Pre-decisional DG, “Technology-Inclusive, Risk-Informed, and Performance-Based Methodology for Seismic Design of Commercial Nuclear Plants,” dated October 3, 2022
ML22276A149.Research Information Letter 2021-04, “Feasibility Study on a Potential Consequence-Based Seismic Design Approach for Nuclear Facilities,” dated April 2021
ML21113A066.RG 1.110, Revision 1, “Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors,” dated October 2013
ML13241A052.RG 1.134, Revision 4, “Medical Assessment Of Licensed Operators Or Applicants For Operator Licenses At Nuclear Power Plants,” dated September 2014
ML14189A385.RG 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,” Revision 3, dated January 2018
ML17317A256.RG 1.208, “A Performance-Based Approach to Define the Site-Specific Earthquake Ground Motion,” dated March 2007
ML070310619.RG 1.232, “Guidance for Developing Principal Design Criteria for Non-Light-Water Reactors,” Revision 0, dated April 2018
ML17325A611.RG 1.233, Revision 0, “Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors,” dated June 2020
ML20091L698.RG 1.247, “Acceptability of Probabilistic Risk Assessment Results for Non-Light-Water Reactor Risk-Informed Activities,” issued March 2022 for trial use
ML21235A008.RG 5.73, “Fatigue Management for Nuclear Power Plant Personnel,” dated March 20, 2009
ML083450028.RG 5.77, “Insider Mitigation Program,” Revision 1, dated September 08, 2022
ML16342B024.RG 5.81, “Target Set Identification and Development for Nuclear Power Reactors,” Revision 1, dated December 2019 (non-public)
ML13151A355.SECY-18-0096, “Functional Containment Performance Criteria For Non-Light-Water-Reactors,” dated September 28, 2018
ML18115A157.SECY-19-0117, “Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors,” dated December 2019
ML18311A264 (package).SECY-20-0032, “Rulemaking Plan on `Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors (RIN-3150-AK31; NRC-2019-0062,'” dated April 13, 2020
ML19340A056.SECY-20-0070, “(Redacted) Technical Evaluation of the Security Bounding Time Concept for Operating Nuclear Power Plants,” dated November 8, 2021
ML20126G265 (package).SECY-22-0072, “Proposed Rule: Alternative Physical Security Requirements for Advanced Reactors (RIN 3150-AK19),” dated August 2, 2022
ML21334A003 (package).SECY-83-293, “Amendments to 10 CFR 50 Related to Anticipated Transients Without Scram (ATWS) Events,” dated July 19, 1983
ML21278A823 (non-public); ML21278A994 (non-public).SECY-93-092, “Issues Pertaining to the Advanced Reactor (PRISM, MHTGR, and PIUS) and CANDU 3 Designs and their Relationship to Current Regulatory Requirements,” dated April 8, 1993
ML040210725.( print page 87015)SRM-SECY-10-0121, “Modifying the Risk-Informed Regulatory Guidance for New Reactors,” dated March 2, 2011
ML110610166.SRM-SECY-17-0100, “Security Baseline Inspection Program Assessment Results and Recommendations for Program Efficiencies,” dated October 8, 2018
ML18283A072.SRM-SECY-20-0032, “Rulemaking Plan on `Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors (RIN-3150-AK31; NRC-2019-0062),'” dated October 2, 2020
ML20276A293.SRM-SECY-20-0045, “Population Related Siting Considerations for Advanced Reactors,” dated July 30, 2022
ML22194A885.SRM-SECY-98-144, “Staff Requirements—SECY-98-144—White Paper on Risk-Informed and Performance-Based Regulations,” dated February 24, 1999
ML003753593.SECY-23-0021, “Proposed Rule: Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors (RIN 3150-AK31),” March 1, 2023
ML21162A095.SECY-23-0021, Enclosure 1, “Draft 
                                **Federal Register**
                                 Notification”
ML21162A102.SECY-23-0021, Enclosure 2, “Draft Environmental Assessment for the Proposed Rule—Risk Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors”
ML21162A104.SECY-23-0021, Enclosure 3, “Draft Regulatory Analysis for the Proposed Rule: Risk Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors”
ML21165A112.SECY-23-0021, Enclosure 4, “Alternative Approaches Considered for Selected Topics During the Development of 10 CFR Part 53”
ML22244A001.SECY-23-0021, Enclosure 5, “Estimated Resources for The Risk-Informed, Technology-Inclusive Regulatory Framework For Advanced Reactors Rulemaking”
ML22304A099 (non-public).Staff Requirements—SECY-23-0021, “Proposed Rule: Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors (RIN 3150-AK31),” March 4, 2024
ML24064A047 (package).
 Throughout the development of this rule, the NRC may post documents related to this rule, including public comments, on the Federal rulemaking website at 
                        *https://www.regulations.gov*
                         under Docket ID NRC-2019-0062. The Federal rulemaking website allows you to receive alerts when changes or additions occur in a docket folder. To subscribe: (1) Navigate to the docket folder (NRC-2019-0062-0012); (2) click the “Sign up for Email Alerts” link; and (3) enter your email address and select how frequently you would like to receive emails (daily, weekly, or monthly). List of Subjects 10 CFR Part 1 Flags
Organization and functions (Government Agencies)
Seals and insignia 10 CFR Part 2 Administrative practice and procedure
Antitrust
Byproduct material
Classified information
Confidential business information
Freedom of information
Environmental protection
Hazardous waste
Nuclear energy
Nuclear materials
Nuclear power plants and reactors
Penalties
Reporting and recordkeeping requirements
Sex discrimination
Source material
Special nuclear material
Waste treatment and disposal 10 CFR Part 10 Administrative practice and procedure
Classified information
Government employees
Security measures 10 CFR Part 11 Hazardous materials transportation
Investigations
Nuclear energy
Nuclear materials
Penalties
Reporting and recordkeeping requirements
Security measures
Special nuclear material 10 CFR Part 19 Criminal penalties
Environmental protection
Nuclear Energy
Nuclear materials
Nuclear power plants and reactors
Occupational safety and health
Penalties
Radiation protection
Reporting and recordkeeping requirements
Sex discrimination 10 CFR Part 20 Byproduct material
Criminal penalties
Hazardous waste
Licensed material
Nuclear energy
Nuclear materials
Nuclear power plants and reactors
Occupational safety and health
Packaging and containers
Penalties
Radiation protection
Reporting and recordkeeping requirements
Source material
Special nuclear material
Waste treatment and disposal 10 CFR Part 21 Nuclear power plants and reactors
Penalties
Radiation protection
Reporting and recordkeeping requirements 10 CFR Part 25 Classified information
Criminal penalties
Investigations
Penalties
Reporting and recordkeeping requirements
Security measures 10 CFR Part 26 Administrative practice and procedure
Alcohol abuse
Alcohol testing
Appeals
Drug abuse
Drug testing
Employee assistance programs
Fitness for duty
Management actions
Nuclear power plants and reactors
Privacy
Protection of information
Radiation protection
Reporting and recordkeeping requirements 10 CFR Part 30 Byproduct material
Criminal penalties
Government contracts
Intergovernmental relations
Isotopes
Nuclear energy
Nuclear materials
Penalties
Radiation protection
Reporting and recordkeeping requirements
Whistleblowing 10 CFR Part 40 Criminal penalties
Exports
Government contracts
Hazardous materials transportation
Hazardous waste
Nuclear energy
Nuclear materials
Penalties
Reporting and recordkeeping requirements
Source material
Uranium
Whistleblowing 10 CFR Part 50 Administrative practice and procedure
Antitrust
Backfitting
Classified information
Criminal penalties
Education
Emergency planning
Fire prevention
Fire protection
Intergovernmental relations
Nuclear power plants and reactors
Penalties
Radiation protection
Reactor siting criteria
Reporting and 10 CFR Part 51 Administrative practice and procedure
Environmental impact statements
Hazardous waste
Nuclear energy
Nuclear materials
Nuclear power plants and reactors
Reporting and recordkeeping requirements 10 CFR Part 53 Administrative practice and procedure
Antitrust
Backfitting
Construction permit
Combined license
Classified information
Criminal penalties
Early site permit
Emergency planning
Fees
Fire prevention
Fire protection
Inspection
Intergovernmental relations
Limited work authorization
Manufacturing license
Nuclear power plants and reactors
Operating license
Penalties
Prototype
Radiation protection
Reactor siting criteria
Reporting and recordkeeping requirements
Standard design
Standard design certification
Training programs 10 CFR Part 70 Classified information
Criminal penalties
Emergency medical services
Hazardous materials transportation
Material control and accounting
Nuclear energy
Nuclear materials
Packaging and containers
Penalties
Radiation protection
Reporting and recordkeeping requirements
Scientific equipment
Security measures
Special nuclear material
Whistleblowing 10 CFR Part 72 Administrative practice and procedure
Hazardous waste
Indians
Intergovernmental relations
Nuclear energy
Penalties
Radiation protection
Reporting and recordkeeping requirements
Security measures
Spent fuel
Whistleblowing 10 CFR Part 73 Criminal penalties
Exports
Hazardous materials transportation
Imports
Nuclear energy
Nuclear materials
Nuclear power plants and reactors
Penalties
Reporting and recordkeeping requirements
Security measures 10 CFR Part 74 Accounting
Criminal penalties
Hazardous materials transportation
Material control and accounting
Nuclear energy
Nuclear materials
Packaging and containers
Penalties
Radiation protection
Reporting and recordkeeping requirements
Scientific equipment
Special nuclear material 10 CFR Part 75 Criminal penalties
Intergovernmental relations
Nuclear energy
Nuclear materials
Nuclear power plants and reactors
Penalties
Reporting and recordkeeping requirements
Security measures
Treaties 10 CFR Part 95 Classified information
Criminal penalties
Penalties
Reporting and recordkeeping requirements
Security measures 10 CFR Part 140 Criminal penalties
Extraordinary nuclear occurrence
Insurance
Intergovernmental relations
Nuclear materials
Nuclear power plants and reactors
Penalties
Reporting and recordkeeping requirements 10 CFR Part 150 Criminal penalties
Hazardous materials transportation
Intergovernmental relations
Nuclear energy
Nuclear materials
Penalties
Reporting and recordkeeping requirements
Security measures
Source material
Special nuclear material 10 CFR Part 170 Byproduct material
Import and export licenses
Intergovernmental relations
Non-payment penalties
Nuclear energy
Nuclear materials
Nuclear power plants and reactors
Source material
Special nuclear material 10 CFR Part 171 Annual charges
Approvals
Byproduct material
Holders of certificates
Intergovernmental relations
Nonpayment penalties
Nuclear materials
Nuclear power plants and reactors
Registrations
Source material
Special nuclear material
 For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended; the Energy Reorganization Act of 1974, as amended; and 5 U.S.C. 552 and 553, the NRC is proposing the following amendments to 10 CFR parts 1, 2, 10, 11, 19, 20, 21, 25, 26, 30, 40, 50, 51, 70, 72, 73, 74, 75, 95, 140, 150, 170, and 171 and adding 10 CFR part 53: PART 1—STATEMENT OF ORGANIZATION AND GENERAL INFORMATION
 1.The authority citation for part 1 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 23, 25, 29, 161, 191 (42 U.S.C. 2033, 2035, 2039, 2201, 2241); Energy Reorganization Act of 1974, secs. 201, 203, 204, 205, 209 (42 U.S.C. 5841, 5843, 5844, 5845, 5849); Administrative Procedure Act (5 U.S.C. 552, 553); Reorganization Plan No. 1 of 1980, 5 U.S.C. Appendix (Reorganization Plans). § 1.43
 2.In § 1.43, in paragraph (a)(2) remove the cross reference “ 10 CFR parts 50, 52, and 54” and add in its place the cross reference “10 CFR parts 50, 52, 53, and 54”. PART 2—AGENCY RULES OF PRACTICE AND PROCEDURE
 3.The authority citation for part 2 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 29, 53, 62, 63, 81, 102, 103, 104, 105, 161, 181, 182, 183, 184, 186, 189, 191, 234 (42 U.S.C. 2039, 2073, 2092, 2093, 2111, 2132, 2133, 2134, 2135, 2201, 2231, 2232, 2233, 2234, 2236, 2239, 2241, 2282); Energy Reorganization Act of 1974, secs. 201, 206 (42 U.S.C. 5841, 5846); Nuclear Waste Policy Act of 1982, secs. 114(f), 134, 135, 141 (42 U.S.C. 10134(f), 10154, 10155, 10161); Administrative Procedure Act (5 U.S.C. 552, 553, 554, 557, 558); National Environmental Policy Act of 1969 (42 U.S.C. 4332); 44 U.S.C. 3504 note. Section 2.205(j) also issued under 28 U.S.C. 2461 note. § 2.1
 4.In § 2.1, in paragraph (e) remove the phrase “part 52” and add in its place the phrase “part 52 or part 53”. 5.In § 2.4, revise the definitions for “ *Contested proceeding*
                        ” and “
                        *Facility*
                        ” to read as follows: § 2.4
  
 *Contested proceeding*
                             means— (1) A proceeding in which there is a controversy between the NRC staff and the applicant for a license or permit concerning the issuance of the license or permit or any of the terms or conditions thereof; (2) A proceeding in which the NRC is imposing a civil penalty or other enforcement action, and the subject of the civil penalty or enforcement action is an applicant for or holder of a license or permit, or is or was an applicant for or holder of a license or permit, or is or was an applicant for a standard design certification under part 52 or part 53 of this chapter; and (3) A proceeding in which a petition for leave to intervene in opposition to an application for a license or permit has been granted or is pending before the Commission. 
 *Facility*
                             means production facility or a utilization facility as defined in §§ 50.2 and 53.020 of this chapter.
 § 2.100
 6.In § 2.100, remove the phrase “subpart E of part 52” and add in its place the phrase “subpart E of part 52 or subpart H of part 53”. 7.In § 2.101, revise paragraphs (a)(3)(i), (a)(5), (a)(9) introductory text and paragraph (a)(9)(i) to read as follows: § 2.101
  (a) * * * (3) * * * (i) Submit to the Director, Office of Nuclear Reactor Regulation, or Director, Office of Nuclear Material Safety and Safeguards, as appropriate, such additional copies as the regulations in part 50, subpart A of part 51, and part 53 of this chapter require; 
 (5) An applicant for a construction permit under parts 50 or 53 of this chapter or a combined license under parts 52 or 53 of this chapter for a production or utilization facility which is subject to § 51.20(b) of this chapter, and is of the type specified in § 50.21(b)(2) or (b)(3); or § 50.22; or part 53, as applicable, of this chapter, or is a testing facility, may submit the information required of applicants by parts 50, 52, or 53 of this chapter in two parts. One part shall be accompanied by the information required by § 50.30(f) of this chapter, § 52.80(b) of this chapter, or § 53.1100(f) of this chapter, as applicable. The other part shall include any information required by § 50.34(a) and, if applicable, § 50.34a of this chapter; or §§ 52.79 and 52.80(a) of this chapter; or §§ 53.1109, 53.1306, 53.1309, and 53.1312 of this chapter; or §§ 53.1109, 53.1413, 53.1416, and 53.1419 of this chapter, as applicable. One part may precede or follow other parts by no longer than 6 months. If it is determined that either of the parts as described above is incomplete and not acceptable for processing, the Director, Office of Nuclear Reactor Regulation, or Director, Office of Nuclear Material Safety and Safeguards, as appropriate, will inform the applicant of this determination and the respects in which the document is deficient. Such a determination of completeness will generally be made within a period of 30 days. Whichever part is filed first shall also include the fee required by § 50.30(e) or § 53.1100(e) and § 170.21 of this chapter and the information required by §§ 50.33, 50.34(a)(1), and 52.79(a)(1) of this chapter; or §§ 53.1109, 53.1309, and 53.1416 of this chapter, as applicable, and § 50.37 or § 53.1115, as applicable, of this chapter. The Director, Office of Nuclear Reactor Regulation, or Director, Office of Nuclear Material Safety and Safeguards, as appropriate, will accept for docketing an application for a construction permit under part 50 or part 53 of this chapter or a combined license under parts 52 or 53 of this chapter for a production or utilization facility which is subject to § 51.20(b) of this chapter, and is of the type specified in § 50.21(b)(2) or (b)(3), or § 50.22, or part 53, as applicable, of this chapter or is a testing facility where one part of the application as described above is complete and conforms to the requirements of part 50 of this chapter. The additional parts will be docketed upon a determination by the Director, Office of Nuclear Reactor Regulation, or Director, Office of Nuclear Material Safety and Safeguards, as appropriate, that it is complete. 
 (9) An applicant for a construction permit for a utilization facility which is subject to § 51.20(b) of this chapter and is of the type specified in § 50.21(b)(2) or (b)(3), or § 50.22, or part 53 of this chapter, an applicant for or holder of an early site permit under part 52 or part 53 of this chapter, or an applicant for a combined license under parts 52 or 53 of this chapter, who seeks to conduct the activities authorized under § 50.10(d) or § 53.1130 of this chapter may submit a complete application under paragraphs (a)(1) through (a)(4) of this section which includes the information required by § 50.10(d) or § 53.1130 of this chapter. Alternatively, the applicant (other than an applicant for or holder of an early site permit) may submit its application in two parts: (i) Part one must include the information required by § 50.33(a) through (f) or § 53.1109(a) through (e) and § 53.1306 of this chapter, and the information required by § 50.10(d)(2) and (d)(3) or § 53.1130(a)(2) and (a)(3) of this chapter, as applicable.
 8.In § 2.104, revise paragraph (a) to read as follows: § 2.104
  (a) In the case of an application on which a hearing is required by the Act or this chapter, or in which the Commission finds that a hearing is required in the public interest, the Secretary will issue a notice of hearing to be published in the 
                            **Federal Register**
                            . The notice must be published at least 15 days, and in the case of an application concerning a limited work authorization, construction permit, early site permit, or combined license for a facility of the type described in § 50.21(b) or 50.22, or subpart H of part 53 of this chapter, as applicable, or a testing facility, at least 30 days, before the date set for hearing in the notice.1
                             In addition, in the case of an application for a limited work authorization, construction permit, early site permit, or combined license for a facility of the type described in § 50.22 or subpart H of part 53 of this chapter, as applicable, or a testing facility, the notice must be issued as soon as practicable after the NRC has docketed the application. If the Commission decides, under § 2.101(a)(2), to determine the acceptability of the application based on its technical adequacy as well as completeness, the notice must be issued as soon as practicable after the application has been tendered. 
 1
                                 If the notice of hearing concerning an application for a limited work authorization, construction permit, early site permit, or combined license for a facility of the type described in § 50.21(b) or § 50.22, or subpart H of part 53 of this chapter, as applicable, or a testing facility, does not specify the time and place of initial hearing, a subsequent notice will be published in the 
                                **Federal Register**
                                 which will provide at least 30-day notice of the time and place of that hearing. After this notice is given, the presiding officer may reschedule the commencement of the initial hearing for a later date or reconvene a recessed hearing without again providing at least 30-day notice.
 9.In § 2.105, revise paragraph (a) introductory text and paragraphs (a)(4), (a)(10), (a)(12), (a)(13), (b)(3) introductory text, (b)(3)(i), (ii), and (iv) to read as follows: § 2.105
  (a) If a hearing is not required by the Act or this chapter, and if the Commission has not found that a hearing is in the public interest, it will, before acting thereon, publish in the 
                            **Federal Register**
                            , as applicable, or on the NRC's website, 
                            *http://www.nrc.gov,*
                             or both, at the Commission's discretion, either a notice of intended operation under § 52.103(a) or § 53.1452(a) of this chapter, as applicable, and a proposed finding that inspections, tests, analyses, and acceptance criteria for a combined license under subpart C of part 52 or under subpart H of part 53 of this chapter, have been or will be met, or a notice of proposed action with respect to an application for: 
 (4) An amendment to an operating license, combined license, or manufacturing license for a facility licensed under § 50.21(b) or § 50.22 or under subpart H of part 53 of this chapter, as applicable, or for a testing facility, as follows: (i) If the Commission determines under § 50.58 or § 53.1515 of this 
                            ( print page 87018)
                            chapter that the amendment involves no significant hazards consideration, though it will provide notice of opportunity for a hearing pursuant to this section, it may make the amendment immediately effective and grant a hearing thereafter; or (ii) If the Commission determines under §§ 50.58 and 50.91 or § 53.1515 of this chapter, as applicable, that an emergency situation exists or that exigent circumstances exist and that the amendment involves no significant hazards consideration, it will provide notice of opportunity for a hearing pursuant to § 2.106 (if a hearing is requested, it will be held after issuance of the amendment); 
 (10) In the case of an application for an operating license for a facility of a type described in § 50.21(b) or § 50.22, or part 53 of this chapter or a testing facility, a notice of opportunity for hearing shall be issued as soon as practicable after the application has been docketed; or 
 (12) An amendment to an early site permit issued under subpart A of part 52, or under subpart H of part 53 of this chapter, as follows: (i) If the early site permit does not provide authority to conduct the activities allowed under § 50.10(e)(1) or § 53.1130(b)(1) of this chapter, the amendment will involve no significant hazards consideration, and though the NRC will provide notice of opportunity for a hearing under this section, it may make the amendment immediately effective and grant a hearing thereafter; and (ii) If the early site permit provides authority to conduct the activities allowed under § 50.10(e)(1) or § 53.1130(b)(1) of this chapter and the Commission determines under §§ 50.58 and 50.91 or § 53.1515 of this chapter that an emergency situation exists or that exigent circumstances exist and that the amendment involves no significant hazards consideration, it will provide notice of opportunity for a hearing under § 2.106 of this chapter (if a hearing is requested, which will be held after issuance of the amendment). (13) A manufacturing license under subpart F of part 52 or subpart H of part 53 of this chapter. (b) * * * (3) For a notice of intended operation under § 52.103(a) or § 53.1452(a) of this chapter, the following information: (i) The identification of the NRC action as making the finding required under § 52.103(g) or § 53.1452(g) of this chapter; (ii) The manner in which the licensee notifications under § 52.99(c) or § 53.1449(c), of this chapter which are required to be made available by § 52.99(e)(2) or § 53.1449(e)(2), of this chapter may be obtained and examined; 
 (iv) Any conditions, limitations, or restrictions to be placed on the license in connection with the finding under § 52.103(g) or § 53.1452(g) of this chapter, and the expiration date or circumstances (if any) under which the conditions, limitations or restrictions will no longer apply.
 10.In § 2.106, revise paragraphs (a)(2), (a)(3), and (b)(2) introductory text to read as follows: § 2.106
  (a) * * * (2) An amendment of a license for a facility of the type described in § 50.21(b) or § 50.22, or part 53 of this chapter, as applicable, or a testing facility, whether or not a notice of proposed action has been previously published; and (3) The finding under § 52.103(g) or § 53.1452(g) of this chapter. (b) * * * (2) In the case of a finding under § 52.103(g) or § 53.1452(g) of this chapter:
 11.In § 2.109, revise paragraphs (b), (c), and (d) to read as follows: § 2.109
  
 (b) If the licensee of a nuclear power plant licensed under § 50.21(b) or § 50.22 or under subpart H of part 53 of this chapter files a sufficient application for renewal of either an operating license or a combined license at least 5 years before the expiration of the existing license, the existing license will not be deemed to have expired until the application has been finally determined. (c) If the holder of an early site permit licensed under subpart A of part 52 or under subpart H of part 53 of this chapter, as applicable, files a sufficient application for renewal under § 52.29 or § 53.1173 of this chapter, as applicable, at least 12 months before the expiration of the existing early site permit, the existing permit will not be deemed to have expired until the application has been finally determined. (d) If the licensee of a manufacturing license under subpart F of part 52, or under subpart H of part 53 of this chapter files a sufficient application for renewal under § 52.177 or § 53.1295 of this chapter at least 12 months before the expiration of the existing license, the existing license will not be deemed to have expired until the application has been finally determined.
 12.In § 2.110, revise paragraphs (a)(1) and (b) to read as follows: § 2.110
  (a)(1) A submittal for a standard design approval under subpart E of part 52 or under subpart H of part 53 of this chapter shall be subject to §§ 2.101(a) and 2.390 to the same extent as if it were an application for a permit or license. 
 (b) Upon initiation of review by the NRC staff of a submittal for an early review of site suitability issues under appendix Q to part 50 of this chapter, or for a standard design approval under subpart E of part 52 or under subpart H of part 53 of this chapter, the Director, Office of Nuclear Reactor Regulation, shall publish in the 
                            **Federal Register**
                             a notice of receipt of the submittal, inviting comments from interested persons within 60 days of publication or other time as may be specified, for consideration by the NRC staff and ACRS in their review.
 13.In § 2.202, revise paragraph (e) to read as follows: § 2.202
  
 (e)(1) If the order involves the modification of a part 50 or a part 53 license and is a backfit, the requirements of § 50.109 or § 53.1590 of this chapter, as applicable, shall be followed, unless the licensee has consented to the action required. (2) If the order involves the modification of combined license under subpart C of part 52, or subpart H of part 53 of this chapter, the requirements of § 52.98 or § 53.1443 of this chapter, as applicable, shall be followed unless the licensee has consented to the action required. (3) If the order involves a change to an early site permit under subpart A of part 52 or under subpart H of part 53 of this chapter, the requirements of § 52.39 or § 53.1188 of this chapter, as applicable, must be followed, unless the applicant or licensee has consented to the action required. (4) If the order involves a change to a standard design certification rule referenced by that plant's application, the requirements, if any, in the referenced design certification rule with respect to changes must be followed, or, in the absence of these requirements, 
                            ( print page 87019)
                            the requirements of § 52.63 or § 53.1263 of this chapter, as applicable, must be followed, unless the applicant or licensee has consented to follow the action required. (5) If the order involves a change to a standard design approval referenced by that plant's application, the requirements of § 52.145 or § 53.1221 of this chapter, as applicable, must be followed unless the applicant or licensee has consented to follow the action required. (6) If the order involves a modification of a manufacturing license under subpart F of part 52 or under subpart H of part 53 of this chapter, the requirements of § 52.171 or § 53.1288 of this chapter, as applicable, must be followed, unless the applicant or licensee has consented to the action required.
 14.In § 2.309, revise paragraphs (a), (f)(1)(i), (f)(1)(vi) and (vii), (g), (h)(2), (i)(2), (j) to read as follows: § 2.309
  (a) 
                            *General requirements.*
                             Any person whose interest may be affected by a proceeding and who desires to participate as a party must file a written request for hearing and a specification of the contentions which the person seeks to have litigated in the hearing. In a proceeding under § 52.103 or § 53.1452 of this chapter, as applicable, the Commission, acting as the presiding officer, will grant the request if it determines that the requestor has standing under the provisions of paragraph (d) of this section and has proposed at least one admissible contention that meets the requirements of paragraph (f) of this section. For all other proceedings, except as provided in paragraph (e) of this section, the Commission, presiding officer, or the Atomic Safety and Licensing Board designated to rule on the request for hearing and/or petition for leave to intervene, will grant the request/petition if it determines that the requestor/petitioner has standing under the provisions of paragraph (d) of this section and has proposed at least one admissible contention that meets the requirements of paragraph (f) of this section. In ruling on the request for hearing/petition to intervene submitted by petitioners seeking to intervene in the proceeding on the HLW repository, the Commission, the presiding officer, or the Atomic Safety and Licensing Board shall also consider any failure of the petitioner to participate as a potential party in the pre-license application phase under subpart J of this part in addition to the factors in paragraph (d) of this section. If a request for hearing or petition to intervene is filed in response to any notice of hearing or opportunity for hearing, the applicant/licensee shall be deemed to be a party. 
 (f) * * * (1) * * * (i) Provide a specific statement of the issue of law or fact to be raised or controverted, provided further, that the issue of law or fact to be raised in a request for hearing under § 52.103(b) or § 53.1452(b) of this chapter, as applicable, must be directed at demonstrating that one or more of the acceptance criteria in the combined license have not been, or will not be met, and that the specific operational consequences of nonconformance would be contrary to providing reasonable assurance of adequate protection of the public health and safety; 
 (vi) In a proceeding other than one under § 52.103 or § 53.1452 of this chapter provide sufficient information to show that a genuine dispute exists with the applicant/licensee on a material issue of law or fact. This information must include references to specific portions of the application (including the applicant's environmental report and safety report) that the petitioner disputes and the supporting reasons for each dispute, or, if the petitioner believes that the application fails to contain information on a relevant matter as required by law, the identification of each failure and the supporting reasons for the petitioner's belief; and (vii) In a proceeding under § 52.103(b) or § 53.1452(b) of this chapter, as applicable, the information must be sufficient, and include supporting information showing, prima facie, that one or more of the acceptance criteria in the combined license have not been, or will not be met, and that the specific operational consequences of nonconformance would be contrary to providing reasonable assurance of adequate protection of the public health and safety. This information must include the specific portion of the report required by § 52.99(c) or § 53.1449(c) of this chapter, as applicable, which the requestor believes is inaccurate, incorrect, and/or incomplete (
                            *i.e.,*
                             fails to contain the necessary information required by § 52.99(c) or § 53.1449(c) of this chapter, as applicable). If the requestor identifies a specific portion of the report under § 52.99(c) or § 53.1449(c) of this chapter, as applicable, as incomplete and the requestor contends that the incomplete portion prevents the requestor from making the necessary 
                            *prima facie*
                             showing, then the requestor must explain why this deficiency prevents the requestor from making the 
                            *prima facie*
                             showing. 
 (g) 
                            *Selection of hearing procedures.*
                             A request for hearing and/or petition for leave to intervene may, except in a proceeding under § 52.103 or § 53.1452 of this chapter, as applicable, also address the selection of hearing procedures, taking into account the provisions of § 2.310. If a request/petition relies upon § 2.310(d), the request/petition must demonstrate, by reference to the contention and the bases provided and the specific procedures in subpart G of this part, that resolution of the contention necessitates resolution of material issues of fact which may be best determined through the use of the identified procedures. (h) * * * (2) If the proceeding pertains to a production or utilization facility (as defined in § 50.2 or § 53.020 of this chapter) located within the boundaries of the State, local governmental body, or Federally-recognized Indian Tribe seeking to participate as a party, no further demonstration of standing is required. If the production or utilization facility is not located within the boundaries of the State, local governmental body, or Federally-recognized Indian Tribe seeking to participate as a party, the State, local governmental body, or Federally-recognized Indian Tribe also must demonstrate standing. 
 (i) * * * (2) Except in a proceeding under § 52.103 or § 53.1452 of this chapter, as applicable, the participant who filed the hearing request, intervention petition, or motion for leave to file new or amended contentions after the deadline may file a reply to any answer. The reply must be filed within 7 days after service of that answer. 
 (j) 
                            *Decision on request/petition.*
                             (1) In all proceedings other than a proceeding under § 52.103 or § 53.1452 of this chapter, as applicable, the presiding officer shall issue a decision on each request for hearing or petition to intervene within 45 days of the conclusion of the initial pre-hearing conference or, if no pre-hearing conference is conducted, within 45 days after the filing of answers and replies 
                            ( print page 87020)
                            under paragraph (i) of this section. With respect to a request to admit amended or new contentions, the presiding officer shall issue a decision on each such request within 45 days of the conclusion of any pre-hearing conference that may be conducted regarding the proposed amended or new contentions or, if no pre-hearing conference is conducted, within 45 days after the filing of answers and replies, if any. In the event the presiding officer cannot issue a decision within 45 days, the presiding officer shall issue a notice advising the Commission and the parties, and the notice shall include the expected date of when the decision will issue. (2) The Commission, acting as the presiding officer, shall expeditiously grant or deny the request for hearing in a proceeding under § 52.103 or § 53.1452 of this chapter, as applicable. The Commission's decision may not be the subject of any appeal under § 2.311.
 15.Amend § 2.310 by: a.In paragraphs (a) and (h) introductory text, removing the cross-reference “parts 30, 32 through 36, 39, 40, 50, 52, 54, 55, 61, 70 and 72 of this chapter” and adding, in its place, the cross reference “parts 30, 32 through 36, 39, 40, 50, 52, 53, 54, 55, 61, 70 and 72 of this chapter”; and b.Revising paragraphs (i) and (j). The revisions read as follows. § 2.310
  
 (i) In design certification rulemaking proceedings under part 52 or part 53 of this chapter, any informal hearing held under § 52.51 or § 53.1242 of this chapter, as applicable, must be conducted under the procedures of subpart O of this part. (j) Proceedings on a Commission finding under § 52.103(c) and (g) or § 53.1452(c) and (g) of this chapter, as applicable, shall be conducted in accordance with the procedures designated by the Commission in each proceeding.
 16.In § 2.329, revise paragraph (a) to read as follows: § 2.329
  (a) 
                            *Necessity for prehearing conference; timing.*
                             The Commission or the presiding officer may, and in the case of a proceeding on an application for a construction permit or an operating license for a facility of a type described in §§ 50.21(b) or 50.22, or part 53 of this chapter, or a testing facility, must direct the parties or their counsel to appear at a specified time and place for a conference or conferences before trial. A prehearing conference in a proceeding involving a construction permit or operating license for a facility of a type described in §§ 50.21(b) or 50.22 or part 53 of this chapter must be held within sixty (60) days after discovery has been completed or any other time specified by the Commission or the presiding officer.
 17.In § 2.339, revise paragraph (d) to read as follows: § 2.339
  
 (d) The provisions of this section do not apply to an initial decision directing the issuance of a limited work authorization under 10 CFR 50.10 or 10 CFR 53.1130; an early site permit under subpart A of part 52 or under subpart H of part 53 of this chapter; a construction permit or construction authorization under part 50 or part 53 of this chapter; a combined license under subpart C of part 52 or under subpart H of part 53 of this chapter; or a manufacturing license under subpart F of part 52 or under subpart H of part 53.
 18.In § 2.340, revise paragraphs (b), (c), (d), (f), (i), and (j) to read as follows: § 2.340
  
 (b) 
                            *Initial decision—combined license under 10 CFR parts 52 or 53.*
                             (1) Matters in controversy; presiding officer consideration of matters not put in controversy by parties. In any initial decision in a contested proceeding on an application for a combined license under parts 52 or 53 of this chapter (including an amendment to or renewal of combined license), the presiding officer shall make findings of fact and conclusions of law on the matters put into controversy by the parties and any matter designated by the Commission to be decided by the presiding officer. The presiding officer shall also make findings of fact and conclusions of law on any matter not put into controversy by the parties, but only to the extent that the presiding officer determines that a serious safety, environmental, or common defense and security matter exists, and the Commission approves of an examination of and decision on the matter upon its referral by the presiding officer under, inter alia, the provisions of §§ 2.323 and 2.341. (2) Presiding officer initial decision and issuance of permit or license. (i) In a contested proceeding for the initial issuance or renewal of a combined license under parts 52 or 53 of this chapter, or the amendment of a combined license where the NRC has not made a determination of no significant hazards consideration, the Commission or the Director, Office of Nuclear Reactor Regulation, as appropriate after making the requisite findings, shall issue, deny, or appropriately condition the permit or license in accordance with the presiding officer's initial decision once that decision becomes effective. (ii) In a contested proceeding for the amendment of a combined license under parts 52 or 53 of this chapter where the NRC has made a determination of no significant hazards consideration, the Commission or the Director, Office of Nuclear Reactor Regulation, as appropriate (appropriate official), after making the requisite findings and complying with any applicable provisions of § 2.1202(a) or § 2.1403(a), may issue the amendment before the presiding officer's initial decision becomes effective. Once the presiding officer's initial decision becomes effective, the appropriate official shall take action with respect to that amendment in accordance with the initial decision. If the presiding officer's initial decision becomes effective before the appropriate official issues the amendment, then the appropriate official, after making the requisite findings, shall issue, deny, or appropriately condition the amendment in accordance with the presiding officer's initial decision. (c) 
                            *Initial decision on findings under 10 CFR 52.103 or 10 CFR 53.1452 with respect to acceptance criteria in nuclear power reactor combined licenses.*
                             In any initial decision under § 52.103(g) or § 53.1452(g) of this chapter with respect to whether acceptance criteria have been or will be met, the presiding officer shall make findings of fact and conclusions of law on the matters put into controversy by the parties, and any matter designated by the Commission to be decided by the presiding officer. Matters not put into controversy by the parties but identified by the presiding officer as matters requiring further examination, shall be referred to the Commission for its determination; the Commission may, in its discretion, treat any of these referred matters as a request for action under § 2.206 and process the matter in accordance with § 52.103(f) or § 53.1452(f) of this chapter. (d) 
                            *Initial decision—manufacturing license under 10 CFR parts 52 or 53.*
                             (1) Matters in controversy; presiding officer consideration of matters not put in controversy by parties. In any initial decision in a contested proceeding on an application for a manufacturing 
                            ( print page 87021)
                            license under subpart C of part 52 or subpart H of part 53 of this chapter (including an amendment to or renewal of a manufacturing license), the presiding officer shall make findings of fact and conclusions of law on the matters put into controversy by the parties and any matter designated by the Commission to be decided by the presiding officer. The presiding officer also shall make findings of fact and conclusions of law on any matter not put into controversy by the parties, but only to the extent that the presiding officer determines that a serious safety, environmental, or common defense and security matter exists, and the Commission approves of an examination of and decision on the matter upon its referral by the presiding officer under, inter alia, the provisions of §§ 2.323 and 2.341. (2) Presiding officer initial decision and issuance of permit or license. (i) In a contested proceeding for the initial issuance or renewal of a manufacturing license under subpart C of part 52 or subpart H of part 53 of this chapter, or the amendment of a manufacturing license, the Commission or the Director, Office of Nuclear Reactor Regulation, as appropriate, after making the requisite findings, shall issue, deny, or appropriately condition the permit or license in accordance with the presiding officer's initial decision once that decision becomes effective. (ii) In a contested proceeding for the initial issuance or renewal of a manufacturing license under subpart C of part 52 or subpart H of part 53 of this chapter, or the amendment of a manufacturing license, the Commission or the Director, Office of Nuclear Reactor Regulation, as appropriate (appropriate official), may issue the license, permit, or license amendment in accordance with § 2.1202(a) or § 2.1403(a) before the presiding officer's initial decision becomes effective. If, however, the presiding officer's initial decision becomes effective before the license, permit, or license amendment is issued under § 2.1202 or § 2.1403, then the Commission or the Director, Office of Nuclear Reactor Regulation, as appropriate, shall issue, deny, or appropriately condition the license, permit, or license amendment in accordance with the presiding officer's initial decision. 
 (f) 
                            *Immediate effectiveness of certain presiding officer decisions.*
                             A presiding officer's initial decision directing the issuance or amendment of a limited work authorization under § 50.10 or § 53.1130 of this chapter; an early site permit under subpart A of part 52 or under subpart H of part 53 of this chapter; a construction permit or construction authorization under part 50 or part 53 of this chapter; an operating license under part 50 or part 53 of this chapter; a combined license under subpart C of part 52 or subpart H or part 53 of this chapter; a manufacturing license under subpart F of part 52 or subpart H of part 53 of this chapter; a renewed license under part 53 or part 54 of this chapter; or a license under part 72 of this chapter to store spent fuel in an independent spent fuel storage facility (ISFSI) or a monitored retrievable storage installation (MRS); an initial decision directing issuance of a license under part 61 of this chapter; or an initial decision under § 52.103(g) or § 53.1452(g) of this chapter that acceptance criteria in a combined license have been met, is immediately effective upon issuance unless the presiding officer finds that good cause has been shown by a party why the initial decision should not become immediately effective. 
 (i) 
                            *Issuance of authorizations, permits, and licenses—production and utilization facilities.*
                             The Commission or the Director, Office of Nuclear Reactor Regulation, as appropriate, shall issue a limited work authorization under § 50.10 or § 53.1130 of this chapter; an early site permit under subpart A of part 52 or subpart H of part 53 of this chapter; a construction permit or construction authorization under part 50 or part 53 of this chapter; an operating license under part 50 or part 53 of this chapter; a combined license under subpart C of part 52 or part 53 of this chapter; or a manufacturing license under subpart F of part 52 or part 53 of this chapter within 10 days from the date of issuance of the initial decision: (1) If the Commission or the Director has made all findings necessary for issuance of the authorization, permit or license, not within the scope of the initial decision of the presiding officer; and (2) Notwithstanding the pendency of a petition for reconsideration under § 2.345, a petition for review under § 2.341, or a motion for stay under § 2.342, or the filing of a petition under § 2.206. (j) 
                            *Issuance of finding on acceptance criteria under 10 CFR 52.103 or 10 CFR 53.1452.*
                             The Commission or the Director, Office of Nuclear Reactor Regulation, as appropriate, shall make the finding under § 52.103(g) or § 53.1452(g) of this chapter, that acceptance criteria in a combined license are met within 10 days from the date of the presiding officer's initial decision: (1) If the Commission or the Director is otherwise able to make the finding under § 52.103(g) or § 53.1452(g) of this chapter, that the prescribed acceptance criteria are met for those acceptance criteria not within the scope of the initial decision of the presiding officer; (2) If the presiding officer's initial decision—with respect to contentions that the prescribed acceptance criteria have not been met—finds that those acceptance criteria have been met, and the Commission or the Director thereafter is able to make the finding that those acceptance criteria are met; (3) If the presiding officer's initial decision—with respect to contentions that the prescribed acceptance criteria will not be met—finds that those acceptance criteria will be met, and the Commission or the Director thereafter is able to make the finding that those acceptance criteria are met; and (4) Notwithstanding the pendency of a petition for reconsideration under § 2.345, a petition for review under § 2.341, or a motion for stay under § 2.342, or the filing of a petition under § 2.206.
 § 2.341
 19.In § 2.341(a)(1), remove the phrase “§ 52.103(c)” and add in its place the phrase “§ 52.103(c) or § 53.1452(c)”. § 2.400
 20.In § 2.400, remove the phrase “parts 50 or 52” and add in its place the phrase “part 50 or part 52, or § 53.1470”. 21.In § 2.401, revise the section heading and paragraph (a) to read as follows: § 2.401
 10 CFR parts 50 52 53
 (a) In the case of applications under appendix N of part 50 or § 53.1470 of this chapter for construction permits for nuclear power reactors of the type described in § 50.22 or part 53 of this chapter, or applications under appendix N of part 52 or § 53.1470 of this chapter for combined licenses, the Secretary will issue notices of hearing pursuant to § 2.104.
 22.In § 2.402, revise paragraph (a) to read as follows: § 2.402
  (a) In the case of applications under appendix N of part 50 or § 53.1470 of 
                            ( print page 87022)
                            this chapter for construction permits for nuclear power reactors of a type described in 10 CFR 50.22 or part 53, or applications pursuant to appendix N of part 52 or § 53.1470 of this chapter for combined licenses, the Commission or the presiding officer may order separate hearings on particular phases of the proceeding, such as matters related to the acceptability of the design of the reactor in the context of the site parameters postulated for the design or environmental matters.
 § 2.403
 23.In § 2.403, remove the phrase “appendix N of part 50” and add in its place the phrase “appendix N to part 50 or § 53.1470”. § 2.404
 24.In § 2.404, remove the phrase “appendix N of part 50” and add in its place the phrase “appendix N to part 50 or § 53.1470”. § 2.405
 25.In § 2.405, remove the phrase “part 52” and add in its place the phrase “part 52 or part 53”. § 2.406
 26.In § 2.406, remove the phrase “appendix N of parts 50 or 52” and add in its place the phrase “appendix N to part 50 or part 52 or § 53.1470”. § 2.500
 27.In § 2.500, remove the phrase “subpart F of part 52” and add in its place the phrase “subpart F of part 52 or subpart H of part 53”. 28.In § 2.501, revise the section heading and paragraph (a) introductory text to read as follows: § 2.501
 10 CFR parts 52 53
 (a) In the case of an application under subpart F of part 52 or subpart H of part 53 of this chapter for a license to manufacture nuclear power reactors of the type described in § 50.22 or part 53 of this chapter to be operated at sites not identified in the license application, the Secretary will issue a notice of hearing to be published in the 
                            **Federal Register**
                             at least 30 days before the date set for hearing in the notice.1
                             The notice shall be issued as soon as practicable after the application has been docketed. The notice will state: 
 1
                                 The thirty-day (30) requirement of this paragraph is not applicable to a notice of the time and place of hearing published by the presiding officer after notice of hearing described in this section has been published.
 29.In § 2.643, revise paragraph (b) to read as follows: § 2.643
  
 (b) The Director will accept for docketing part one of an application for a construction permit for a utilization facility which is subject to § 51.20(b) of this chapter and is of the type specified in § 50.21(b)(2) or (3) or § 50.22 or part 53 of this chapter or an application for a combined license where part one of the application as described in § 2.101(a)(9) is complete. Part one will not be considered complete unless it contains the information required by § 50.10(d)(3) or § 53.1130(a)(3) of this chapter. Upon assignment of a docket number, the procedures in § 2.101(a)(3) and (4) relating to formal docketing and the submission and distribution of additional copies of the application must be followed.
 § 2.645
 30.In § 2.645, in paragraph (a), remove the phrase “§ 50.33(a) through (f) of this chapter” and add in its place the phrase “§§ 50.33(a) through (f), 53.1109, and 53.1306(a) or 53.1413 of this chapter, as applicable,”. § 2.649
 31.In § 2.649, remove the phrase “ 10 CFR 50.10(d)” and add in its place the phrase “10 CFR 50.10(d) or 10 CFR 53.1130(a)”. § 2.800
 32.In § 2.800, amend paragraphs (c) and (d) by removing the phrase “subpart B of part 52” and adding in its place the phrase “subpart B of part 52 or subpart H of part 53”. § 2.801
 33.In § 2.801, remove the phrase “subpart B of part 52” and add in its place the phrase “subpart B of part 52 or subpart H of part 53”. § 2.813
 34.In § 2.813(a), remove the phrase “parts 50, 52, and 100” and add in its place the phrase “parts 50, 52, 53, and 100”. § 2.1103
 35.In § 2.1103, remove the phrase “part 50 of this chapter” and add in its place the phrase “parts 50 or 53 of this chapter”. 36.In § 2.1202, revise paragraphs (a)(1) through (3) and (a)(6) to read as follows: § 2.1202
  (a) * * * (1) An application to construct and/or operate a production or utilization facility (including an application for a limited work authorization under §§ 50.12 or 53.1130 of this chapter, or an application for a combined license under subpart C of 10 CFR part 52, or under subpart H of 10 CFR part 53; (2) An application for an early site permit under subpart A of 10 CFR part 52 or under subpart H of 10 CFR part 53; (3) An application for a manufacturing license under subpart F of 10 CFR part 52 or under subpart H of 10 CFR part 53; 
 (6) Production or utilization facility licensing actions that involve significant hazards considerations as defined in §§ 50.92 or 53.1520 of this chapter.
 § 2.1301
 37.In § 2.1301(b), remove “part 50 and part 52” and add in its place “parts 50, 52, and 53”. § 2.1403
 38.In § 2.1403, remove the phrase “ 10 CFR 50.92” and add in its place the phrase “10 CFR 50.92 or 10 CFR 53.1520”. § 2.1500
 39.In § 2.1500, remove the phrase “subpart B of part 52” and add in its place the phrase “subpart B of part 52 or under subpart H of part 53”. § 2.1502
 40.In § 2.1502, in paragraph (a), remove the phrase “§ 52.51(b)” and add in its place the phrase “§§ 52.51(b) or 53.1242(b)(2)”; and in paragraph (b)(1), wherever it appears, remove the phrase “§ 52.51(a)” and add in its place the phrase “§§ 52.51(a) or 53.1242(b)”. PART 10—CRITERIA AND PROCEDURES FOR DETERMINING ELIGIBILITY FOR ACCESS TO RESTRICTED DATA OR NATIONAL SECURITY INFORMATION OR AN EMPLOYMENT CLEARANCE
 41.The authority citation for part 10 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 145, 161 (42 U.S.C. 2165, 2201); Energy Reorganization Act of 1974, sec. 201 (42 U.S.C. 5841); E.O. 10450, 18 FR 2489, 3 CFR, 1949-1953 Comp., p. 936, as amended; E.O. 10865, 25 FR 1583, 3 CFR, 1959-1963 Comp., p. 398, as amended; E.O. 12968, 60 FR 40245, 3 CFR, 1995 Comp., p. 391. § 10.1
 42.In § 10.1, in paragraph (a)(3) remove the phrase “under part 52” and ( print page 87023)
                        add in its place the phrase “under parts 52 or 53”. § 10.2
 43.In § 10.2, in paragraph (b), wherever it appears, remove the phrase “under part 52” and add in its place the phrase “under parts 52 or 53”. PART 11—CRITERIA AND PROCEDURES FOR DETERMINING ELIGIBILITY FOR ACCESS TO OR CONTROL OVER SPECIAL NUCLEAR MATERIAL
 44.The authority citation for part 11 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 161, 223 (42 U.S.C. 2201, 2273); Energy Reorganization Act of 1974, sec. 201 (42 U.S.C. 5841); 44 U.S.C. 3504 note. Section 11.15(e) also issued under 31 U.S.C. 9701; 42 U.S.C. 2214. § 11.7
 45.In § 11.7, in the introductory text, remove the phrase “parts 10, 25, 50, 70, 72, 73, and 95 of this chapter” and add in its place the phrase “parts 10, 25, 50, 53, 70, 72, 73, and 95 of this chapter”. PART 19—NOTICES, INSTRUCTIONS AND REPORTS TO WORKERS: INSPECTION AND INVESTIGATIONS
 46.The authority citation for part 19 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 53, 63, 81, 103, 104, 161, 223, 234, 1701 (42 U.S.C. 2073, 2093, 2111, 2133, 2134, 2201, 2273, 2282, 2297f); Energy Reorganization Act of 1974, secs. 201, 211, 401 (42 U.S.C. 5841, 5851, 5891); 44 U.S.C. 3504 note. 47.In § 19.2, revise paragraph (a) to read as follows: § 19.2
  (a) * * * (1) All persons who receive, possess, use, or transfer material licensed by the NRC under the regulations in parts 30 through 36, 39, 40, 60, 61, 63, 70, or 72 of this chapter, including persons licensed to operate a production or utilization facility under part 50, part 52, or part 53 of this chapter, persons licensed to possess power reactor spent fuel in an independent spent fuel storage installation (ISFSI) under part 72 of this chapter, and in accordance with 10 CFR 76.60 to persons required to obtain a certificate of compliance or an approved compliance plan under part 76 of this chapter; (2) All applicants for and holders of licenses (including construction permits and early site permits) under parts 50, 52, 53, and 54 of this chapter; (3) All applicants for and holders of a standard design approval under subpart E of part 52 or under subpart H of part 53 of this chapter; and (4) All applicants for a standard design certification under subpart B of part 52 or under subpart H of part 53 of this chapter, and those (former) applicants whose designs have been certified under that subpart.
 48.In § 19.3, revise the definitions for “ *License*
                        ” and “
                        *Regulated entities*
                        ” to read as follows: § 19.3
  
 *License*
                             means a license issued under the regulations in parts 30 through 36, 39, 40, 60, 61, 63, 70, or 72 of this chapter, including licenses to manufacture, construct and/or operate a production or utilization facility under parts 50, 52, 53, or 54 of this chapter. 
 *Regulated entities*
                             means any individual, person, organization, or corporation that is subject to the regulatory jurisdiction of the NRC, including (but not limited to) an applicant for or holder of a standard design approval under subpart E of part 52 or under subpart H of part 53 of this chapter or a standard design certification under subpart B of part 52 or under subpart H of part 53 of this chapter.
 § 19.11
 49.In § 19.11, in paragraph (a) introductory text, paragraph (b) introductory text, and paragraph (e)(1), remove the phrase “of part 52” wherever may appears and add in its place the phrase “of part 52 or under subpart H of part 53”. § 19.14
 50.In § 19.14, in paragraph (a), wherever it may appear, remove the phrase “of part 52” and add in its place the phrase “of part 52 or under subpart H of part 53”. § 19.20
 51.In § 19.20, add the number “53,” in sequential order. PART 20—STANDARDS FOR PROTECTION AGAINST RADIATION
 52.The authority citation for part 20 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 11, 53, 63, 65, 81, 103, 104, 161, 170H, 182, 186, 223, 234, 274, 1701 (42 U.S.C. 2014, 2073, 2093, 2095, 2111, 2133, 2134, 2201, 2210h, 2232, 2236, 2273, 2282, 2021, 2297f); Energy Reorganization Act of 1974, secs. 201, 202 (42 U.S.C. 5841, 5842); Low-Level Radioactive Waste Policy Amendments Act of 1985, sec. 2 (42 U.S.C. 2021b); 44 U.S.C. 3504 note. § 20.1002
 53.In § 20.1002, remove the phrase “parts 30 through 36, 39, 40, 50, 52, 60, 61, 63, 70, or 72 of this chapter” and add in its place the phrase “parts 30 through 36, 39, 40, 50, 52, 53, 60, 61, 63, 70, or 72 of this chapter”. 54.In § 20.1003, revise the definition for “ *License*
                        ” to read as follows: § 20.1003
  
 *License*
                             means a license issued under the regulations in parts 30 through 36, 39, 40, 50, 53, 60, 61, 63, 70, or 72 of this chapter.
 § 20.1101
 55.In § 20.1101, in paragraph (d), remove the phrase “subject to § 50.34a” and add in its place the phrase “subject to §§ 50.34a or 53.260 of this chapter”. § 20.1401
 56.Amend § 20.1401 by: a.In paragraph (a), removing the phrase “parts 30, 40, 50, 52, 60, 61, 63, 70, and 72 of this chapter”, and adding in its place the phrase “parts 30, 40, 50, 52, 53, 60, 61, 63, 70, and 72 of this chapter”; and b.In paragraphs (a) and (c) removing the phrase “in accordance with § 50.83” and adding in its place the phrase “in accordance with §§ 50.83 or 53.1080”. 57.In § 20.1403, revise paragraph (d) introductory text to read as follows: § 20.1403
  
 (d) The licensee has submitted a decommissioning plan or License Termination Plan (LTP) to the Commission indicating the licensee's intent to decommission in accordance with §§ 30.36(d), 40.42(d), 50.82 (a) and (b), subpart G of part 53, 70.38(d), or 72.54 of this chapter, and specifying that the licensee intends to decommission by restricting use of the site. The licensee shall document in the LTP or decommissioning plan how the advice of individuals and institutions in the community who may be affected by the decommissioning has been sought and incorporated, as appropriate, following analysis of that advice.
 58.In § 20.1404, revise paragraph (a)(4) introductory text to read as follows: § 20.1404
  (a) * * *
                            ( print page 87024) (4) Has submitted a decommissioning plan or License Termination Plan (LTP) to the Commission indicating the licensee's intent to decommission in accordance with § 30.36(d), 40.42(d), 50.82 (a) and (b), subpart G of part 53, 70.38(d), or 72.54 of this chapter, and specifying that the licensee proposes to decommission by use of alternate criteria. The licensee shall document in the decommissioning plan or LTP how the advice of individuals and institutions in the community who may be affected by the decommissioning has been sought and addressed, as appropriate, following analysis of that advice. In seeking such advice, the licensee shall provide for:
 § 20.1406
 59.In § 20.1406, in paragraphs (a) and (b), wherever it appears, remove the phrase “under part 52” and add in its place the phrase “under parts 52 or 53”. 60.In § 20.1501, revise paragraph (b) to read as follows: § 20.1501
  
 (b) Notwithstanding § 20.2103(a) of this part, records from surveys describing the location and amount of subsurface residual radioactivity identified at the site must be kept with records important for decommissioning, and such records must be retained in accordance with § 30.35(g), § 40.36(f), § 50.75(g), subpart G of part 53, § 70.25(g), or § 72.30(d) of this chapter, as applicable.
 § 20.1905
 61.In § 20.1905, in paragraph (g) introductory text, remove the phrase “Parts 50 or 52” and add in its place the phrase “parts 50, 52, or 53”. 62.In § 20.2004, revise paragraph (b)(1) to read as follows: § 20.2004
  
 (b)(1) Waste oils (petroleum derived or synthetic oils used principally as lubricants, coolants, hydraulic or insulating fluids, or metalworking oils) that have been radioactively contaminated in the course of the operation or maintenance of a nuclear power reactor licensed under parts 50 or 53 of this chapter may be incinerated on the site where generated provided that the total radioactive effluents from the facility, including the effluents from such incineration, conform to the requirements of appendix I to part 50 or § 53.425(d) of this chapter and the effluent release limits contained in applicable license conditions other than effluent limits specifically related to incineration of waste oil. The licensee shall report any changes or additions to the information supplied under §§ 50.34, 50.34a, or under subpart H of part 53 of this chapter associated with this incineration pursuant to §§ 50.71 or 53.1620 of this chapter, as appropriate. The licensee shall also follow the procedures of §§ 50.59 or 53.1565 of this chapter with respect to such changes to the facility or procedures.
 63.In § 20.2201, revise paragraphs (a)(2)(i), (b)(2)(i), and (c) to read as follows: § 20.2201
  (a) * * * (2) * * * (i) Licensees having an installed Emergency Notification System shall make the reports to the NRC Operations Center under §§ 50.72 or 53.1630 of this chapter, and 
 (b) * * * (2) * * * (i) For holders of an operating license for a nuclear power plant, the events included in paragraph (b) of this section must be reported under the procedures described in §§ 50.73(b), (c), (d), (e), and (g) or 53.1640(b), (c), (d), and (e) of this chapter and must include the information required in paragraph (b)(1) of this section, and 
 (c) A duplicate report is not required under paragraph (b) of this section if the licensee is also required to submit a report pursuant to §§ 30.55(c), 37.57, 37.81, 40.64(c), 50.72, 50.73, 53.1630, 53.1640, 70.52, 73.27(b), 73.67(e)(3)(vii), 73.67(g)(3)(iii), 73.1205, or 150.19(c) of this chapter.
 § 20.2202
 64.In § 20.2202, in paragraph (d)(1), remove the phrase “ 10 CFR 50.72” and add in its place the phrase “§§ 50.72 or 53.1630 of this chapter;”. 65.In § 20.2203, revise paragraph (c) to read as follows: § 20.2203
  
 (c) For holders of an operating license or a combined license for a nuclear power plant, the occurrences included in paragraph (a) of this section must be reported under the procedures described in §§ 50.73(b), (c), (d), (e), and (g) or 53.1640(b), (c), (d), and (e) of this chapter, and must include the information required by paragraph (b) of this section. Occurrences reported under §§ 50.73 or 53.1640 of this chapter need not be reported by a duplicate report under paragraph (a) of this section.
 § 20.2206
 66.In § 20.2206, in paragraph (a)(1), remove the phrase “or § 50.22” and add in its place the phrase “, § 50.22, or part 53”. PART 21—REPORTING OF DEFECTS AND NONCOMPLIANCE
 67.The authority citation for part 21 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 53, 63, 81, 103, 104, 161, 223, 234, 1701 (42 U.S.C. 2073, 2093, 2111, 2133, 2134, 2201, 2273, 2282, 2297f); Energy Reorganization Act of 1974, secs. 201, 206 (42 U.S.C. 5841, 5846); Nuclear Waste Policy Act of 1982, secs. 135, 141 (42 U.S.C. 10155, 10161); 44 U.S.C. 3504 note. 68.In § 21.2, revise paragraphs (a)(2) through (4), (b), and (c) to read as follows: § 21.2
  (a) * * * (1) * * * (2) Each individual, corporation, partnership, or other entity doing business within the United States, and each director and responsible officer of such an organization, that constructs a production or utilization facility licensed for manufacture, construction, or operation under parts 50, 52, or 53 of this chapter, an ISFSI for the storage of spent fuel licensed under part 72 of this chapter, an MRS for the storage of spent fuel or high-level radioactive waste under part 72 of this chapter, or a geologic repository for the disposal of high-level radioactive waste under parts 60 or 63 of this chapter; or supplies basic components for a facility or activity licensed, other than for export, under parts 30, 40, 50, 52, 53, 60, 61, 63, 70, 71, or 72 of this chapter; (3) Each individual, corporation, partnership, or other entity doing business within the United States, and each director and responsible officer of such an organization, applying for a design certification rule under parts 52 or 53 of this chapter; or supplying basic components with respect to that design certification, and each individual, corporation, partnership, or other entity doing business within the United States, and each director and responsible officer of such an organization, whose 
                            ( print page 87025)
                            application for design certification has been granted under parts 52 or 53 of this chapter, or who has supplied or is supplying basic components with respect to that design certification; (4) Each individual, corporation, partnership, or other entity doing business within the United States, and each director and responsible officer of such an organization, applying for or holding a standard design approval under parts 52 or 53 of this chapter; or supplying basic components with respect to a standard design approval under parts 52 or 53 of this chapter; (b) For persons licensed to construct a facility under either a construction permit issued under §§ 50.23 or 53.1333 of this chapter or a combined license under parts 52 or 53 of this chapter (for the period of construction until the date that the Commission makes the finding under §§ 52.103(g) or 53.1452(g) of this chapter), or to manufacture a facility under parts 52 or 53 of this chapter, evaluation of potential defects and failures to comply and reporting of defects and failures to comply under §§ 50.55(e) or 53.605 of this chapter satisfies each person's evaluation, notification, and reporting obligation to report defects and failures to comply under this part and the responsibility of individual directors and responsible officers of these licensees to report defects under Section 206 of the Energy Reorganization Act of 1974. (c) For persons licensed to operate a nuclear power plant under part 50, part 52, or part 53 of this chapter, evaluation of potential defects and appropriate reporting of defects under §§ 50.72, 50.73, 53.1630, 53.1640, or 73.1200 and 73.1205 of this chapter, satisfies each person's evaluation, notification, and reporting obligation to report defects under this part, and the responsibility of individual directors and responsible officers of these licensees to report defects under Section 206 of the Energy Reorganization Act of 1974.
 69.In § 21.3, revise the definitions for “Basic component”, “Commercial grade item”, “Critical characteristics”, “Dedicating entity”, “Dedication”, “Defect”, and “Substantial safety hazard” to read as follows: § 21.3
  
 *Basic component.*
                             (1)(i) When applied to nuclear power plants licensed under part 53 of this chapter, basic component means a safety-related structure, system, or component (SSC), or part thereof, and when applied to nuclear power plants licensed under parts 50 or 52, of this chapter, basic component means an SSC, or part thereof that affects its safety function necessary to assure: (A) The integrity of the reactor coolant pressure boundary; (B) The capability to shut down the reactor and maintain it in a safe-shutdown condition; or (C) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to those referred to in §§ 50.34(a)(1), 50.67(b)(2), or 100.11 of this chapter, as applicable. (ii) Basic components are items designed and manufactured under a quality assurance program complying with appendix B to part 50 of this chapter, or commercial grade items which have successfully completed the dedication process. (2) When applied to standard design certifications and approvals under part 53 of this chapter, basic component means the design or procurement information approved or to be approved within the scope of the design certification or approval for a safety-related SSC, or part thereof. When applied to standard design certifications under subpart B of part 52 of this chapter and standard design approvals under part 52 of this chapter, basic component means the design or procurement information approved or to be approved within the scope of the design certification or approval for an SSC, or part thereof, that affects its safety function necessary to assure: (i) The integrity of the reactor coolant pressure boundary; (ii) The capability to shut down the reactor and maintain it in a safe-shutdown condition; or (iii) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to those referred to in §§ 50.34(a)(1), 50.67(b)(2), or 100.11 of this chapter, as applicable. (3) When applied to other facilities and other activities licensed under 10 CFR parts 30, 40, 50 (other than nuclear power plants), 60, 61, 63, 70, 71, or 72 of this chapter, basic component means a structure, system, or component, or part thereof, that affects their safety function, that is directly procured by the licensee of a facility or activity subject to the regulations in this part and in which a defect or failure to comply with any applicable regulation in this chapter, order, or license issued by the Commission could create a substantial safety hazard. (4) In all cases, basic component includes safety-related design, analysis, inspection, testing, fabrication, replacement of parts, or consulting services that are associated with the component hardware, design certification, design approval, or information in support of an early site permit application under part 52 or part 53 of this chapter, whether these services are performed by the component supplier or others. *Commercial grade item.*
                             (1) When applied to nuclear power plants licensed under parts 50 or 53 of this chapter, commercial grade item means an SSC, or part thereof that affects its safety function, that was not designed and manufactured as a basic component. Commercial grade items do not include items where the design and manufacturing process require in-process inspections and verifications to ensure that defects or failures to comply are identified and corrected (
                            *i.e.,*
                             one or more critical characteristics of the item cannot be verified). (2) When applied to facilities and activities licensed pursuant to parts 30, 40, 50 (other than nuclear power plants), 60, 61, 63, 70, 71, or 72 of this chapter, commercial grade item means an item that is: (i) Not subject to design or specification requirements that are unique to those facilities or activities; (ii) Used in applications other than those facilities or activities; and (iii) To be ordered from the manufacturer/supplier on the basis of specifications set forth in the manufacturer's published product description (for example, a catalog). 
 *Critical characteristics.*
                             When applied to nuclear power plants licensed under parts 50, 52, or 53 of this chapter, critical characteristics are those important design, material, and performance characteristics of a commercial grade item that, once verified, will provide reasonable assurance that the item will perform its intended safety function. *Dedicating entity.*
                             When applied to nuclear power plants licensed under parts 50, 52, or 53 of this chapter, dedicating entity means the organization that performs the dedication process. Dedication may be performed by the manufacturer of the item, a third-party dedicating entity, or the licensee itself. The dedicating entity, under § 21.21(c) of this part, is responsible for identifying and evaluating deviations, reporting defects and failures to comply for the dedicated item, and maintaining auditable records of the dedication process. *Dedication.*
                             (1) When applied to nuclear power plants licensed pursuant to 10 CFR parts 30, 40, 50, 53, or 60, dedication is an acceptance process 
                            ( print page 87026)
                            undertaken to provide reasonable assurance that a commercial grade item to be used as a basic component will perform its intended safety function and, in this respect, is deemed equivalent to an item designed and manufactured under a 10 CFR part 50, appendix B, quality assurance program. This assurance is achieved by identifying the critical characteristics of the item and verifying their acceptability by inspections, tests, or analyses performed by the purchaser or third-party dedicating entity after delivery, supplemented as necessary by one or more of the following: commercial grade surveys; product inspections or witness at holdpoints at the manufacturer's facility, and analysis of historical records for acceptable performance. In all cases, the dedication process must be conducted under the applicable provisions of 10 CFR part 50, appendix B. The process is considered complete when the item is designated for use as a basic component. (2) When applied to facilities and activities licensed pursuant to 10 CFR parts 30, 40, 50 (other than nuclear power plants), 60, 61, 63, 70, 71, or 72, dedication occurs after receipt when that item is designated for use as a basic component. *Defect*
                             means: (1) A deviation in a basic component delivered to a purchaser for use in a facility or an activity subject to the regulations in this part if, on the basis of an evaluation, the deviation could create a substantial safety hazard; (2) The installation, use, or operation of a basic component containing a defect as defined in this section; (3) A deviation in a portion of a facility subject to the early site permit, standard design certification, standard design approval, construction permit, combined license or manufacturing licensing requirements of parts 50, 52, or 53 of this chapter, provided the deviation could, on the basis of an evaluation, create a substantial safety hazard and the portion of the facility containing the deviation has been offered to the purchaser for acceptance; (4) A condition or circumstance involving a basic component that could contribute to the exceeding of a safety limit, as defined in the technical specifications of a license for operation issued under part 50, part 52, or part 53 of this chapter; or (5) An error, omission or other circumstance in a design certification, or standard design approval that, on the basis of an evaluation, could create a substantial safety hazard. 
 *Substantial safety hazard*
                             means a loss of safety function to the extent that there is a major reduction in the degree of protection provided to public health and safety for any facility or activity licensed or otherwise approved or regulated by the NRC, other than for export, under part 30, 40, 50, 52, 53, 60, 61, 63, 70, 71, or 72 of this chapter.
 § 21.21
 70.Amend § 21.21 by: a.In paragraph (a)(3), removing the phrase “under part 52” and add in its place the phrase “under parts 52 or 53”; and b.In paragraphs (d)(1)(i) and (ii) removing the phrase “parts 30, 40, 50, 52, 60, 61, 63, 70, 71, or 72 of this chapter” and adding in its place the phrase “parts 30, 40, 50, 52, 53, 60, 61, 63, 70, 71, or 72 of this chapter”. § 21.51
 71.In § 21.51, in paragraphs (a)(4) and (5) remove the phrase “of part 52” and add in its place the phrase “of part 52 or under subpart H of part 53”. § 21.61
 72.In § 21.61, in paragraph (b) remove the phrase “under part 52” and add in its place the phrase “under parts 52 or 53”. PART 25—ACCESS AUTHORIZATION
 73.The authority citation for part 25 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 145, 161, 223, 234 (42 U.S.C. 2165, 2201, 2273, 2282); Energy Reorganization Act of 1974, sec. 201 (42 U.S.C. 5841); 44 U.S.C. 3504 note; E.O. 10865, 25 FR 1583, as amended, 3 CFR, 1959-1963 Comp., p. 398; E.O. 12829, 58 FR 3479, 3 CFR, 1993 Comp., p. 570; E.O. 13526, 75 FR 707, 3 CFR, 2009 Comp., p. 298; E.O. 12968, 60 FR 40245, 3 CFR, 1995 Comp., p. 391. Section 25.17(f) and Appendix A also issued under 31 U.S.C. 9701; 42 U.S.C. 2214. 74.In § 25.5, revise the definition for “ *License*
                        ” to read as follows: § 25.5
  
 *License*
                             means a license issued pursuant to 10 CFR parts 50, 52, 53, 60, 63, 70, or 72.
 § 25.17
 75.In § 25.17, in paragraph (a), remove the phrase “under 10 CFR parts 50, 52, 54, 60, 63, 70, 72, or 76” and add in its place the phrase “under 10 CFR parts 50, 52, 53, 54, 60, 63, 70, 72, or 76”. § 25.35
 76.In § 25.35, in paragraph (a), wherever it appears, remove the phrase “under part 52” and add in its place the phrase “under parts 52 or 53”. PART 26—FITNESS FOR DUTY PROGRAMS
 77.The authority citation for part 26 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 53, 103, 104, 107, 161, 223, 234, 1701 (42 U.S.C. 2073, 2133, 2134, 2137, 2201, 2273, 2282, 2297f); Energy Reorganization Act of 1974, secs. 201, 202 (42 U.S.C. 5841, 5842); 44 U.S.C. 3504 note. 78.In § 26.3, revise paragraph (d) and add paragraph (f) to read as follows: § 26.3
  
 (d) Contractor/vendors (C/Vs) who implement FFD programs or program elements, to the extent that the licensees and other entities specified in paragraphs (a) through (c) and (f) of this section rely on those C/V FFD programs or program elements to meet the requirements of this part, shall comply with the requirements of this part. 
 (f) No later than the start of construction activities, licensees and other entities that have applied for or have been issued a license under part 53 of this chapter, other than a manufacturing license (ML), must implement the requirements in subpart M of this part or all the requirements of this part except subparts K and M. Holders of an ML under part 53 of this chapter must implement the requirements in subpart M or all the requirements of this part except subparts K and M, before commencing activities that assemble a manufactured reactor.
 79.In § 26.4, revise paragraphs (a) introductory text, (a)(1), (a)(4), (b), (c), (e) introductory text, (e)(4), (f), (g) introductory text, and (h) to read as follows: § 26.4
  (a) All persons who are granted unescorted access to nuclear power reactor protected areas by the licensees in § 26.3(a) and, as applicable, (c) and perform the following duties shall be subject to an FFD program that meets all of the requirements of this part, except subpart K of this part, and those persons who are granted unescorted access to either nuclear power reactor protected areas or remote facilities where safety-significant systems or components may be operated within the design basis of 
                            ( print page 87027)
                            a licensed commercial nuclear plant, by the licensees and other entities in § 26.3(f) and perform the following duties must be subject to an FFD program that satisfies the requirements in subpart M of this part, unless the licensee or other entity subjects these individuals to an FFD program that satisfies all of the requirements of this part except for those requirements in subparts K and M: (1) For persons who are granted unescorted access by the licensees in § 26.3(a) and, as applicable, (c), operating or onsite directing of the operation of systems and components that a risk-informed evaluation process has shown to be significant to public health and safety; for those persons who are granted unescorted access by the licensees and other entities in § 26.3(f), operating or directing of the operation of systems and components that a risk-informed evaluation process has shown to be significant to public health and safety; 
 (4) For persons who are granted unescorted access to nuclear power reactor protected areas by the licensees in § 26.3(a) and, as applicable, (c), performing maintenance or onsite directing of the maintenance of SSCs that a risk-informed evaluation process has shown to be significant to public health and safety; for those persons who are granted unescorted access to nuclear power reactor protected areas by the licensees and other entities in § 26.3(f), performing maintenance or directing of the maintenance of SSCs that a risk-informed evaluation process has shown to be significant to public health and safety; and 
 (b) All persons who are granted unescorted access to nuclear power reactor protected areas by the licensees in § 26.3(a) and, as applicable, (c) and who do not perform the duties described in paragraph (a) of this section shall be subject to an FFD program that meets all of the requirements of this part, except §§ 26.205 through 26.209 and subpart K of this part. All persons who are granted unescorted access to a facility licensed under part 53 of this chapter, and who do not perform or direct the performance of the duties described in § 26.4(a), must be subject to the requirements in subpart M of this part, unless the licensee or other entity implements an FFD program that satisfies all of the requirements of this part, except §§ 26.205 through 26.209 and subparts K and M. (c) All persons who are required by a licensee in § 26.3(a) and, as applicable, (c) to physically report to the licensee's Technical Support Center or Emergency Operations Facility by licensee emergency plans and procedures shall be subject to an FFD program that meets all of the requirements of this part, except §§ 26.205 through 26.209 and subpart K of this part. Also, for licensees or other entities in § 26.3(f), all persons without unescorted access to the facility who make decisions and/or direct actions regarding plant safety and security, and all persons who participate remotely in emergency response activities or physically report to the Technical Support Center or Emergency Operations Facility (or an equivalent facility), must be subject to an FFD program that satisfies all of the requirements described in subpart M of this part, unless the licensee or other entity implements an FFD program that satisfies all of the requirements of this part, except §§ 26.205 through 26.209 and subparts K and M. 
 (e) When construction activities, as defined in § 26.5, begin, any individual whose duties for the licensees and other entities in § 26.3(c) require him or her to have the following types of access or perform the following activities at the location where the nuclear power plant will be constructed and operated shall be subject to an FFD program that meets all of the requirements of this part, except subparts I, K, and M of this part, and for any individual whose duties for the licensees and other entities in § 26.3(f) require him or her to have the following types of access, perform construction activities as defined in § 26.5, or perform the following activities must be subject to an FFD program as described in subpart M or an FFD program that satisfies all of the requirements of this part, except subparts I, K, and M: 
 (4) Witnesses or determines inspections, tests, and analyses certification required under part 52 or part 53 of this chapter; 
 (f) Any individual who is constructing or directing the construction of safety- or security-related SSCs shall be subject to an FFD program that meets the requirements of subpart K, or, if applicable, subpart M of this part, unless the licensee or other entity subjects these individuals to an FFD program that meets all of the requirements of this part, except for subparts I, K, and M of this part. (g) All FFD program personnel who are involved in the day-to-day operations of the program, as defined by the procedures of the licensees and other entities in § 26.3(a) through (c), and, as applicable, (d) and whose duties require them to have the following types of access or perform the following activities shall be subject to an FFD program that meets all of the requirements of this part, except subparts I, K, and M of this part, and, at the licensee's or other entity's discretion, subpart C of this part. All personnel whose duties require them to have the following types of access or perform the following activities at facilities licensed under part 53 of this chapter must be subject to the requirements in subpart M or an FFD program that satisfies all of the requirements of this part, except subparts I, K, and M, and, at the licensee's or other entity's discretion, subpart C of this part: 
 (h) Individuals who have applied for authorization to have the types of access or perform the activities described in paragraphs (a) through (d) of this section shall be subject to §§ 26.31(c)(1), 26.35(b), 26.37, 26.39, and the applicable requirements of subparts C, E through H, and M of this part.
 80.Amend § 26.5 by: a.Adding the definitions for “Biological marker” and “Change”; b.Revising the definitions for “Constructing or construction activities” “Contractor/vendor (C/V)”; c.Adding the definition of “Illicit substance”; d.Revising the definitions of “Other entity” and “Questionable validity”; e.Adding the definitions of “Reduction in FFD program effectiveness”; f.Revising the definitions of “Reviewing official”, “Safety-related structures, systems, and components (SSCs)”, and “Security-related SSCs”; g.Adding the definitions of “Special nuclear material”; and h.Revising the definition of “Unit outage”. The additions and revisions read as follows: § 26.5
  
 *Biological marker*
                             means, for a part 53 licensee implementing subpart M of this part, an endogenous substance that is used to validate that the biological specimen collected for testing was produced by the donor. 
 
 
 *Change*
                             as used in § 26.603(e) means an action that results in a modification of, addition to, or removal from the licensee's or other entity's FFD program. 
 *Constructing or construction activities*
                             means, for the purposes of this part, the tasks involved in building a nuclear power plant that are performed at the location where the nuclear power plant will be constructed and operated. These tasks include fabricating, erecting, integrating, and testing safety- and security-related SSCs, and the installation of their foundations, including the placement of concrete. For a licensee or other entity described in § 26.3(f), construction is defined in § 53.020 of this chapter. *Contractor/vendor (C/V)*
                             means any company, or any individual not employed by a licensee or other entity specified in § 26.3(a) through (c) and (f), who is providing work or services to a licensee or other entity covered in § 26.3(a) through (c) and (f), either by contract, purchase order, oral agreement, or other arrangement. 
 *Illicit substance*
                             means a substance that causes impairment and possible addiction but is not an illegal drug as defined in § 26.5. 
 *Other entity*
                             means any corporation, firm, partnership, limited liability company, association, C/V, or other organization who is subject to this part under § 26.3(a) through (c) and (f) but is not licensed by the NRC. 
 *Questionable validity*
                             means the results of validity screening or initial validity tests at a licensee testing facility indicating that a urine specimen may be adulterated, substituted, dilute, or invalid. For a part 53 licensee or other entity, 
                            *questionable validity*
                             means the results of validity screening or initial validity tests indicating that a biological specimen obtained from an individual pursuant to subpart M of this part may be adulterated, substituted, dilute, or invalid. *Reduction in FFD program effectiveness*
                             means, for a part 53 licensee or other entity implementing subpart M of this part, a change or series of changes to an element of the FFD program that reduces or eliminates the licensee's ability to satisfy or maintain site-specific FFD program performance when compared to historical site-specific performance, the licensee's fleet-level program performance, or industry performance. 
 *Reviewing official*
                             means an employee of a licensee or other entity specified in § 26.3(a) through (c) and (f), who is designated by the licensee or other entity to be responsible for reviewing and evaluating any potentially disqualifying FFD information about an individual, including, but not limited to, the results of a determination of fitness, as defined in § 26.189, in order to determine whether the individual may be granted or maintain authorization. *Safety-related structures, systems, and components (SSCs)*
                             means, for part 50 or part 52 licensees and other entities described in § 26.3(a) through (d), those SSCs that are relied on to remain functional during and following design basis events to ensure the integrity of the reactor coolant pressure boundary, the capability to shut down the reactor and maintain it in a safe shutdown condition, or the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposure comparable to the guidelines in § 50.34(a)(1) of this chapter. For part 53 licensees and other entities described in § 26.3(d) and (f), safety-related has the same meaning as that in § 53.020 of this chapter. *Security-related SSCs*
                             means, for the purposes of this part, those structures, systems, and components that the licensee will rely on to implement the licensee's physical security and safeguards contingency plans that either are required under part 73 of this chapter if the licensee is a construction permit applicant or holder or an early site permit holder, as described in § 26.3(c)(3) through (c)(5), respectively, or are included in the licensee's application if the licensee is a combined license applicant or holder, as described in § 26.3(c)(1) and (c)(2), respectively, or a licensee or other entity described in § 26.3(d) or (f). 
 *Special nuclear material (SNM)*
                             has the same meaning as that in § 70.4 of this chapter. 
 *Unit outage*
                             means, for the purposes of this part, for electricity-generation units, that the reactor unit is disconnected from the electrical grid. 
                            *Unit outage*
                             means, for the purposes of this part, for non-electricity-generation units, that the reactor unit is disconnected from the loads to which its output is supplied under normal operating conditions.
 81.In § 26.8, revise paragraph (b) to read as follows: § 26.8
  
 (b) The approved information collection requirements contained in this part appear in §§ 26.9, 26.27, 26.29, 26.31, 26.33, 26.35, 26.37, 26.39, 26.41, 26.53, 26.55, 26.57, 26.59, 26.61, 26.63, 26.65, 26.67, 26.69, 26.75, 26.77, 26.85, 26.87, 26.89, 26.91, 26.93, 26.95, 26.97, 26.99, 26.101, 26.103, 26.107, 26.109, 26.111, 26.113, 26.115, 26.117, 26.119, 26.125, 26.127, 26.129, 26.135, 26.137, 26.139, 26.153, 26.157, 26.159, 26.163, 26.165, 26.167, 26.168, 26.169, 26.183, 26.185, 26.187, 26.189, 26.202, 26.203, 26.205, 26.207, 26.211, 26.401, 26.403, 26.405, 26.406, 26.407, 26.411, 26.413, 26.415, 26.417, 26.603, 26.604, 26.605, 26.606, 26.607, 26.608, 26.609, 26.611, 26.613, 26.617, 26.619, 26.711, 26.713, 26.715, 26.717, 26.719, and 26.821.
 82.Revise § 26.21 to read as follows: § 26.21
  The licensees and other entities specified in § 26.3(a) through (c) and (f) (for those licensees and other entities that do not implement the requirements in subparts M and K of this part) shall establish, implement, and maintain FFD programs that, at a minimum, comprise the program elements contained in this subpart. The individuals specified in § 26.4(a) through (e) and (g), and, at the licensee's or other entity's discretion, § 26.4(f), and, if necessary, § 26.4(j) shall be subject to these FFD programs. Licensees and other entities may rely on the FFD program or program elements of a C/V, as defined in § 26.5, if the C/V's FFD program or program elements satisfy the applicable requirements of this part.
 83.Revise § 26.51 to read as follows: § 26.51
  The requirements in this subpart apply to the licensees and other entities identified in § 26.3(a), (b), and, as applicable, (c) for the categories of individuals in § 26.4(a) through (d), and, at the licensee's or other entity's discretion, in § 26.4(g) and, if necessary, § 26.4(j). The requirements in this subpart also apply to the licensees and other entities specified in § 26.3(c), as applicable, for the categories of individuals in § 26.4(e). At the discretion of a licensee or other entity in § 26.3(c), the requirements of this subpart also may be applied to the categories of individuals identified in § 26.4(f). In addition, the requirements in this subpart apply to the entities in § 26.3(d) to the extent that a licensee or other entity relies on the C/V to satisfy the requirements of this subpart. Certain requirements in this subpart also apply 
                            ( print page 87029)
                            to the individuals specified in § 26.4(h). The requirements in this subpart apply to the FFD programs of licensees and other entities identified in § 26.3(f) that elect not to implement the requirements in subpart M for the categories of individuals in § 26.4 and those licensees and other entities that elect to implement the requirements in § 26.605.
 § 26.53
 84.Amend § 26.53 by: a.In paragraph (e), wherever it appears, remove the phrase “§ 26.3(a) through (c)” and add in its place the phrase “§ 26.3(a) through (c) and (f)”; and b.In paragraphs (g), (h), and (i), wherever it appears, remove the phrase “(c) and (d)” and add in its place the phrase “(c), (d), and (f)”. § 26.63
 85.In § 26.63, in paragraph (d) remove the phrase “§ 26.3(a) through (d)” and add in its place the phrase “§ 26.3(a) through (d) and (f)”. 86.Revise § 26.73 to read as follows: § 26.73
  The requirements in this subpart apply to the licensees and other entities identified in § 26.3(a), (b), and, as applicable, (c) for the categories of individuals specified in § 26.4(a) through (d) and (g). The requirements in this subpart also apply to the licensees and other entities specified in § 26.3(c), as applicable, for the categories of individuals in § 26.4(e). At the discretion of a licensee or other entity in § 26.3(c), the requirements of this subpart also may be applied to the categories of individuals identified in § 26.4(f). In addition, the requirements in this subpart apply to the entities in § 26.3(d) to the extent that a licensee or other entity relies on the C/V to satisfy the requirements of this subpart. The regulations in this subpart also apply to the individuals specified in § 26.4(h) and (j), as appropriate. The requirements in this subpart apply to the FFD programs of licensees and other entities identified in § 26.3(f) that elect not to implement the requirements in subpart M for the categories of individuals in § 26.4 and those licensees and other entities that elect to implement the requirements in § 26.605(b).
 87.Revise § 26.81 to read as follows: § 26.81
  This subpart contains requirements for collecting specimens for drug testing and conducting alcohol tests by or on behalf of the licensees and other entities in § 26.3(a) through (d) for the categories of individuals specified in § 26.4(a) through (d) and (g). At the discretion of a licensee or other entity in § 26.3(c), specimen collections and alcohol tests must be conducted either under this subpart for the individuals specified in § 26.4(e) and (f) or the licensee or other entity may rely on specimen collections and alcohol tests conducted under the requirements of 49 CFR part 40 for the individuals specified in § 26.4(e) and (f). The requirements of this subpart do not apply to specimen collections and alcohol tests that are conducted under the requirements of 49 CFR part 40, as permitted in this paragraph and under §§ 26.4(j) and 26.31(b)(2) and subpart K. The requirements in this subpart apply to the FFD programs of licensees and other entities identified in § 26.3(f) that elect not to implement the requirements in subpart M for the categories of individuals in § 26.4 and those licensees and other entities that elect to implement the requirements in § 26.605.
 88.Revise § 26.201 to read as follows: § 26.201
  (a) The requirements in this subpart, with the exception of § 26.202, apply to the licensees and other entities identified in § 26.3(a); if applicable, (c), (d), and (f), for licensees and other entities not implementing the requirements in subparts K and M. For the licensees and other entities to whom the requirements in this subpart, with the exception of § 26.202, apply, the requirements in §§ 26.203 and 26.211 apply to the individuals identified in § 26.4(a) through (c). In addition, the requirements in §§ 26.205 through 26.209 apply to the individuals identified in § 26.4(a). (b) The requirements in this subpart, with the exception of § 26.203, apply to the licensees or other entities identified in § 26.3(f) implementing this subpart under §§ 26.604 and 26.605. For these licensees and other entities, the requirements in §§ 26.202 and 26.211 apply to the individuals identified in § 26.4(a) through (c) and any person licensed to operate under 10 CFR part 53; and the requirements in §§ 26.205 through 26.209 apply to the individuals identified in § 26.4(a).
 89.Add § 26.202 to read as follows: § 26.202
  (a) 
                            *Policy.*
                             Licensees must establish a policy for the management of fatigue for all individuals who are subject to the licensee's FFD program and incorporate it into the written policy required in § 26.606(a). (b) 
                            *Procedures.*
                             In addition to the procedures required in § 26.606(b), licensees must develop, implement, and maintain procedures that— (1) Describe the process to be followed when any individual identified in § 26.4(a) through (c) makes a self-declaration that he or she is not fit to safely and competently perform his or her duties for any part of a working tour as a result of fatigue. The procedure must— (i) Describe the individual's and licensee's rights and responsibilities related to self-declaration; (ii) Describe requirements for establishing controls and conditions under which an individual may be permitted or required to perform work after that individual declares that he or she is not fit due to fatigue; and (iii) Describe the process to be followed if the individual disagrees with the results of a fatigue assessment that is required under § 26.211(a)(2); (2) Describe the process for implementing the controls required under § 26.205 for the individuals who are performing the duties listed in § 26.4(a); (3) Describe the process to be followed in conducting fatigue assessments under § 26.211; and (4) Describe the disciplinary actions that the licensee may impose on an individual following a fatigue assessment, and the conditions and considerations for taking those disciplinary actions. (c) 
                            *Training and assessments.*
                             Licensees must include the following KAs in the content of the training and trainee assessments required in § 26.608: (1) Knowledge of the contributors to worker fatigue, circadian variations in alertness and performance, indications and risk factors for common sleep disorders, shiftwork strategies for obtaining adequate rest, and the effective use of fatigue countermeasures; and (2) Ability to identify symptoms of worker fatigue and contributors to decreased alertness in the workplace. (d) 
                            *Recordkeeping.*
                             Licensees must retain the following records for at least 3 years or until the completion of all related legal proceedings, whichever is later: (1) Records of work hours for individuals who are subject to the work hour controls in § 26.205; (2) For licensees implementing the requirements of § 26.205(d)(3), records of shift schedules and shift cycles, or, for licensees implementing the requirements of § 26.205(d)(7), records of shift schedules and records showing the beginning and end times and dates of all averaging periods, of individuals 
                            ( print page 87030)
                            who are subject to the work hour controls in § 26.205; (3) The documentation of waivers that is required in § 26.207(a)(4), including the bases for granting the waivers; (4) The documentation of work hour reviews that is required in § 26.205(e)(3) and (e)(4); and (5) The documentation of fatigue assessments that is required in § 26.211(g). (e) 
                            *Reporting.*
                             Licensees must include the following information in a standard format in the annual FFD program performance report required under § 26.617(b)(2): (1) A summary for each nuclear power plant site of all instances during the previous calendar year when the licensee waived one or more of the work hour controls specified in § 26.205(d)(1) through (d)(5)(i) and (d)(7) for individuals described in § 26.4(a). The summary must include only those waivers under which work was performed. If it was necessary to waive more than one work hour control during any single extended work period, the summary of instances must include each of the work hour controls that were waived during the period. For each category of individuals specified in § 26.4(a), the licensee must report— (i) The number of instances when each applicable work hour control specified in § 26.205(d)(1)(i) through (iii), (d)(2)(i) and (ii), (d)(3)(i) through (v), and (d)(7) was waived for individuals not working on outage activities; (ii) The number of instances when each applicable work hour control specified in § 26.205(d)(1)(i) through (iii), (d)(2)(i) and (ii), (d)(3)(i) through (v), (d)(4) and (d)(5)(i), and (d)(7) was waived for individuals working on outage activities; and (iii) A summary that shows the distribution of waiver use among the individuals applicable within each category of individuals identified in § 26.4(a) (
                            *e.g.,*
                             a table that shows the number of individuals who received only one waiver during the reporting period, the number of individuals who received a total of two waivers during the reporting period). (2) A summary of corrective actions, if any, resulting from the analyses of these data, including fatigue assessments. (f) 
                            *Audits.*
                             Licensees must audit the management of worker fatigue under § 26.615.
 90.In § 26.205, revise paragraphs (d)(7)(iii) and (d)(8) to read as follows: § 26.205
  
 (d) * * * (7) * * * (iii) Each licensee shall state, in its FFD policy and procedures required by either §§ 26.27 and 26.203(a) and (b) or §§ 26.202(a) and (b) and 26.606, the work hour counting system in § 26.205(d)(7)(ii) the licensee is using. (8) Each licensee shall state, in its FFD policy and procedures required by either §§ 26.27 and 26.203(a) and (b) or §§ 26.202(a) and (b) and 26.606, the requirements with which the licensee is complying: the minimum days off requirements in § 26.205(d)(3) or maximum average work hours requirements in § 26.205(d)(7).
 91.In § 26.207, revise paragraph (a)(1)(ii) to read as follows: § 26.207
  (a) * * * (1) * * * (ii) A supervisor assesses the individual face to face and determines that there is reasonable assurance that the individual will be able to safely and competently perform his or her duties during the additional work period for which the waiver will be granted. The supervisor performing the assessment shall be trained as required by either §§ 26.29 and 26.203(c) or §§ 26.202(c) and 26.608 and shall be qualified to direct the work to be performed by the individual. If there is no supervisor on site who is qualified to direct the work, the assessment may be performed by a supervisor who is qualified to provide oversight of the work to be performed by the individual. At a minimum, the assessment must address the potential for acute and cumulative fatigue considering the individual's work history for at least the past 14 days, the potential for circadian degradations in alertness and performance considering the time of day for which the waiver will be granted, the potential for fatigue-related degradations in alertness and performance to affect risk-significant functions, and whether any controls and conditions must be established under which the individual will be permitted to perform work. For licensees and other entities in § 26.3(f), the assessment may be performed remotely using electronic communications. In such instances, the assessment must be supported by someone who is present in-person with the individual whose alertness may be impaired, and that supporting person must be trained under the requirements of either § 26.29 and § 26.203(c) or § 26.202(c) and § 26.608.
 92.In § 26.211, revise paragraphs (a)(1) and (3) and paragraph (b) introductory text to read as follows: § 26.211
  (a) * * * (1) For-cause. In addition to any other test or determination of fitness that may be required under §§ 26.31(c), 26.77, 26.607(b), and 26.619, a fatigue assessment must be conducted in response to an observed condition of impaired individual alertness creating a reasonable suspicion that an individual is not fit to safely and competently perform his or her duties, except if the condition is observed during an individual's break period. If the observed condition is impaired alertness with no other behaviors or physical conditions creating a reasonable suspicion of possible substance abuse, then the licensee need only conduct a fatigue assessment. If the licensee has reason to believe that the observed condition is not due to fatigue, the licensee need not conduct a fatigue assessment; 
 (3) Post-event. A fatigue assessment must be conducted in response to events requiring post-event drug and alcohol testing as specified in § 26.31(c) or post-event tests in § 26.607(b)(4). Licensees may not delay necessary medical treatment in order to conduct a fatigue assessment; and 
 (b) Only supervisors and FFD program personnel who are trained under either §§ 26.29 and 26.203(c) or §§ 26.202(c) and 26.608 may conduct a fatigue assessment. The fatigue assessment must be conducted face to face with the individual whose alertness may be impaired. For licensees and other entities in § 26.3(f), a fatigue assessment may be performed remotely using electronic communications. In such instances, the fatigue assessment must be supported by someone who is present in-person with the individual whose alertness may be impaired, and that supporting person must be trained in accordance with the requirements of either §§ 26.29 and 26.203(c) or §§ 26.202(c) and 26.608.
 93.Add Subpart M, consisting of §§ 26.601 through 26.619, to read as follows: Subpart M—Fitness for Duty Programs for Facilities Licensed Under 10 CFR Part 53
 26.601
                        Applicability.
                        26.603
                        General provisions.
                            ( print page 87031)
                        26.604
                        FFD program requirements for facilities that satisfy the § 26.603(c) criterion.
                        26.605
                        FFD program requirements for facilities that do not implement § 26.604.
                        26.606
                        Written policy and procedures.
                        26.607
                        Drug and alcohol testing.
                        26.608
                        FFD program training.
                        26.609
                        Behavioral observation.
                        26.610
                        Sanctions.
                        26.611
                        Protection of information.
                        26.613
                        Appeals process.
                        26.615
                        Audits.
                        26.617
                        Recordkeeping and reporting.
                        26.619
                        Suitability and fitness determinations.
 § 26.601
  A licensee or other entity in § 26.3(f), at its discretion, may establish, implement, and maintain a fitness-for-duty (FFD) program that satisfies the requirements of this subpart for those categories of individuals in § 26.4, as applicable, and any person licensed to operate under 10 CFR part 53. If a licensee or other entity in § 26.3(f) does not elect to implement an FFD program that satisfies the requirements of this subpart, then those categories of individuals in § 26.4, as applicable, and any person licensed to operate under 10 CFR part 53 must be subject to an FFD program that satisfies all part 26 requirements, except for those requirements in subparts K and M.
 § 26.603
  (a) 
                            *FFD program description.*
                             An applicant's description of the FFD program in its Final Safety Analysis Report, required by subpart H of part 53 of this chapter, must include— (1) If the applicant performed the analysis under paragraph (c) of this section, a summary of the analysis, including the assumptions, methodology, conclusion, and references; (2) A statement whether the FFD program will be implemented pursuant to § 26.604 or § 26.605, or will satisfy all part 26 requirements, except for the requirements in subparts K and M; (3) A discussion of the applicability of the FFD program to those individuals described in § 26.4 and how the program will be implemented offsite at a U.S. Nuclear Regulatory Commission (NRC)-licensed facility authorized to assemble or test a manufactured reactor, if applicable; (4) A description of the drug and alcohol testing and fitness determination process to be implemented through the licensee's or other entity's procedures, including the collection and testing facilities to be used, biological specimens to be collected, and sanctions to be imposed upon a confirmed FFD policy violation; and (5) A summary of the FFD performance monitoring and review program (PMRP), including the measures and thresholds required by paragraph (d)(1) of this section. (b) 
                            *FFD program implementation and availability.*
                             For the licensees and other entities in § 26.3(f), other than the holder of a manufacturing license (ML), the FFD program must be implemented no later than the start of construction activities, as defined in § 26.5, and maintained until the NRC's docketing of the license holder's certifications described in § 53.1070 of this chapter. For holders of an ML, the FFD program must be implemented no later than the start of activities that assemble the manufactured reactor and maintained until expiration of the ML. (c) 
                            *Criterion and analysis for an FFD program.*
                             For a licensee or other entity to implement an FFD program under § 26.604, the licensee or other entity must perform a site-specific analysis to demonstrate that the facility and its operation satisfy the criterion in § 53.860(a)(2) of this chapter. The licensee or other entity must maintain the analysis, including updates to reflect changes made to the staffing, FFD programs, or offsite support resources described in the analysis, to show that the facility and its operation continue to satisfy the criterion, until permanent cessation of operations under § 53.1070 of this chapter. (d) 
                            *FFD performance monitoring and review.*
                             A licensee or other entity must establish performance measures and associated thresholds as described in paragraph (d)(1) of this section and monitor the effectiveness of its FFD program by comparing performance data against these performance measures and thresholds, in a manner sufficient to satisfy the § 26.23 performance objectives. (1) 
                            *PRMP elements.*
                             The PMRP must be documented and maintained and include the following program elements: (i) 
                            *Performance measures.*
                             Performance measures must be identified and designed to monitor FFD program performance. (A) If the licensee or other entity is subject to the requirements in § 26.604, then the monitoring program must include performance measures for the following: the behavioral observation program; occurrence of FFD policy violations categorized by licensee employee, contractor/vendor, and labor category; and occurrence of individuals with potentially disqualifying information or who possessed FFD prohibited items. (B) If the licensee or other entity is subject to the requirements in § 26.604 and has implemented a drug testing program at its discretion or is subject to the requirements of § 26.605, then the monitoring program must include performance measures identified in paragraph (d)(1)(i)(A) of this section. This monitoring program must also include performance measures for the pre-access and random positive testing rates, random testing rate for licensee employees and contractor/vendors, and the number of subversion attempts categorized by licensee employee, contractor/vendor, and labor category. (ii) 
                            *Thresholds.*
                             Licensee- or other entity-specific thresholds for its site-specific performance measures must be established and used to facilitate corrective actions to maintain FFD program performance. Initial thresholds must be based on FFD performance data from comparable facilities subject to part 26, the licensee's or other entity's fleet-level program performance if applicable, and industry FFD performance data. (iii) 
                            *Monitoring program.*
                             Licensees and other entities must monitor the performance of their FFD programs against licensee- or other entity-established performance measures and thresholds as FFD performance data is received to determine whether a threshold has been exceeded. Licensees and other entities must perform year-to-year comparisons of site-specific performance; site-specific performance to the licensee's or other entity's fleet-level program performance, if applicable; and site-specific to industry performance. (iv) 
                            *Quantitative and qualitative reviews.*
                             The PMRP must include a documented review of the elements in paragraph (d)(1)(i) through (iii) of this section and the following qualitative elements. (A) 
                            *Worker protections.*
                             The review must include a documented assessment of the licensee's or other entity's implementation of the protections described in §§ 26.606(b)(1)(iii), 26.611, and 26.613. (B) 
                            *Laboratory test results and Medical Review Officer performance.*
                             The review must include a documented assessment of whether the actions taken by the Medical Review Officer (MRO) met the requirements in § 26.185 based on the laboratory test results reported under § 26.169. This review must include a comparative analysis between the point of collection testing and assessment (POCTA) screening result(s) and the corresponding specimen test results obtained from the U.S. Department of Health and Human 
                            ( print page 87032)
                            Services (HHS)-certified laboratory if the POCTA indicated a positive, adulterated, substituted, or invalid screening result or discrepant biological marker, to assess the effectiveness of the POCTA and to inform MRO decisions under § 26.185 or § 26.607(m)(6). (C) 
                            *Change control.*
                             The review must include a documented assessment of the changes made under paragraph (e) of this section to verify that the summation of program changes has not resulted in a reduction in FFD program effectiveness. (2) 
                            *Corrective actions.*
                             Corrective actions must be implemented to address when FFD performance meets a licensee-established performance threshold or to resolve a finding resulting from a qualitative review or audit in a manner that restores performance and corrects root causes, contributing causes, or both. (3) 
                            *Program review periodicity.*
                             The documented review in paragraph (d)(1)(iv) of this section must be conducted biennially to assess and modify licensee or other entity implementation of its FFD program. This documented review must demonstrate that the performance measures and thresholds are appropriate and adjusted as necessary based on site-level and licensee's or other entity's fleet-level, if applicable, program performance, and industry performance. (i) Identified program weaknesses and corrective actions must be summarized in the annual reporting requirement described in § 26.617(b)(2) or § 26.717, as applicable. (ii) The program review must be completed and approved by the licensee or other entity to support the reporting of PMRP weaknesses and corrective actions as required in paragraph (d)(3)(i) of this section every odd-numbered year, and the implementation of corrective actions before May 15 of that odd-numbered year. (e) 
                            *FFD program change control.*
                             (1) The licensee or other entity may make changes to its FFD program under this subpart if— (i) The licensee or other entity performs and retains an analysis demonstrating that the changes do not reduce the effectiveness of the FFD program; or (ii) The change was necessitated or justified by a change to part 26, laboratory processes or procedures, or guidance issued by the HHS or NRC, as implemented by the licensee or other entity though its procedures. (2) A licensee or other entity desiring to make a change that decreases FFD program effectiveness must implement a mitigating strategy so the FFD program, as revised, will continue to satisfy the performance objectives in § 26.23 and not result in a reduction in program effectiveness. (3) Except for phencyclidine, and notwithstanding paragraph (e)(1)(ii) of this section, the change control process may not be used to reduce the minimum panel of drugs to be tested in § 26.607(c)(1). (4) The licensee must retain a record of each change made under this section for a period of at least 5 years from the date the change was implemented and summarize this change in its annual FFD performance report required by § 26.617(b)(2) or § 26.717, as applicable.
 § 26.604
  (a) 
                            *FFD program.*
                             A licensee or other entity with an analysis that demonstrates that its facility and operation satisfy the criterion in § 26.603(c) may elect to establish, implement, and maintain an FFD program under this section. That FFD program must contain the following elements: (1) Applies to those individuals described in § 26.4, as applicable; and (2) Implements the following requirements and subparts in this part: (i) § 26.23, Performance objectives; (ii) § 26.603, General provisions; (iii) § 26.606, Written policies and procedures, (a) and, if applicable (b); (iv) § 26.608, FFD program training; (v) § 26.609, Behavioral observation; (vi) § 26.610, Sanctions; (vii) § 26.611, Protection of information; (viii) § 26.613, Appeals process; (ix) § 26.615, Audits; (x) § 26.617, Recordkeeping and reporting; (xi) § 26.619, Suitability and fitness determinations; (xii) Subpart A—Administrative Provisions; (xiii) Subpart I—Managing Fatigue; and (xiv) Subpart O—Inspections, Violations, and Penalties. (b) [Reserved]
 § 26.605
  (a) Licensees and other entities who satisfy the criterion in § 26.603(c), at their discretion, and licensees and other entities who do not satisfy the criterion in § 26.603(c), must establish, implement, and maintain an FFD program under this section either during construction activities as defined in § 26.5, or during activities performed under an ML that allows the assembly, testing, or both of a manufactured reactor, as applicable. This FFD program must contain the following elements: (1) Applies to those individuals described in § 26.4, as applicable; and, (2) Implements the following requirements and subparts in this part— (i) § 26.23, Performance objectives; (ii) § 26.603, General provisions; (iii) § 26.606, Written policy and procedures; (iv) § 26.607, Drug and alcohol testing; (v) § 26.608, FFD program training; (vi) § 26.609, Behavioral observation; (vii) § 26.610, Sanctions; (viii) § 26.611, Protection of information; (ix) § 26.613, Appeals process; (x) § 26.615, Audits; (xi) § 26.617, Recordkeeping and reporting; (xii) § 26.619, Suitability and fitness determinations; (xiii) Subpart A—Administrative Provisions; (xiv) Subpart I—Managing Fatigue, in the case of holders of an ML that allows the assembly, testing, or both of a manufactured reactor; and (xv) Subpart O—Inspections, Violations, and Penalties. (b) Licensees and other entities who satisfy the criterion in § 26.603(c), at their discretion, and licensees and other entities who do not satisfy the criterion in § 26.603(c), before the loading of fuel onsite into a reactor vessel; before receiving a manufactured reactor; or before individuals subject to part 26 operate, test, perform maintenance of, or direct the maintenance or surveillance of security-related equipment or equipment that a risk-informed evaluation process has shown to be significant to public health and safety, must establish, implement, and maintain an FFD program that— (1) Applies to those individuals described in § 26.4, as applicable; and, (2) Implements the following requirements and subparts— (i) § 26.23, Performance objectives; (ii) § 26.603, General provisions; (iii) § 26.606, Written policy and procedures; (iv) § 26.607, Drug and alcohol testing; (v) § 26.608, FFD program training; (vi) § 26.609, Behavioral observation; (vii) § 26.611, Protection of information; (viii) § 26.613, Appeals process; (ix) § 26.615, Audits; (x) Subpart A—Administrative Provisions; (xi) Subpart C—Granting and Maintaining Authorization; (xii) Subpart D—Management Actions and Sanctions to be Imposed; (xiii) Subpart H—Determining Fitness-for-Duty Policy Violations and 
                            ( print page 87033)
                            Determining Fitness, unless using the HHS Guidelines for MRO evaluation of drug test results, and determining fitness; (xiv) Subpart I—Managing Fatigue; (xv) Subpart N—Recordkeeping and Reporting Requirements; and (xvi) Subpart O—Inspections, Violations, and Penalties.
 § 26.606
  (a) Licensees and other entities that implement an FFD program under this subpart must ensure that— (1) A written FFD policy statement is provided to each individual who is subject to the program before the individual is subject to behavioral observation, drug and alcohol testing, or both. (2) The FFD policy statement describes the performance objectives in § 26.23. (3) The FFD policy statement describes the minimum days off requirements in § 26.205(d)(3) or maximum average work hours requirements in § 26.205(d)(7). (4) The FFD policy statement must be written in sufficient detail to provide affected individuals with information on what is expected of them and what consequences may result from a lack of adherence to the policy, including those elements described in § 26.606(b), part 26-required sanctions, and required medical/clinical treatment and follow-up testing for FFD policy violations. (5) The FFD policy statement describes the individual's responsibilities to report for work in a physiological and psychological condition that enables the safe and competent performance of assigned duties and responsibilities and inform a licensee- or other entity-designated representative when the individual determines that this cannot be accomplished. (b) Licensees and other entities must establish, implement, and maintain written procedures that address the following topics: (1) If implementing a drug and alcohol testing program under this subpart, (i) The methods and techniques to collect and test for drugs and alcohol and for the shipping and temporary storage of biological specimens used for drug testing at HHS-certified laboratories, (ii) The urine specimen volumes, techniques for split specimen collections, and the acceptability of a urine specimen as described in § 26.111 or as described in the HHS Guidelines, (iii) Protecting the privacy of an individual who provides a specimen, protecting the integrity of the specimen, and ensuring that the test results are valid and attributable to the correct individual, and (iv) If the licensee or other entity elects to use the HHS Guidelines, the name of the specific HHS Guideline and revision being implemented by the licensee or other entity and a description of the specific sections in the guideline that are being implemented in the procedure, including specimen collections, drug testing, and evaluation of test results. (2) The immediate and follow-up actions that will be taken, and the procedures to be used, in those cases in which individuals who are subject to the FFD program: (i) Have been involved in the use, sale, or possession of illegal substances, illegal drugs, or illicit substances; (ii) Are impaired by any illegal substances, illegal drugs, or illicit substances or the consumption of alcohol as determined by behavioral observation or a test that measures blood alcohol concentration; (iii) If drug and alcohol testing is conducted, attempted to subvert the testing process by adulterating or diluting specimens (
                            *in vivo*
                             or 
                            *in vitro*), substituting specimens, or by any other means; (iv) If drug and alcohol testing is conducted, refused to provide a specimen for analysis or follow instructions provided by FFD program personnel; (v) Had legal action taken relating to drug or alcohol use; or (vi) Demonstrated character or actions indicating that the individual cannot be trusted or relied upon to perform those duties and responsibilities or maintain access to NRC-licensed facilities, special nuclear material (SNM), or sensitive information. (3) The process, including the duties and responsibilities of FFD program personnel, to be followed if an individual's behavior or condition raises a concern regarding the possible use, sale, or possession of illegal drugs on- or offsite; the possible use or possession of alcohol on the NRC-licensed facility; impairment from any cause that in any way could adversely affect the individual's ability to safely and competently perform the individual's duties; or the receipt of credible information indicating that the individual cannot be trusted or relied on to perform those duties and responsibilities making the individual subject to this part. (4) Operation and oversight of an onsite or offsite collection facility. (5) The fatigue management requirements in §§ 26.202(b) and either 26.205(d)(3) or (d)(7). (6) Measures to prevent subversion of drug and alcohol tests conducted onsite and offsite.
 § 26.607
  Licensees and other entities implementing § 26.604, at their discretion, and licensees and other entities implementing § 26.605 must perform drug and alcohol testing that complies with the following requirements— (a) 
                            *Split specimens.*
                             Split specimen collections of oral fluid or urine must be used for the test conditions described in paragraph (b) of this section. A split specimen collection need not be used if the licensee or other entity elects to use a POCTA device for a screening test conducted during random testing under paragraphs (b)(2) and (h) of this section or a protected area portal monitor indication that drugs or alcohol were detected under paragraph (j) of this section. Testing of the split specimen (specimen B) requires the donor's permission unless ordered by the MRO to resolve an invalid test result obtained for specimen A. (b) 
                            *Test conditions.*
                             Individuals identified in § 26.4 must be subject to drug and alcohol testing under the following conditions: (1) 
                            *Pre-access.*
                             A pre-access test must be conducted for drugs and alcohol before performing or directing the conduct of roles and responsibilities making the individual subject to this subpart or being granted unescorted access to the protected areas of the NRC-licensed facility. A pre-access test must have been conducted no more than 14 days before the individual is granted unescorted access. (2) 
                            *Random.*
                             Random testing for drugs and alcohol must— (i) Be administered in a manner that provides reasonable assurance that individuals are unable to predict the time periods during which specimens will be collected; (ii) Require individuals who are selected for random testing to report to the onsite collection site as soon as reasonably practicable after notification, within the time period specified in the FFD program procedure; (iii) Ensure that all individuals in the population that is subject to random testing on a given day have an equal probability of being selected and tested; (iv) Ensure that an individual completing a test is immediately eligible for another random test; and (v) Ensure that the sampling process used to select individuals for random 
                            ( print page 87034)
                            testing provides that the number of random tests performed annually is equal to at least 50 percent for licensee employees and 50 percent for contractor/vendors at the NRC-licensed site. (3) 
                            *For-cause.*
                             A for-cause drug test, alcohol test, or both, must be conducted onsite in response to an individual's observed behavior or physical condition indicating possible substance abuse or after receiving credible information that an individual is engaging in substance abuse, as defined in § 26.5; (4) 
                            *Post-event.*
                             A post-event test for drugs and alcohol must be conducted— (i) As soon as practical after an event involving a human error that was committed by an individual specified in § 26.4, where the human error may have caused or contributed to the event. This test must be conducted onsite unless the individual requires offsite medical care. The licensee or other entity must test the individual(s) who committed or directed the error and need not test individuals who were affected by the event and whose actions likely did not cause or contribute to the event. The licensee or other entity must describe in its procedures what constitutes a human error. (ii) Within 4 hours of an event unless immediate medical intervention precludes the conduct of the test on the individual(s) who caused or contributed to the accident(s), if the event results in— (A) An illness or personal injury to any individual which results in death, days away from work, restricted work, transfer to another job, medical treatment beyond first aid, loss of consciousness, or other significant illness or injury, as diagnosed by a licensee- or other entity-designated physician or other licensed health care professional, even if the illness or injury does not result in death, days away from work, restricted work or job transfer, medical treatment beyond first aid, or loss of consciousness; or (B) Damage to any safety- or security-related structures, systems, and components; and (5) 
                            *Follow-up.*
                             An individual subject to part 26 who has violated the FFD policy for substance use or abuse, or the sale, use, or possession of illegal drugs must be subject to a follow-up series of tests for drugs, alcohol, or both to verify an individual's continued abstinence from substance abuse. (c) 
                            *Urine and oral fluid specimens.*
                             (1) All urine or oral fluid specimens must be subject to validity testing, including an adulterant and biological marker, and tested for the substances listed in § 26.31(d)(1), except as allowed by § 26.603(e)(3). (2) For the use of urine as the biological specimen to be tested, the following requirements must be implemented— (i) § 26.115, Collecting a urine specimen under direct observation; (ii) § 26.119, Determining “shy” bladder; and (iii) § 26.163, Cutoff levels for drugs and drug metabolites, (a)(2) regarding special analysis testing. (3) For alcohol testing onsite, the following requirements must be implemented— (i) § 26.91, Acceptable devices for conducting initial and confirmatory tests for alcohol and methods of use; (ii) § 26.93, Preparing for alcohol testing; (iii) § 26.95, Conducting an initial test for alcohol using a breath specimen; (iv) § 26.97, Collecting oral fluid specimens for alcohol and drug testing; (v) § 26.99, Determining the need for a confirmatory test for alcohol; (vi) § 26.101, Conducting a confirmatory test for alcohol; and, (vii) § 26.103, Determining a confirmed positive test result for alcohol. (4) For all test conditions in paragraph (b) of this section, except for the use of a POCTA screening device in paragraph (h) of this section, and for MRO-directed tests under § 26.185, drug testing must be performed at an HHS-certified laboratory for the specific biological specimen to be tested. Only HHS-certified laboratory test results from urine and oral fluid specimens may be used for the issuance of a part 26-required sanction. The licensee or other entity must establish and maintain a contract with a primary and a back-up HHS-certified laboratory (with a different Certifying Scientist) for the specimen(s) to be tested. These contracts must stipulate that the laboratories are subject to inspection or audit by the licensee or other entity and that records and documents must be provided and/or able to be photocopied and removed from the premises to support the inspection or audit. (d) 
                            *Privacy and integrity.*
                             The specimen collection and drug and alcohol testing procedures of FFD programs must protect the donor's privacy and the integrity of the specimen and implement quality controls to ensure that test results are valid and attributable to the correct individual. (e) 
                            *Offsite collection facilities.*
                             At the licensee's or other entity's discretion, specimen collections and alcohol testing may be conducted at a local hospital or other facility licensed to conduct specimen collections and perform alcohol testing and audited by the State or a State-designated entity. The licensee or other entity must audit these facilities, if used, before their initial use and then on a biennial basis to confirm that the facility procedures are comparable to those described in subpart E of this part or the HHS Guidelines for urine and oral fluid. (f) 
                            *Initial testing.*
                             A licensee or other entity subject to this subpart performing an initial test must use an immunoassay, or an alternative technology established in its FFD program through § 26.603(e), that satisfies the requirements of the U.S. Food and Drug Administration (FDA) for commercial distribution. Specimens that yield positive, adulterated, substituted, or invalid initial validity or drug test results or discrepant biological markers must be subject to confirmatory testing by an HHS-certified laboratory, certified for that biological specimen, except for invalid specimens that cannot be tested. (g) 
                            *Oral fluid testing.*
                             If the licensee or other entity elects to use oral fluid for drug or alcohol testing, the collection, packaging, and temporary storage of the drug or alcohol test device, and shipment of an oral fluid specimen to an HHS-certified laboratory or the collection of an oral fluid specimen for alcohol testing must be performed in accordance with licensee- or other entity-established procedures based either on the requirements in part 26 or the procedures in HHS Guidelines identified by the licensee or other entity in § 26.606(b)(1)(iv). The device must have received premarket approval from the FDA and must not expire before laboratory testing. The drugs, drug metabolites, initial and confirmatory testing cutoffs, and biological markers, if applicable, must be those established by HHS for oral fluid testing and the alcohol cutoffs in this part or, if not established by HHS or the NRC for the panel of drugs and drug metabolites to be tested, as determined and documented by a forensic toxicologist review conducted pursuant to § 26.31(d)(1)(i)(D). (h) 
                            *Point of collection testing and assessment.*
                             (1) If the licensee or other entity elects to use a POCTA device, then it may only be used for pre-access and random drug and alcohol initial testing in paragraph (b) of this section, the alcohol testing process in paragraph (c)(3) of this section, and the portal area screening process in paragraph (j) of this section. Before the licensee or other entity uses a POCTA device, a forensic toxicologist must review and document 
                            ( print page 87035)
                            their evaluation that the validity and accuracy of the device for alcohol and/or the drugs and drug metabolites listed in § 26.31(d) are comparable to the performance achieved by initial testing conducted using a similar technology at an HHS-certified laboratory. For initial testing of drugs and drug metabolites using a POCTA device, this review must include a documented evaluation of POCTA device performance against the requirements in § 26.161(b) for a urine specimen or the procedures in the HHS Guidelines for urine or oral fluid, as implemented by the licensee or other entity through its procedures. (2) If the performance of the POCTA device is not comparable to that achieved from initial testing conducted by an HHS-certified laboratory as determined by the forensic toxicologist, then the licensee or other entity must implement a mitigating strategy to maintain program effectiveness under § 26.603(e)(2), as applicable. (3) The licensee and other entity must implement procedures for the use of a POCTA that ensures the effectiveness of the collection process, assessment of the screening results, and prevention of subversion attempts. (4) If the use of a POCTA device indicates a discrepant biological marker or that a test result exceeds the initial test cutoff, the specimen is invalid, or the individual subverted the drug or alcohol test, then the individual must be immediately removed from duties, responsibilities, and access making the individual subject to this subpart. (i) The individual must be subject to an immediate drug and alcohol test using the alcohol testing process in paragraph (c)(3) of this section for a positive alcohol screen and either oral fluid or urine by a collection kit that is not a POCTA device, but of the same type of biological specimen collected by the POCTA, for validity, if required, and initial and confirmatory testing by an HHS-certified laboratory. (ii) If this individual shows any signs of impairment, the individual's authorization must be temporarily removed until the MRO reviews the laboratory test result(s), interviews the individual, and performs a determination of fitness under § 26.189 or § 26.619, as applicable, that enables the restoration of authorization. (i) 
                            *Hair testing.*
                             The testing of hair specimens may only be used to inform a licensee's or other entity's determination of whether the individual is trustworthy and reliable under the test condition in paragraph (b)(1) of this section to supplement the information gained from a pre-access test using oral fluid or urine as the test specimen and must be conducted at an HHS-certified laboratory certified for hair specimens. (1) If used, this process must be described in the licensee's or other entity's FFD policy and described in detail in its procedure. The panel of drugs and drug metabolites to be evaluated must only include those listed as Schedule I or II of section 202 of the Controlled Substances Act [21 U.S.C. 812]. The collection, packaging, and temporary storage of a hair specimen and shipment of the specimen to an HHS-certified laboratory must be conducted in accordance with the HHS Guidelines. The test kit must be FDA cleared, and licensee- or other entity-designated FFD program personnel must conduct the collection, packaging, temporary storage, shipping, and custody and control of the specimen. (2) Before the licensee or other entity begins to conduct hair testing, the initial and confirmatory testing cutoffs must be the cutoffs established by HHS for hair testing or, if not established by HHS or the NRC, as determined by a forensic toxicologist review conducted pursuant to § 26.31(d)(1)(i)(D). (3) Confirmed positive test results must be considered potentially disqualifying FFD information until proven otherwise by a review under § 26.613. Sanctions under this subpart must not be issued for any FFD policy violation involving a drug test using a hair specimen unless the licensee or other entity determines that the individual subverted, as defined in § 26.5, the hair test. (j) 
                            *Portal area screening.*
                             A non-invasive point of collection testing instrument may be used to screen individuals for drugs, drug metabolites, and alcohol before the individuals' entry into or exit from a protected or vital area. (1) If a licensee or other entity uses such an instrument, then before such use, a forensic toxicologist must review the instrument and document an evaluation that the instrument and setpoints used in the instrument are acceptable for use for the detection and screening of the drugs and drug metabolites selected for screening from the panel of drugs and drug metabolites to be tested under the FFD program and alcohol and its metabolites. (2) The instrument must be operated in accordance with the manufacturer's specifications. If screening detects the presence of drugs, drug metabolites, or alcohol at or above the instrument set point(s), the individual screened by the instrument must be subject to a POCTA screening test using the process described in paragraph (h) of this section or an oral fluid or urine test that is sent to an HHS-certified laboratory. (3) A part 26 sanction may not be issued to an individual based solely on a portal area screening instrument detection that drugs or alcohol exceed the instrument's established setpoint. (k) 
                            *Blood testing.*
                             The testing of blood specimens may only be conducted under the order of the licensee- or other entity-designated MRO for a valid medical reason as confirmed by the MRO pursuant to § 26.31(d)(5). This specimen must be subject to testing by a laboratory that satisfies quality control requirements that are comparable to those required for certification by the HHS. (l) 
                            *Custody-and-control form.*
                             For the collection and packaging of urine, oral fluid, and hair specimens, the licensee or other entity must use a custody-and-control form approved by the U.S. Office of Management and Budget. For the use of a POCTA device, the licensee or other entity must implement a licensee- or other entity-approved and -maintained procedure that ensures the reliability of the tracking, handling, and storage of a specimen from the point of specimen collection to the final disposition of the specimen and the reliability of an identification system to uniquely assign the specimen to the donor. (m) 
                            *Medical Review Officer.*
                             Licensees or other entities must— (1) Require their designated MRO to review positive, adulterated, substituted, and dilute confirmatory drug and validity test results and test results of questionable validity to determine whether the donor has violated the FFD policy for urine and oral fluid specimens. The review must be completed before reporting the results to the individual designated by the licensee or other entity to assess authorization or perform the suitability and fitness determinations required under § 26.619, or, if required, that are described in subpart H of this part. (2) Require their MRO to satisfy the requirements in § 26.183 and, prior to conducting any activities under this part, attend and pass a medical- or clinical-based training session to improve his/her knowledge of MRO duties and responsibilities, drug and alcohol testing processes and procedures, and evaluation of drug testing results. This training session must be conducted by a nationally recognized MRO training and certification organization that has been assessed by the licensee's or other entity's FFD program personnel to include the technical elements an MRO must implement under § 26.185. An 
                            ( print page 87036)
                            MRO who performed the duties and responsibilities in §§ 26.185 and 26.187 for at least 3 continuous years in the last 10 years prior to being hired or contracted by the licensee or other entity satisfies the requirements in this paragraph. (3) Require their MRO to attend a medical- or clinical-based training session on a triennial basis to improve his/her knowledge of changes in drug and alcohol testing processes and procedures and evaluation of drug testing results. (4) Require their MRO to determine whether a biological specimen is positive, adulterated, substituted, dilute or of questionable validity by implementing the requirements in § 26.185 or the HHS Guidelines through the licensee's or other entity's procedures. (i) If § 26.185 or the HHS Guidelines, as used by the licensee or other entity in its procedures, are insufficient to make this determination, then guidance issued by a State agency in the state in which the NRC-licensed facility is located, Federal agencies, or nationally recognized MRO training and certification organizations may be used to inform an MRO determination. (ii) An MRO need not review a confirmed alcohol positive test result determined by an evidentiary breath testing device under paragraphs (c)(3)(vi) and (vii) of this section. (5) Require their MRO to determine and approve the use of oral fluid or urine as an alternative biological specimen when the donor cannot provide a specimen for testing. This determination and the retest must be documented and completed as soon as reasonably practicable. (6) Require the MRO to review all specimens screened and tested associated with a drug-related FFD policy violation. This review includes POCTA, split specimens, and all specimens taken to resolve a discrepant condition, such as a possible subversion attempt, impairment without a known cause, or a donor-requested or MRO-directed re-test. To resolve a discrepant condition, the MRO is authorized to test a specimen for a biological marker, adulterants, or additional drugs. (n) 
                            *Limitations of screening and testing.*
                             Specimens collected under NRC regulations may only be designated or approved for screening and testing as described in this part and may not be used to conduct any other analysis or test without the written permission of the donor. Analyses, screens, and tests that may not be conducted include, but are not limited to, DNA testing, serological typing, or any other medical or genetic test used for diagnostic or specimen identification purposes. No biological specimens may be passively sampled and analyzed in a manner different than described in this subpart. (o) 
                            *Specimen collectors.*
                             All onsite specimen collections, except a collection by a portal area screening instrument in paragraph (j) of this section, must be conducted by licensee- or other entity-designated and -trained personnel.
 § 26.608
  (a) 
                            *FFD program training.*
                             (1) Individuals must be trained in the FFD policy and procedure, including fatigue management, and their FFD program responsibilities. Individuals who collect specimens for testing or screening must also be trained in specimen collector duties and responsibilities, including, at a minimum, specimen collection, custody and control, identification and response to subversion attempts, and privacy. For licensees and other entities of commercial nuclear plants, the FFD program training program must use a systems approach to training as defined in § 53.725 of this chapter and described in § 53.830 of this chapter for those individuals in § 26.4. (2) FFD program training must include training on the behavioral observation program. The behavioral observation program training must include the detection of physiological behaviors or conditions that may indicate— (i) Possible use, sale, or possession of illegal drugs or illicit drugs, or substance abuse on- or offsite; (ii) Use or possession of alcohol onsite or use while on duty offsite; (iii) Impairment from fatigue or any cause that, if left unattended, could result in inattentiveness or human errors; and (iv) Any individual's inability to safely and competently perform assigned duties and responsibilities or act in a trustworthy and reliable manner while having access to protected areas, SNM, or sensitive information. (3) Training must explain that an individual's FFD policy violation will— (i) Subject the individual to an FFD program-required sanction designed to preclude recurrence of an FFD policy violation; (ii) Contribute to the licensee's or other entity's assessment of whether the individual can be trusted and relied upon to safely and competently perform the assigned duties and responsibilities making the individual subject to this subpart; (iii) Be used to inform the licensee's or other entity's insider mitigation and access authorization programs under §§ 73.55, 73.56, 73.100 or 73.120 of this chapter; and (iv) Be used to inform other NRC licensees and other entities subject to part 26 when FFD program information is requested to support authorization determinations under subpart C of this part or §§ 73.56 or 73.120 of this chapter. (b) 
                            *Training and assessments.*
                             Training and a trainee assessment must be conducted before pre-access testing, and refresher training and trainee assessments must be conducted periodically thereafter. (c) 
                            *Training program review.*
                             The licensee or other entity must periodically evaluate its FFD training program and revise it as appropriate to reflect industry experience as well as applicable changes to the regulations in this part, the HHS Guidelines, if used, and specimen collection and testing processes implemented by the licensee or other entity.
 § 26.609
  (a) Licensees and other entities must ensure that the individuals who are subject to this subpart are subject to behavioral observation and that behavioral observation is performed by all individuals subject to this subpart. (b) Licensees and other entities must require all individuals subject to the FFD program to report to the licensee- or other entity-designated representative any onsite or offsite behaviors or activities by individuals subject to this part that may constitute an unreasonable risk to the safety or security of the NRC-licensed facility or SNM or may cause harm to others. This reporting must include any information relating to character or reputation of the individual indicating that the individual cannot be trusted or relied upon to perform those duties and responsibilities or maintain access to NRC-licensed facilities, SNM, or sensitive information that makes them subject to part 26. (c) Behavioral observation must be performed visually, in-person, and, when necessary, remotely by live video and audible streaming and capture, to observe the behavior of individuals in the workforce subject to the requirements in this subpart. (d) Not withstanding paragraph (c) of this section, for a reactor facility where individual task loading does not allow for the effective conduct of behavior observation in addition to assigned operational tasks, the licensee or other entity must implement a live video and audible streaming and capture system to conduct behavioral observation of 
                            ( print page 87037)
                            persons licensed to operate under 10 CFR part 53 who manipulate the controls of any commercial nuclear plant licensed under 10 CFR part 53.
 § 26.610
  Licensees and other entities that implement an FFD program under this subpart must establish sanctions for FFD policy violations that, at a minimum, prohibit the individuals specified in § 26.4 from being assigned to perform or direct those duties and responsibilities or maintaining authorization making them subject to this subpart. The severity of the sanction must escalate with the number of occurrences and severity of the FFD policy violation. The sanction must be long enough to act as a deterrent and, if the individual is retained as a licensee employee or contractor/vendor, facilitate the individual to complete counseling or treatment. The sanctions must include a minimum 5-year denial of access to the NRC-licensed facility for any individual who is determined to have been involved in the sale, use, or possession of illegal drugs or the consumption of alcohol within a protected area of any facility licensed under part 53 of this chapter or within a transporter's facility or vehicle used in the conveyance of formula quantities of strategic SNM while the individual is subject to this subpart, and a permanent denial of access to the NRC-licensed facility for three FFD policy violations or any subversion attempt of any drug or alcohol test or screening process, including subversion attempts at any licensee or other entity subject to this part.
 § 26.611
  (a) Licensees and other entities that collect personal information about an individual for the purpose of complying with this subpart must establish and maintain a system of files and procedures to prevent unauthorized disclosure. (b) Licensees and other entities must obtain a signed consent that documents the individual's acceptance of being subject to the FFD program and authorizes the disclosure of the personal information collected and maintained under this subpart, except for disclosures to the individuals and entities specified in § 26.37(b)(1) through (b)(6), (b)(8), and persons deciding matters under review in § 26.613. This signed and dated consent must be obtained before making the individual subject to the FFD program.
 § 26.613
  Licensees and other entities that implement an FFD program under this subpart must establish and implement procedures for the review of a determination that an individual in § 26.4 has violated the FFD policy. The procedure must provide for an objective and impartial review of the facts related to the determination that the individual has violated the FFD policy and a schedule for the completion of the review.
 § 26.615
  (a) Licensees and other entities that implement an FFD program under this subpart must audit their programs at a frequency that ensures the continuing effectiveness of their FFD program, FFD program elements that are provided by C/Vs, and the FFD programs of C/Vs that are accepted by the licensee or other entity. Corrective actions must be as soon as reasonably practicable to resolve any problems identified in an audit and preclude recurrence. (b) The subject matter, scope, and frequency of audits must be revised as necessary to improve or maintain program performance based on findings resulting from licensee or other entity implementation of its FFD PMRP in § 26.603(d). (c) Licensees and other entities may conduct joint audits or accept audits of C/Vs so long as the audit addresses the relevant services of the C/Vs. (d) Licensees and other entities must audit HHS-certified laboratories unless the licensee's or other entity's panel of drugs and drug metabolites to be tested is equivalent to the panel by which the laboratory is certified by HHS or is subject to the standards and procedures for drug testing and evaluation used by the laboratory under the HHS Guidelines. Licensees and other entities must audit any hospital or other facility licensed by the State (or State-designated entity) if used to conduct specimen collections and perform alcohol testing under this part on a biennial basis to confirm that the facility procedures are comparable to those described in subpart E of this part, for urine and oral fluid.
 § 26.617
  (a) Licensees and other entities that implement FFD programs under this subpart must ensure that records pertaining to the administration of their program, which may be stored and archived electronically, are maintained so that they are available for NRC inspection purposes and for any legal proceedings resulting from the administration of the program. Records pertaining to the administration of the FFD program and FFD performance data required by § 26.717 must be retained until license termination. (b) Licensees and other entities must make the following reports: (1) Reports to the NRC Operations Center by telephone within 24 hours after the licensee or other entity discovers any intentional act that casts doubt on the integrity of the FFD program and any programmatic failure, degradation, or discovered vulnerability of the FFD program that may permit undetected drug or alcohol use or abuse by individuals who are subject to this subpart. These events must be reported under this subpart, rather than under the provisions of § 73.1200 of this chapter; and (2) Annual program performance reports for the FFD program, including the FFD program performance data listed in § 26.717(b), as applicable. Licensees and other entities must submit FFD program performance data (for January through December) to the NRC annually, before March 1 of the following year and must use unexpired NRC-provided forms for the electronic submission of FFD information to the NRC. (c) Licensees and other entities subject to this subpart must describe in sufficient detail to support an authorization determination, an individual's FFD policy violation (while protecting privacy information under § 26.611) and FFD program weakness to NRC, licensees, and other entities subject to this part when requested to support authorization determinations under subpart C of this part or § 73.120 of this chapter, as applicable, or to support licensee or other entity performance monitoring.
 § 26.619
  Licensees and other entities that implement FFD programs under this subpart must develop, implement, and maintain procedures for evaluating whether to assign individuals to perform or direct those duties and responsibilities making them subject to this subpart. A suitability or fitness determination conducted for cause must be performed face to face. A suitability or fitness determination conducted for cause may be performed remotely using electronic communications only when supported by someone who is present in-person with the individual being assessed, and that supporting person must be trained in accordance with the requirements of either §§ 26.29 or 26.608.
 94.Revise § 26.709 to read as follows: § 26.709
  (a) The requirements of this subpart apply to the FFD programs of licensees and other entities specified in § 26.3(a) through (d), except for FFD programs that are implemented under subpart K of this part. (b) The requirements in this subpart apply to the FFD programs of licensees and other entities specified in § 26.3(f) that elect not to implement the requirements in subpart M or elect to implement the requirements in § 26.605(b).
 § 26.711
 95.In § 26.711, in paragraphs (c) and (d), remove the phrase “(c) and (d),” and add in its place the phrase “(c), (d), and (f),”. § 26.825
 96.In § 26.825, in paragraph (b) add remove the phrase “§§ 26.1, 26.3, 26.5, 26.7, 26.8, 26.9, 26.11, 26.51, 26.81, 26.121, 26.151, 26.181, 26.201, 26.823, and 26.825” and add in its place the phrase “§§ 26.1, 26.3, 26.5, 26.7, 26.8, 26.9, 26.11, 26.51, 26.81, 26.121, 26.151, 26.181, 26.201, 26.601, 26.823, and 26.825”. PART 30—RULES OF GENERAL APPLICABILITY TO DOMESTIC LICENSING OF BYPRODUCT MATERIAL
 97.The authority citation for part 30 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 11, 81, 161, 181, 182, 183, 184, 186, 187, 223, 234, 274 (42 U.S.C. 2014, 2111, 2201, 2231, 2232, 2233, 2234, 2236, 2237, 2273, 2282, 2021); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); 44 U.S.C. 3504 note. 98.In § 30.4, revise the definition for “ *Utilization facility*
                        ” to read as follows: § 30.4
  
 *Utilization facility*
                             means a utilization facility as defined in the regulations contained in part 50 or part 53 of this chapter;
 99.In § 30.50, revise paragraph (c)(3) to read as follows: § 30.50
  
 (c) * * * (3) The provisions of this section do not apply to licensees subject to the notification requirements in §§ 50.72 or 53.1630 of this chapter. They do apply to those part 50 licensees possessing material licensed under this part, who are not subject to the notification requirements in § 50.72 of this chapter.
 PART 40—DOMESTIC LICENSING OF SOURCE MATERIAL
 100.The authority citation for part 40 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 62, 63, 64, 65, 69, 81, 83, 84, 122, 161, 181, 182, 183, 184, 186, 187, 193, 223, 234, 274, 275 (42 U.S.C. 2092, 2093, 2094, 2095, 2099, 2111, 2113, 2114, 2152, 2201, 2231, 2232, 2233, 2234, 2236, 2237, 2243, 2273, 2282, 2021, 2022); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); Uranium Mill Tailings Radiation Control Act of 1978, sec. 104 (42 U.S.C. 7914); 44 U.S.C. 3504 note. 101.In § 40.60, revise paragraph (c)(3) to read as follows: § 40.60
  
 (c) * * * (3) The provisions of this section do not apply to licensees subject to the notification requirements in §§ 50.72 or 53.1630 of this chapter. They do apply to those part 50 licensees possessing material licensed under this part who are not subject to the notification requirements in § 50.72 of this chapter.
 PART 50—DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES
 102.The authority citation for part 50 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 11, 101, 102, 103, 104, 105, 108, 122, 147, 149, 161, 181, 182, 183, 184, 185, 186, 187, 189, 223, 234 (42 U.S.C. 2014, 2131, 2132, 2133, 2134, 2135, 2138, 2152, 2167, 2169, 2201, 2231, 2232, 2233, 2234, 2235, 2236, 2237, 2239, 2273, 2282); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); Nuclear Waste Policy Act of 1982, sec. 306(42 U.S.C. 10226); National Environmental Policy Act of 1969 (42 U.S.C. 4332); 44 U.S.C. 3504 note; Sec. 109, Pub. L. 96-295, 94 Stat. 783. 103.In § 50.47, revise paragraphs (a)(1) and (e) to read as follows: § 50.47
  (a)(1)(i) Except as provided in paragraph (d) of this section, no initial operating license for a nuclear power reactor will be issued under this part or under part 53 of this chapter unless a finding is made by the NRC that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. No finding under this section is necessary for issuance of a renewed nuclear power reactor operating license. (ii) No initial combined license under parts 52 or 53 of this chapter will be issued unless a finding is made by the NRC that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. No finding under this section is necessary for issuance of a renewed combined license. (iii) If an application for an early site permit under subpart A of part 52 of this chapter includes complete and integrated emergency plans under § 52.17(b)(2)(ii) of this chapter or an application for an early site permit under subpart H of part 53 of this chapter includes complete and integrated emergency plans under § 53.1146(b)(2)(ii) of this chapter, no early site permit will be issued unless a finding is made by the NRC that the emergency plans provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. (iv) If an application for an early site permit proposes major features of the emergency plans under §§ 52.17(b)(2)(i) or 53.1146(b)(2)(i) of this chapter, no early site permit will be issued unless a finding is made by the NRC that the major features are acceptable in accordance with the applicable standards of either § 50.47 and appendix E to this part or the applicable requirements of § 50.160, within the scope of emergency preparedness matters addressed in the major features. 
 (e) Notwithstanding the requirements of paragraph (b) of this section and the provisions of § 52.103 or § 53.1452 of this chapter, a holder of a combined license under part 52 or part 53 of this chapter, as applicable, that is complying with the requirements of § 50.47(b) and appendix E to this part may not load fuel or operate except as provided in accordance with appendix E to this part and § 50.54(gg), and a holder of a combined license under part 52 or part 53 of this chapter that is complying with the requirements of § 50.160 may not load fuel or operate except as provided in accordance with § 50.160(c)(2) and § 50.54(gg).
 104.In § 50.54, revise paragraphs (q)(2), (q)(4), and (gg)(1) introductory text to read as follows: § 50.54
  
 (q) * * * (2)(i) Except as provided in paragraph (q)(2)(ii) of this section, a holder of a license under this part, or a combined license under parts 52 or 53 of this chapter after the Commission makes the finding under §§ 52.103(g) or 53.1452(g) 
                            ( print page 87039)
                            of this chapter, as applicable, shall follow and maintain the effectiveness of an emergency plan that meets the requirements in appendix E to this part and, for nuclear power reactor licensees, the planning standards of § 50.47(b). (ii) A holder of a license under this part for a non-power production or utilization facility, a holder of a license under this part or part 53 of this chapter for a small modular reactor or a non-light-water reactor, or a holder of a combined license under parts 52 or 53 of this chapter after the Commission makes the finding under §§ 52.103(g) or 53.1452(g) of this chapter, as applicable, for a small modular reactor or a non-light-water reactor, shall follow and maintain the effectiveness of either an emergency plan that meets the requirements in § 50.160 or an emergency plan that meets the requirements in appendix E to this part and, for nuclear power reactor licensees, the planning standards of § 50.47(b). 
 (4) The changes to a licensee's emergency plan that reduce the effectiveness of the plan as defined in paragraph (q)(1)(iv) of this section may not be implemented without prior approval by the NRC. A licensee desiring to make such a change shall submit an application for an amendment to its license. In addition to the filing requirements of §§ 50.90 and 50.91 or §§ 53.1510 and 53.1515 of this chapter, as applicable, the request must include all emergency plan pages affected by that change and must be accompanied by a forwarding letter identifying the change, the reason for the change, and the basis for concluding that the licensee's emergency plan, as revised, will continue to meet either the requirements in § 50.160 or the requirements in appendix E to this part and, for nuclear power reactor licensees, the planning standards of § 50.47(b). 
 (gg)(1) Notwithstanding §§ 52.103 or 53.1452 of this chapter, if following the conduct of the exercise required by paragraph IV.f.2.a of appendix E to this part or § 50.160(c)(2), as applicable, FEMA identifies one or more deficiencies in the state of offsite emergency preparedness, the holder of a combined license under 10 CFR part 52 or under 10 CFR part 53, as applicable, may operate at up to 5 percent of rated thermal power only if the Commission finds that the state of onsite emergency preparedness provides reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. The NRC will base this finding on its assessment of the applicant's onsite emergency plans against the pertinent standards in either § 50.47 and appendix E to this part, or § 50.160, as applicable. Review of the applicant's emergency plans will include the following standards with offsite aspects:
 105.In § 50.160, revise paragraphs (b)(3) and (c)(2) to read as follows: § 50.160
  
 (b) * * * (3) 
                            *Emergency planning zone.*
                             For an applicant whose analysis required by § 50.33(g)(2) or § 53.1109(g)(2) of this chapter meets the criteria in § 50.33(g)(2)(i) or § 53.1109(g)(2)(i) of this chapter, as applicable, determine and describe the boundary and physical characteristics of the EPZ in the emergency plan. 
 (c) * * * (2) A holder of a combined license issued under parts 52 or 53 of this chapter before the Commission has made the finding under §§ 52.103(g) or 53.1452(g) of this chapter, as applicable, must establish, implement, and maintain an emergency preparedness program that meets the requirements of paragraph (b) of this section, as described in the approved emergency plan and license, and conduct an initial exercise to demonstrate this compliance within 2 years before the scheduled date for initial loading of fuel (or, for a fueled manufactured reactor, within 2 years before the scheduled date for initiating the physical removal of any one of the independent physical mechanisms to prevent criticality required under § 53.620(d)(1) of this chapter).
 106.In appendix B to part 50, revise the first paragraph in the Introduction section, the first paragraph of section III, Design Control, and section IV, Procurement Document Control, to read as follows: Appendix B to Part 50—Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants *Introduction.*
                             Every applicant for a construction permit is required by the provisions of § 50.34 or § 53.1309 of this chapter to include in its preliminary safety analysis report a description of the quality assurance program to be applied to the design, fabrication, construction, and testing of the structures, systems, and components of the facility. Every applicant for an operating license is required by the provisions of § 50.34 or § 53.1369 of this chapter to include, in its final safety analysis report, information pertaining to the managerial and administrative controls to be used to assure safe operation. Every applicant for a combined license is required by the provisions of §§ 52.79 or 53.1416 of this chapter to include in its final safety analysis report a description of the quality assurance applied to the design, and to be applied to the fabrication, construction, and testing of the structures, systems, and components of the facility and to the managerial and administrative controls to be used to assure safe operation. For applications submitted after September 27, 2007, every applicant for an early site permit is required by the provisions of §§ 52.17 or 53.1146 of this chapter to include in its site safety analysis report a description of the quality assurance program applied to site activities related to the design, fabrication, construction, and testing of the structures, systems, and components of a facility or facilities that may be constructed on the site. Every applicant for a design approval is required by the provisions of §§ 52.137 or 53.1209 of this chapter to include in its final safety analysis report a description of the quality assurance program applied to the design of the structures, systems, and components of the facility. Every applicant for a design certification is required by the provisions of §§ 52.47 or 53.1239 of this chapter to include in its final safety analysis report a description of the quality assurance program applied to the design of the structures, systems, and components of the facility. Every applicant for a manufacturing license is required by the provisions of §§ 52.157 or 53.1279 of this chapter to include in its final safety analysis report a description of the quality assurance program applied to the design, and to be applied to the manufacture of, the structures, systems, and components of the reactor. Nuclear power plants and fuel reprocessing plants include structures, systems, and components that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public. This appendix establishes quality assurance requirements for the design, manufacture, construction, and operation of those structures, systems, and components. The pertinent requirements of this appendix apply to all activities affecting the safety-related functions of those structures, systems, and components; these activities include designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, installing, inspecting, testing, operating, maintaining, repairing, refueling, and modifying. 
 III. Design Control Measures shall be established to assure that applicable regulatory requirements and the design bases, as defined in § 50.2 and as specified in the license application, or the functional design criteria, as defined in § 53.020 of this chapter and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. These measures shall include provisions to assure that appropriate quality 
                            ( print page 87040)
                            standards are specified and included in design documents and that deviations from such standards are controlled. Measures shall also be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems and components. 
 IV. Procurement Document Control Measures shall be established to assure that applicable regulatory requirements, design bases or functional design criteria, and other requirements which are necessary to assure adequate quality are suitably included or referenced in the documents for procurement of material, equipment, and services, whether purchased by the applicant or by its contractors or subcontractors. To the extent necessary, procurement documents shall require contractors or subcontractors to provide a quality assurance program consistent with the pertinent provisions of this appendix. 
 PART 51—ENVIRONMENTAL PROTECTION REGULATIONS FOR DOMESTIC LICENSING AND RELATED REGULATORY FUNCTIONS
 107.The authority citation for part 51 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 161, 193 (42 U.S.C. 2201, 2243); Energy Reorganization Act of 1974, secs. 201, 202 (42 U.S.C. 5841, 5842); National Environmental Policy Act of 1969 (42 U.S.C. 4332, 4334, 4335); Nuclear Waste Policy Act of 1982, secs. 144(f), 121, 135, 141, 148 (42 U.S.C. 10134(f), 10141, 10155, 10161, 10168); 44 U.S.C. 3504 note. 108.In § 51.20, revise paragraphs (b)(1) and (2) to read as follows: § 51.20
  
 (b) * * * (1) Issuance of a limited work authorization or a permit to construct a nuclear power reactor, testing facility, or fuel reprocessing plant under part 50 of this chapter, issuance of an early site permit under part 52 of this chapter, or issuance of a limited work authorization, construction permit, or early site permit under part 53 of this chapter. (2) Issuance or renewal of a full power or design capacity license to operate a nuclear power reactor, testing facility, or fuel reprocessing plant under parts 50 or 53 of this chapter, or a combined license under parts 52 or 53 of this chapter.
 109.In § 51.22, revise paragraphs (c)(3) introductory text, (c)(9) introductory text, (c)(12) introductory text, (c)(17), (c)(22) and (23) to read as follows: § 51.22
  
 (c) * * * (3) Amendments to parts 20, 30, 31, 32, 33, 34, 35, 37, 39, 40, 50, 51, 52, 53, 54, 60, 61, 63, 70, 71, 72, 73, 74, 81, and 100 of this chapter which relate to— 
 (9) Issuance of an amendment to a permit or license for a reactor under part 50, part 52, or part 53 of this chapter that changes a requirement or issuance of an exemption from a requirement, with respect to installation or use of a facility component located within the restricted area, as defined in part 20 of this chapter; or the issuance of an amendment to a permit or license for a reactor under part 50, part 52, or part 53 of this chapter that changes an inspection or a surveillance requirement; provided that: 
 (12) Issuance of an amendment to a license under parts 50, 52, 53, 60, 61, 63, 70, 72, or 75 of this chapter relating solely to safeguards matters (
                            *i.e.,*
                             protection against sabotage or loss or diversion of special nuclear material) or issuance of an approval of a safeguards plan submitted under parts 50, 52, 53, 70, 72, and 73 of this chapter, provided that the amendment or approval does not involve any significant construction impacts. These amendments and approvals are confined to— 
 (17) Issuance of an amendment to a permit or license under part 30, part 40, part 50, part 52, part 53, or part 70 of this chapter which deletes any limiting condition of operation or monitoring requirement based on or applicable to any matter subject to the provisions of the Federal Water Pollution Control Act. 
 (22) Issuance of a standard design approval under part 52 or part 53 of this chapter. (23) The Commission finding for a combined license under § 52.103(g) or § 53.1452(g) of this chapter.
 § 51.26
 110.In § 51.26, in paragraph (d) remove the phrase “under part 52” and add in its place the phrase “under 10 CFR parts 52 or 53,”. 111.In § 51.30, revise paragraph (a) introductory text and paragraphs (d) and (e) to read as follows: § 51.30
  (a) An environmental assessment for proposed actions, other than those for a standard design certification under 10 CFR parts 52 or 53, or a manufacturing license under 10 CFR parts 52 or 53, shall identify the proposed action and include: 
 (d) An environmental assessment for a standard design certification under subpart B of part 52 of this chapter, or under subpart H of part 53 of this chapter must identify the proposed action and will be limited to the consideration of the costs and benefits of severe accident mitigation design alternatives and the bases for not incorporating severe accident mitigation design alternatives in the design certification. An environmental assessment for an amendment to a design certification will be limited to the consideration of whether the design change which is the subject of the proposed amendment renders a severe accident mitigation design alternative previously rejected in the earlier environmental assessment to become cost beneficial, or results in the identification of new severe accident mitigation design alternatives, in which case the costs and benefits of new severe accident mitigation design alternatives and the bases for not incorporating new severe accident mitigation design alternatives in the design certification must be addressed. (e) An environmental assessment for a manufacturing license under subpart F of part 52 of this chapter or under subpart H of part 53 of this chapter must identify the proposed action and will be limited to the consideration of the costs and benefits of severe accident mitigation design alternatives and the bases for not incorporating severe accident mitigation design alternatives in the manufacturing license. An environmental assessment for an amendment to a manufacturing license will be limited to consideration of whether the design change which is the subject of the proposed amendment either renders a severe accident mitigation design alternative previously rejected in an environmental assessment to become cost beneficial, or results in the identification of new severe accident mitigation design alternatives, in which case the costs and benefits of new severe accident mitigation design alternatives and the bases for not incorporating new severe accident mitigation design alternatives in the manufacturing license must be addressed. In either case, the environmental assessment will not address the environmental impacts 
                            ( print page 87041)
                            associated with manufacturing the reactor under the manufacturing license.
 § 51.31
 112.In § 51.31, in paragraph (a) remove the phrase “under part 52” and add in its place the phrase “under parts 52 or 53”. § 51.32
 113.In § 51.32, in paragraphs (b)(1) and (3) remove the phrase “of part 52 of this chapter” and add in its place the phrase “of part 52 of this chapter or subpart H of part 53 of this chapter”. § 51.49
 114.In § 51.49, in paragraph (c) introductory text, remove the phrase “of part 52 of this chapter” and add in its place the phrase “of part 52 of this chapter or under subpart H of part 53 of this chapter”. § 51.50
 115.In § 51.50, wherever it appears, remove the phrase “in accordance with § 50.36b of this chapter” and add in its place the phrase “in accordance with §§ 50.36b or 53.1112 of this chapter”. § 51.53
 116.In § 51.53, in paragraph (d) remove the phrase “under § 50.82 of this chapter” and add in its place the phrase “under §§ 50.82 or 53.1080 of this chapter”. § 51.54
 117.In § 51.54, in paragraph (a), remove the phrase “of part 52 of this chapter” and add in its place the phrase “of part 52 of this chapter or under subpart H of part 53 of this chapter”. § 51.55
 118.In § 51.55, in paragraph (a) remove the phrase “of part 52 of this chapter” and add in its place the phrase “of part 52 of this chapter or under subpart H of part 53 of this chapter”. 119.In § 51.58, revise paragraph (b) to read as follows: § 51.58
  
 (b) Each applicant for a license to manufacture a nuclear power reactor, or for an amendment to a license to manufacture, seeking approval of the final design of the nuclear power reactor under subpart F of part 52 of this chapter or under subpart H of part 53 of this chapter, shall submit to the Commission an environmental report or any supplement to an environmental report in the manner specified in §§ 52.3 or 53.040 of this chapter. The applicant shall maintain the capability to generate additional copies of the environmental report or any supplement to the environmental report for subsequent distribution to parties and Boards in the NRC proceeding; Federal, State, and local officials; and any affected Indian Tribes, in accordance with written instructions issued by the Director, Office of Nuclear Reactor Regulation.
 120.In § 51.77, revise paragraph (a) introductory text to read as follows: § 51.77
  (a) In addition to the distribution authorized by § 51.74, a copy of a draft environmental statement for a licensing action for a production or utilization facility, except an action authorizing issuance, amendment, or renewal of a license to manufacture a nuclear power reactor pursuant to 10 CFR part 52, subpart F or 10 CFR part 53, subparts H or I will also be distributed to:
 § 51.92
 121.In § 51.92, in paragraph (b), wherever it may appear, remove the phrase “ 10 CFR part 52” and add in its place the phrase “10 CFR parts 52 or 53”. § 51.95
 122.In § 51.95, in paragraph (c) introductory text remove the phrase “under 10 CFR parts 52 or 54” and add in its place the phrase “under 10 CFR parts 52, 53, or 54”. 123.In § 51.101, revise paragraph (a)(2) to read as follows: § 51.101
  (a) * * * (2) Any action concerning the proposal taken by an applicant which would— (i) Have an adverse environmental impact, or (ii) Limit the choice of reasonable alternatives that may be grounds for denial of the license. In the case of an application covered by §§ 30.32(f), 40.31(f), 50.10(c), 53.1130, 70.21(f), or 72.16 and 72.34 of this chapter, the provisions of this paragraph will be applied in accordance with § 30.33(a)(5), 40.32(e), 50.10(c), 53.1130, 70.23(a)(7), or 72.40(b) of this chapter, as appropriate.
 § 51.103
 124.In § 51.103, in paragraph (a)(6) remove the phrase “under 10 CFR 50.10” and add in its place the phrase “under §§ 50.10 or 53.1130 of this chapter”. 125.In § 51.105, revise paragraph (c)(1) introductory text to read as follows: § 51.105
  
 (c)(1) In addition to complying with the applicable provisions of § 51.104, in any proceeding for the issuance of a construction permit for a nuclear power plant or an early site permit under parts 52 or 53 of this chapter, where the applicant requests a limited work authorization under §§ 50.10(d) or 53.1130 of this chapter, the presiding officer will—
 126.In § 51.107, revise paragraphs (a) introductory text, (b) introductory text, and (d)(1) introductory text to read as follows: § 51.107
  (a) In addition to complying with the applicable requirements of § 51.104, in a proceeding for the issuance of a combined license for a nuclear power reactor under parts 52 or 53 of this chapter, the presiding officer will: 
 (b) If a combined license application references an early site permit, then the presiding officer in the combined license hearing must not admit any contention proffered by any party on environmental issues that have been accorded finality under §§ 52.39 or 53.1188 of this chapter, unless the contention: 
 (d)(1) In any proceeding for the issuance of a combined license where the applicant requests a limited work authorization under §§ 50.10(d) or § 53.1130(a) of this chapter, the presiding officer, in addition to complying with any applicable provision of § 51.104, will:
 127.Revise § 51.108 to read as follows: § 51.108
  In any public hearing requested under §§ 52.103(b) or 53.1452(b) of this chapter, the Commission will not admit any contentions on environmental issues, the adequacy of the environmental impact statement for the combined license issued under subpart C of part 52 of this chapter or under 
                            ( print page 87042)
                            subpart H of part 53 of this chapter, or the adequacy of any other environmental impact statement or environmental assessment referenced in the combined license application. The Commission will not make any environmental findings in connection with the finding under § 52.103(g) or § 53.1452(g) of this chapter.
 128.Add part 53, consisting of §§ 53.000 through 53.9010, to read as follows: PART 53—RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK FOR COMMERCIAL NUCLEAR PLANTS 53.000
                            Purpose.
                            Subpart A—General Provisions53.015
                                Scope.
                                53.020
                                Definitions.
                                53.030
                                Reserved.
                                53.040
                                Written communications.
                                53.050
                                Deliberate misconduct.
                                53.060
                                Employee protection.
                                53.070
                                Completeness and accuracy of information.
                                53.080
                                Specific exemptions.
                                53.090
                                Standards for review.
                                53.100
                                Jurisdictional limits.
                                53.110
                                Attacks and destructive acts.
                                53.115
                                Rights related to special nuclear material.
                                53.117
                                License suspension and rights of recapture.
                                53.120
                                Information collection requirements: OMB approval.
                            Subpart B—Technology-Inclusive Safety Requirements53.210
                                Safety criteria for design-basis accidents.
                                53.220
                                Safety criteria for licensing-basis events other than design-basis accidents.
                                53.230
                                Safety functions.
                                53.240
                                Licensing-basis events.
                                53.250
                                Defense in depth.
                                53.260
                                Normal operations.
                                53.270
                                Protection of plant workers.
                            Subpart C—Design and Analysis Requirements53.400
                                Design features for licensing-basis events.
                                53.410
                                Functional design criteria for design-basis accidents.
                                53.415
                                Protection against external hazards.
                                53.420
                                Functional design criteria for licensing-basis events other than design-basis accidents.
                                53.425
                                Design features and functional design criteria for normal operations.
                                53.430
                                Design features and functional design criteria for protection of plant workers.
                                53.440
                                Design requirements.
                                53.450
                                Analysis requirements.
                                53.460
                                Safety categorization and treatments.
                                53.470
                                Maintaining analytical safety margins used to justify operational flexibilities.
                                53.480
                                Earthquake engineering.
                            Subpart D—Siting Requirements53.500
                                General siting and siting assessment.
                                53.510
                                External hazards.
                                53.520
                                Site characteristics.
                                53.530
                                Population-related considerations.
                                53.540
                                Siting interfaces.
                            Subpart E—Construction and Manufacturing Requirements53.600
                                Construction and manufacturing—scope and purpose.
                                53.605
                                Reporting of defects and noncompliance.
                                53.610
                                Construction.
                                53.620
                                Manufacturing.
                            Subpart F—Requirements for Operation53.700
                                Operational objectives.
                                53.710
                                Maintaining capabilities and availability of structures, systems, and components.
                                53.715
                                Maintenance, repair, and inspection programs.
                                53.720
                                Response to seismic events.
                                53.725
                                General staffing, training, personnel qualifications, and human factors requirements.
                                53.726
                                Communications.
                                53.728
                                Completeness and accuracy of information.
                                53.730
                                Defining, fulfilling, and maintaining the role of personnel in ensuring safe operations.
                                53.735
                                General exemptions.
                                53.740
                                Facility licensee requirements—General.
                                53.745
                                Operator license requirements.
                                53.760
                                Operator licensing.
                                53.765
                                Medical requirements.
                                53.770
                                Incapacitation because of disability or illness.
                                53.775
                                Applications for operators and senior operators.
                                53.780
                                Training, examination, and proficiency program.
                                53.785
                                Conditions of operator and senior operator licenses.
                                53.790
                                Issuance, modification, and revocation of operator and senior operator licenses.
                                53.795
                                Expiration and renewal of operator and senior operator licenses.
                                53.800
                                Facility licensees for self-reliant-mitigation facilities.
                                53.805
                                Facility licensee requirements related to generally licensed reactor operators.
                                53.810
                                Generally licensed reactor operators.
                                53.815
                                Generally licensed reactor operator training, examination, and proficiency programs.
                                53.820
                                Cessation of individual applicability.
                                53.830
                                Training and qualification of commercial nuclear plant personnel.
                                53.845
                                Programs.
                                53.850
                                Radiation protection.
                                53.855
                                Emergency preparedness.
                                53.860
                                Security programs.
                                53.865
                                Quality assurance.
                                53.870
                                Integrity assessment programs.
                                53.875
                                Fire protection.
                                53.880
                                Inservice inspection and inservice testing.
                                53.910
                                Procedures and guidelines.
                            Subpart G—Decommissioning Requirements53.1000
                                Scope and purpose.
                                53.1010
                                Financial assurance for decommissioning.
                                53.1020
                                Cost estimates for decommissioning.
                                53.1030
                                Annual adjustments to cost estimates for decommissioning.
                                53.1040
                                Methods for providing financial assurance for decommissioning.
                                53.1045
                                Limitations on the use of decommissioning trust funds.
                                53.1050
                                NRC oversight.
                                53.1060
                                Reporting and recordkeeping requirements.
                                53.1070
                                Termination of license.
                                53.1075
                                Program requirements during decommissioning.
                                53.1080
                                Release of part of a commercial nuclear plant or site for unrestricted use.
                            Subpart H—Licenses, Certifications, and Approvals53.1100
                                Filing of application for licenses, certifications, or approvals; oath or affirmation.
                                53.1101
                                Requirement for license.
                                53.1103
                                Combining applications and licenses.
                                53.1106
                                Elimination of repetition.
                                53.1109
                                Contents of applications; general information.
                                53.1112
                                Environmental conditions.
                                53.1115
                                Agreement limiting access to classified information.
                                53.1118
                                Ineligibility of certain applicants.
                                53.1120
                                Exceptions and exemptions from licensing requirements.
                                53.1121
                                Public inspection of applications.
                                53.1124
                                Relationship between sections.
                                53.1130
                                Limited work authorizations.
                                53.1140
                                Early site permits.
                                53.1143
                                Filing of applications.
                                53.1144
                                Contents of applications for early site permits; general information.
                                53.1146
                                Contents of applications for early site permits; technical information.
                                53.1149
                                Review of applications.
                                53.1155
                                Referral to the Advisory Committee on Reactor Safeguards.
                                53.1158
                                Issuance of early site permit.
                                53.1161
                                Extent of activities permitted.
                                53.1164
                                Duration of permit.
                                53.1167
                                Limited work authorization after issuance of early site permit.
                                53.1170
                                Transfer of early site permit.
                                53.1173
                                Application for renewal.
                                53.1176
                                Criteria for renewal.
                                53.1179
                                Duration of renewal.
                                53.1182
                                Use of site for other purposes.
                                53.1188
                                Finality of early site permit determinations.
                                53.1200
                                Standard design approvals.
                                53.1203
                                Filing of applications.
                                53.1206
                                Contents of applications for standard design approvals; general information.
                                53.1209
                                Contents of applications for standard design approvals; technical information.
                                53.1210
                                Contents of applications for standard design approvals; other application content.
                                53.1212
                                Standards for review of applications.
                                    ( print page 87043)
                                53.1215
                                Referral to the Advisory Committee on Reactor Safeguards.
                                53.1218
                                Staff approval of design.
                                53.1221
                                Finality of standard design approvals; information requests.
                                53.1230
                                Standard design certifications.
                                53.1233
                                Filing of applications.
                                53.1236
                                Contents of applications for standard design certifications; general information.
                                53.1239
                                Contents of applications for standard design certifications; technical information.
                                53.1241
                                Contents of applications for standard design certifications; other application content.
                                53.1242
                                Review of applications.
                                53.1245
                                Referral to the Advisory Committee on Reactor Safeguards.
                                53.1248
                                Issuance of standard design certification.
                                53.1251
                                Duration of certification.
                                53.1254
                                Application for renewal.
                                53.1257
                                Criteria for renewal.
                                53.1260
                                Duration of renewal.
                                53.1263
                                Finality of standard design certifications.
                                53.1270
                                Manufacturing licenses.
                                53.1273
                                Filing of applications.
                                53.1276
                                Contents of applications for manufacturing licenses; general information.
                                53.1279
                                Contents of applications for manufacturing licenses; technical information.
                                53.1282
                                Contents of applications for manufacturing licenses; other application content.
                                53.1285
                                Review of applications.
                                53.1286
                                Referral to the Advisory Committee on Reactor Safeguards.
                                53.1287
                                Issuance of manufacturing licenses.
                                53.1288
                                Finality of manufacturing licenses.
                                53.1291
                                Duration of manufacturing licenses.
                                53.1293
                                Transfer of manufacturing licenses.
                                53.1295
                                Renewal of manufacturing licenses.
                                53.1300
                                Construction permits.
                                53.1306
                                Contents of applications for construction permits; general information.
                                53.1309
                                Contents of applications for construction permits; technical information.
                                53.1312
                                Contents of applications for construction permits; other application content.
                                53.1315
                                Review of applications.
                                53.1318
                                Finality of referenced NRC approvals, permits, and certifications.
                                53.1324
                                Referral to the Advisory Committee on Reactor Safeguards.
                                53.1327
                                Authorization to conduct limited work authorization activities.
                                53.1330
                                Exemptions, departures, and variances.
                                53.1333
                                Issuance of construction permits.
                                53.1336
                                Finality of construction permits.
                                53.1342
                                Duration of construction permits.
                                53.1345
                                Transfer of construction permits.
                                53.1348
                                Termination of construction permits.
                                53.1360
                                Operating licenses.
                                53.1366
                                Contents of applications for operating licenses; general information.
                                53.1369
                                Contents of applications for operating licenses; technical information.
                                53.1372
                                Contents of applications for operating licenses; other application content.
                                53.1375
                                Review of applications.
                                53.1381
                                Referral to the Advisory Committee on Reactor Safeguards.
                                53.1384
                                Exemptions, departures, and variances.
                                53.1387
                                Issuance of operating licenses.
                                53.1390
                                Backfitting of operating licenses.
                                53.1396
                                Duration of operating licenses.
                                53.1399
                                Transfer of an operating license.
                                53.1402
                                Application for renewal.
                                53.1405
                                Continuation of an operating license.
                                53.1410
                                Combined licenses.
                                53.1413
                                Contents of applications for combined licenses; general information.
                                53.1416
                                Contents of applications for combined licenses; technical information.
                                53.1419
                                Contents of applications for combined licenses; other application content.
                                53.1422
                                Review of applications.
                                53.1425
                                Finality of referenced NRC approvals.
                                53.1431
                                Referral to the Advisory Committee on Reactor Safeguards.
                                53.1434
                                Authorization to conduct limited work authorization activities.
                                53.1437
                                Exemptions, departures, and variances.
                                53.1440
                                Issuance of combined licenses.
                                53.1443
                                Finality of combined licenses.
                                53.1449
                                Inspection during construction.
                                53.1452
                                Operation under a combined license.
                                53.1455
                                Duration of combined license.
                                53.1456
                                Transfer of a combined license.
                                53.1458
                                Application for renewal.
                                53.1461
                                Continuation of combined license.
                                53.1470
                                Standardization of commercial nuclear plant designs: licenses to construct and operate nuclear power reactors of identical design at multiple sites.
                            Subpart I—Maintaining and Revising Licensing-Basis Information53.1500
                                Licensing-basis information.
                                53.1502
                                Specific terms and conditions of licenses.
                                53.1505
                                Changes to licensing-basis information requiring prior NRC approval.
                                53.1510
                                Application for amendment of license.
                                53.1515
                                Public notices; State consultation.
                                53.1520
                                Issuance of amendment.
                                53.1525
                                Revising certification information within a design certification rule.
                                53.1530
                                Revising design information within a manufacturing license.
                                53.1535
                                Amendments during construction.
                                53.1540
                                Updating licensing-basis information and determining the need for NRC approval.
                                53.1545
                                Updating Final Safety Analysis Reports.
                                53.1550
                                Evaluating changes to facility as described in Final Safety Analysis Reports.
                                53.1560
                                Updating program documents included in licensing-basis information.
                                53.1565
                                Evaluating changes to programs included in licensing-basis information.
                                53.1570
                                Transfer of licenses.
                                53.1575
                                Termination of licenses.
                                53.1580
                                Information requests.
                                53.1585
                                Revocation, suspension, modification of licenses and approvals for cause.
                                53.1590
                                Backfitting.
                                53.1595
                                Renewal.
                            Subpart J—Reporting and Other Administrative Requirements53.1600
                                General information.
                                53.1610
                                Unfettered access for inspections.
                                53.1620
                                Maintenance of records, making of reports.
                                53.1630
                                Immediate notification requirements for operating commercial nuclear plants.
                                53.1640
                                Licensee event report system.
                                53.1645
                                Reports of radiation exposure to members of the public.
                                53.1650
                                Facility information and verification.
                                53.1660
                                Financial requirements.
                                53.1670
                                Financial qualifications.
                                53.1680
                                Annual financial reports.
                                53.1690
                                Licensee's change of status; financial qualifications.
                                53.1700
                                Creditor regulations.
                                53.1710
                                Financial protection.
                                53.1720
                                Insurance required to stabilize and decontaminate plant following an accident.
                                53.1730
                                Financial protection requirements.
                            Subpart M—Enforcement53.9000
                                Violations.
                                53.9010
                                Criminal penalties.
 Authority:
                            Atomic Energy Act of 1954, secs. 11, 101, 103, 108, 122, 147, 161, 181, 182, 183, 184, 185, 186, 187, 189, 223, 234 (42 U.S.C. 2014, 2131, 2132, 2133, 2134, 2135, 2138, 2152, 2167, 2169, 2201, 2231, 2232, 2233, 2234, 2235, 2236, 2237, 2239, 2273, 2282); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); Nuclear Waste Policy Act of 1982, sec. 306 (42 U.S.C. 10226); National Environmental Policy Act of 1969 (42 U.S.C. 4332); 44 U.S.C. 3504 note; Sec. 109, Pub. L. 96-295, 94 Stat. 783; Pub. L. 115-439, 132 Stat. 5571. § 53.000
  This part provides an optional technology-inclusive, performance-based framework for the issuance, amendment, renewal, and termination of licenses, permits, certifications, and approvals for commercial nuclear plants licensed under section 103 of the Atomic Energy Act of 1954, as amended (the Act)(68 Stat. 919), and Title II of the Energy Reorganization Act of 1974, as amended (88 Stat. 1242). Also, this part gives notice to all persons who knowingly provide to any holder of or applicant for an approval, certification, permit, or license, or to a contractor, subcontractor, or consultant of any of 
                                ( print page 87044)
                                them, components, equipment, materials, or other goods or services that relate to the activities of a holder of or applicant for an approval, certification, permit, or license, subject to this part, that they may be individually subject to U.S. Nuclear Regulatory Commission enforcement action for violation of the provisions in § 53.050.
 Subpart A—General Provisions § 53.015
  Subpart A provides general provisions applicable to all applicants and licensees subject to the rules of this part.
 § 53.020
  For the purpose of this part: *Anticipated event sequence*
                                     means event sequences expected to occur one or more times during the life of a commercial nuclear plant. Anticipated event sequences take into account the expected response of all structures, systems, and components (SSCs) within the plant, regardless of safety classification. *Applicant*
                                     means a person applying for a license, permit, or other form of Commission permission or approval under this part. *Certified fuel handler*
                                     means, for a commercial nuclear plant, either— (1) A non-licensed operator who has qualified in accordance with a fuel handler training program approved by the Commission; or (2) A non-licensed operator who demonstrates compliance with the following criteria: (i) Has qualified in accordance with a fuel handler training program that demonstrates compliance with the same requirements as training programs for non-licensed operators required by § 53.830, and (ii) Is responsible for decisions on— (A) Safe conduct of decommissioning activities, (B) Safe handling and storage of spent fuel, and (C) Appropriate response to plant emergencies. *Combined license (COL)*
                                     means a combined construction permit (CP) and operating license (OL) with conditions for a commercial nuclear plant issued under this part. *Commercial nuclear plant*
                                     means a facility consisting of one or more commercial nuclear reactors and associated co-located support facilities, including the collection of buildings, radionuclide sources, and SSCs for which a license, certification, or approval is being sought under this part, that is or will be used for producing power for commercial electric power or other commercial purposes. For the purposes of requirements in this part that reference requirements in part 50 of this chapter, a commercial nuclear plant is equivalent to a nuclear power plant. *Commercial nuclear reactor*
                                     means an apparatus, other than an atomic weapon, designed or used to sustain nuclear fission. For the purposes of requirements in this part that reference requirements in 10 CFR part 50, a commercial nuclear reactor is equivalent to a nuclear reactor as defined in 10 CFR 50.2. *Commission*
                                     means the U.S. Nuclear Regulatory Commission (NRC) or its duly authorized representatives. *Consensus code or standard*
                                     means any technical standard that is— (1) Developed or adopted by a voluntary consensus standard body under procedures that assure that persons having interests within the scope of the standard that are affected by the provisions of the standard have reached substantial agreement on its adoption; (2) Formulated in a manner that afforded an opportunity for diverse views to be considered; and (3) Designated by the standards body as a consensus code or standard. *Construction*
                                     means the activities in paragraph (1) below and does not mean the activities in paragraph (2) below. (1) Activities constituting construction are those activities credited or relied upon for demonstrating compliance with the safety criteria defined in subpart B of this part which are conducted on-site to build the commercial nuclear plant, including the driving of piles; subsurface preparation; placement of backfill, concrete, or permanent retaining walls within an excavation; installation of foundations; or in-place assembly, erection, fabrication, or testing, which are for— (i) Safety-related (SR) and non-safety-related but safety-significant (NSRSS) SSCs of a facility; (ii) SSCs necessary to comply with 10 CFR part 73; or (iii) Onsite emergency facilities necessary to comply with § 53.855. (2) Construction does not include— (i) Changes for temporary use of the land for public recreational purposes; (ii) Site exploration, including necessary borings to determine foundation conditions or other preconstruction monitoring to establish background information related to the suitability of the site, the environmental impacts of construction or operation, or the protection of environmental values; (iii) Preparation of a site for construction of a facility, including clearing of the site, grading, installation of drainage, erosion, and other environmental mitigation measures, and construction of temporary roads and borrow areas; (iv) Erection of fences and other access control measures; (v) Excavation; (vi) Erection of support buildings (such as construction equipment storage sheds, warehouse and shop facilities, utilities, concrete mixing plants, docking and unloading facilities, and office buildings) for use in connection with the construction of the facility; (vii) Building of service facilities (such as paved roads, parking lots, railroad spurs, exterior utility and lighting systems, potable water systems, sanitary sewage treatment facilities, and transmission lines); (viii) Procurement or fabrication of components or portions of the proposed facility occurring at locations other than the final, in-place location at the facility; or (ix) Manufacture of a nuclear power reactor under a manufacturing license (ML) under subpart H of this part to be installed at the proposed site and to be part of the proposed facility. *Custom combined license (custom COL)*
                                     means a COL that does not reference a standard design approval or design certification. *Decommission or decommissioning*
                                     means to remove a plant or site safely from service and reduce residual radioactivity to a level that permits— (1) Release of the property for unrestricted use and termination of the license; or (2) Release of the property under restricted conditions and termination of the license. *Defense in depth*
                                     means inclusion of two or more independent and redundant layers of defense in the design of a facility and its operating procedures to compensate for uncertainties such that no single layer of defense, no matter how robust, is exclusively relied upon. Defense in depth includes, but is not limited to, the use of access controls, physical barriers, redundant and diverse safety functions, and emergency response measures. *Design-basis accidents (DBAs)*
                                     means postulated event sequences that are used to set functional design criteria and performance objectives for the design of SR SSCs through deterministic analyses. Design-basis accidents are a type of licensing-basis event and are based on the capabilities and reliabilities of SR SSCs needed to mitigate and prevent event sequences, respectively. *Design-basis external hazard level*
                                     means the level of severity or intensity 
                                    ( print page 87045)
                                    of an external hazard for which the SR SSCs are protected against or designed to withstand without losing their capability to perform their safety functions. *Design features*
                                     means the active and passive SSCs and the inherent characteristics of those SSCs that contribute to limiting the total effective dose equivalent to individual members of the public during normal operations and prevent or mitigate the consequences of event sequences. *Electric utility*
                                     means any entity that generates or distributes electricity and that recovers the cost of this electricity, either directly or indirectly, through rates established by the entity itself or by a separate regulatory authority. Investor-owned utilities, including generation or distribution subsidiaries, public utility districts, municipalities, rural electric cooperatives, and State and Federal agencies, including associations of any of the foregoing, are included within the meaning of “electric utility.” *Event sequence*
                                     means a postulated initiating event defined for a set of initial plant conditions followed by system, safety function, and operator successes or failures, and terminating in a specified end state depending on the system, safety function, and operator successes and failures (
                                    *e.g.,*
                                     prevention of release of radioactive material or release in one of the reactor-specific release categories). An event sequence may include many unique variations of events that are similar in terms of results or end states. *Exclusion area*
                                     means that area surrounding the reactor, in which the reactor licensee has the authority to determine all activities including exclusion or removal of personnel and property from the area. This area may be traversed by a highway, railroad, or waterway, provided these are not so close to the facility as to interfere with normal operations of the facility and provided appropriate and effective arrangements are made to control traffic on the highway, railroad, or waterway, in case of emergency, to protect the public health and safety. Residence within the exclusion area must normally be prohibited. In any event, residents must be subject to ready removal in case of necessity. Activities unrelated to operation of the reactor may be permitted in an exclusion area under appropriate limitations, provided that no significant hazards to the public health and safety will result. *Fission product release*
                                     means the amount and composition of radioactive material released to the environment, after accounting for any retention of radionuclides provided by reactor design features. *Fuel*
                                     means special nuclear material (SNM) or source material, discrete elements that physically contain SNM or source material, and homogeneous mixtures that contain SNM or source material, intended to or used to create power in a commercial nuclear plant. *Functional design criteria*
                                     means metrics for the performance of SSCs. For SR SSCs, these criteria define performance metrics necessary to demonstrate compliance with the safety criteria in § 53.210. For NSRSS SSCs, these criteria define performance metrics necessary to demonstrate compliance with the safety criteria in § 53.220. *License,*
                                     when used in the context of a facility, means a limited work authorization, CP, OL, early site permit, COL, or ML under this part, or a renewed license issued by the Commission under this part. When used in the context of a license authorizing an individual to manipulate the controls of a facility, 
                                    *license*
                                     means a license issued by the Commission to perform the function of an operator, senior operator, or generally licensed reactor operator as defined in this part. *Licensee*
                                     means a person who is authorized to conduct activities under a license issued under this part by the Commission. *Licensing-basis events*
                                     means a collection of event sequences considered in the design and licensing of the commercial nuclear plant. Licensing-basis events are unplanned events and include anticipated event sequences, unlikely event sequences, very unlikely event sequences, and DBAs. *Licensing-basis information*
                                     means the information contained in regulations, orders, licenses, certifications, or approvals issued by the NRC for a commercial nuclear plant licensed under this part and that information submitted to the NRC by an applicant or licensee in a Safety Analysis Report, program description, or other licensing-related document required under this part. *Low-population zone*
                                     means the area immediately surrounding the exclusion area which contains residents, the total number and density of which are such that there is a reasonable probability that appropriate protective measures could be taken on their behalf in the event of a serious accident. A permissible population density or total population within this zone is not included in this definition because the situation may vary from case to case. Whether a specific number of people can, for example, be evacuated from a specific area or instructed to take shelter on a timely basis, will depend on many factors such as location, number and size of highways, scope and extent of advance planning, and actual distribution of residents within the area. *Major decommissioning activity*
                                     means, for a commercial nuclear plant, any activity that results in permanent removal of major radioactive components, permanently modifies the structure of the containment, if applicable, or results in dismantling components for shipment containing greater than class C waste in accordance with 10 CFR 61.55. *Major feature of the emergency*
                                     plans means an aspect of those plans necessary to: (1) Address in whole or part either one or more of the 16 standards in 10 CFR 50.47(b) or the requirements of 10 CFR 50.160(b), as applicable; or (2) Describe the emergency planning zones as required in § 53.1109(g). *Manufactured reactor*
                                     means the essential portions of a nuclear reactor that are manufactured under an ML and subsequently transported and incorporated into a commercial nuclear plant under a COL. *Manufacturing license*
                                     means a license issued under this part that authorizes the manufacture of manufactured reactors but not its construction, installation, or operation. *Non-Safety-Related but Safety-Significant (NSRSS) SSCs*
                                     means those SSCs which are not SR but are relied on to achieve adequate defense in depth or perform risk-significant functions and warrant special treatment. *Non-Safety-Significant SSCs*
                                     means those SSCs that are not SR or NSRSS, are not relied on to achieve adequate defense in depth or to perform risk-significant functions, and do not warrant special treatment. *Person*
                                     means— (1) any individual, corporation, partnership, firm, association, trust, estate, public or private institution, group, government agency other than the Commission or the Department, except that the Department shall be considered a person to the extent that its facilities are subject to the licensing and related regulatory authority of the Commission pursuant to section 202 of the ERA, any State or any political subdivision of, or any political entity within a State, any foreign government or nation or any political subdivision of any such government or nation, or other entity; and (2) any legal successor, representative, agent, or agency of the foregoing.
                                    ( print page 87046) *Population center distance*
                                     means the distance from the reactor to the nearest boundary of a densely populated center containing more than about 25,000 residents. *Probabilistic risk assessment*
                                     means a quantitative assessment of the risk associated with plant operation and maintenance that is measured in terms of event sequence occurrence frequencies and consequences. *Programmatic controls*
                                     means administrative procedures that govern human action in implementing programs and operating, monitoring, and maintaining SSCs and equipment of a commercial nuclear plant. Programmatic controls considered to be licensing basis information are specified in an application for a requested activity of the Commission. *Quality assurance (QA)*
                                     means all those planned and systematic actions necessary to ensure that a structure, system, or component will perform satisfactorily in service. Quality assurance includes quality control, which comprises those QA actions related to the physical characteristics of a material, structure, component, or system which provide a means to control the quality of the material, structure, component, or system to predetermined requirements. *Safety criteria*
                                     means performance-based metrics that establish a level of safety provided in requirements in §§ 53.210 and 53.220. *Safety-related structures, systems, or components*
                                     means those SSCs that are relied upon to demonstrate compliance with the safety criteria in § 53.210 and warrant special treatment. *Small modular reactor*
                                     means a power reactor, which may be of modular design as defined in 10 CFR 52.1, licensed under this part to produce heat energy up to 1,000 megawatts thermal per module. *Site characteristics*
                                     means the actual physical, environmental, and demographic features of a site. Site characteristics are specified in an early site permit or in a Preliminary or Final Safety Analysis Report for a limited work authorization, CP, or COL, as applicable. *Site parameters*
                                     are the postulated physical, environmental, and demographic features of an assumed site. Site parameters are specified in a standard design approval, standard design certification, or ML. *Source material*
                                     means source material as defined in subsection 11z. of the Atomic Energy Act of 1954, as amended, (the Act) and in the regulations contained in part 40 of this chapter. *Special nuclear material (SNM)*
                                     means: (1) Plutonium, uranium-233, uranium enriched in the isotope-233 or in the isotope-235, and any other material which the Commission, pursuant to the provisions of section 51 of the Act, determines to be SNM, but does not include source material; or (2) Any material artificially enriched by any of the foregoing, but does not include source material. *Special treatment*
                                     means those requirements, such as QA and programmatic controls, that ensure that SR and NSRSS SSCs will provide defense in depth or perform risk-significant functions. The requirements also ensure that the SSCs will perform under the service conditions and with the reliability assumed in the analysis performed under § 53.450 to demonstrate compliance with the safety criteria in §§ 53.210 and 53.220. *Standard design*
                                     means a design which is sufficiently detailed and complete to support certification or approval in accordance with subpart H of this part, and which is usable under of this part for a multiple number of units or at a multiple number of sites without reopening or repeating the review. *Standard design approval or design approval*
                                     means an NRC staff approval, issued under subpart H of this part, of a final standard design for a commercial nuclear plant. The approval may be for either the final design for the entire reactor facility or the final design of major portions thereof. *Standard design certification or design certification*
                                     means a Commission approval, issued under subpart H of this part, of a final standard design for a nuclear power facility. This design may be referred to as a certified standard design. *Total effective dose equivalent*
                                     means the sum of the effective dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures). *Utilization facility*
                                     means any commercial nuclear reactor other than one designed or used primarily for the formation of plutonium or uranium-233. *Unlikely event sequences*
                                     means event sequences that are not expected to occur in the life of a commercial nuclear plant and are less likely than anticipated event sequences, but are infrequent rather than rare. Unlikely event sequences take into account the expected response of all SSCs within the plant regardless of safety classification. *Very unlikely event sequences*
                                     means event sequences that are not expected to occur in the life of a commercial nuclear plant, are less likely than an unlikely event sequence, and are rare. Very unlikely event sequences take into account the expected response of all SSCs within the plant regardless of safety classification.
 § 53.030
 § 53.040
  (a) 
                                    *General requirements.*
                                     All correspondence, reports, applications, and other written communications from the applicant or licensee to the NRC concerning the regulations in this part or individual license conditions must be sent either by mail addressed: ATTN: Document Control Desk, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; by hand delivery to the NRC's offices at 11555 Rockville Pike, Rockville, Maryland, between the hours of 8:15 a.m. and 4 p.m. eastern time; or, where practicable, by electronic submission, for example, via Electronic Information Exchange, email, or CD-ROM. Electronic submissions must be made in a manner that enables the NRC to receive, read, authenticate, distribute, and archive the submission, and process and retrieve it a single page at a time. Detailed guidance on making electronic submissions can be obtained by visiting the NRC's website at 
                                    *http://www.nrc.gov/​site-help/​e-submittals.html;*
                                     by email to 
                                    *MSHD.Resource@nrc.gov;*
                                     or by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. The guidance discusses, among other topics, the formats the NRC can accept, the use of electronic signatures, and the treatment of nonpublic information. If the communication is on paper, the signed original must be sent. If a submission due date falls on a Saturday, Sunday, or Federal holiday, the next Federal working day becomes the official due date. (b) 
                                    *Distribution requirements.*
                                     Copies of all correspondence, reports, and other written communications concerning the regulations in this part or individual license conditions, or the terms and conditions of an early site permit or standard design approval, must be submitted to the persons listed below (addresses for the NRC Regional Offices are listed in appendix D to 10 CFR part 20). (1) 
                                    *Applications for amendment of permits and licenses, reports, and other communications.*
                                     All written communications (including responses to generic letters, bulletins, information notices, regulatory information summaries, inspection reports, and 
                                    ( print page 87047)
                                    miscellaneous requests for additional information) that are required of holders of licenses, permits, and design approvals issued pursuant to this part, must be submitted as follows, except as otherwise specified in paragraphs (b)(2) through (7) of this section: to the NRC's Document Control Desk (if on paper, the signed original), with a copy to the appropriate Regional Office, and a copy to the appropriate NRC Resident Inspector if one has been assigned to the site of the facility or the place of manufacture of a reactor licensed under this part. (2) 
                                    *Applications for permits and licenses, and amendments to applications.*
                                     Applications for licenses, permits, and design approvals and amendments to any of these types of applications must be submitted to the NRC's Document Control Desk, with a copy to the appropriate Regional Office, and a copy to the appropriate NRC Resident Inspector if one has been assigned to the facility or the place of manufacture of a reactor licensed under this part, except as otherwise specified in paragraphs (b)(3) through (9) of this section. If the application or amendment is on paper, the submission to the Document Control Desk must be the signed original. (3) 
                                    *Acceptance review application.*
                                     Written communications required for an application for determination of suitability for docketing must be submitted to the NRC's Document Control Desk, with a copy to the appropriate Regional Office. If the communication is on paper, the submission to the Document Control Desk must be the signed original. (4) 
                                    *Security plan and related submissions.*
                                     Written communications, as defined in paragraphs (b)(4)(i) through (v) of this section, must be submitted to the NRC's Document Control Desk, with a copy to the appropriate Regional Office. If the communication is on paper, the submission to the Document Control Desk must be the signed original. Submissions should include the following as appropriate: (i) Physical security plan; (ii) Safeguards contingency plan; (iii) Cybersecurity plan; (iv) Change to security plan, guard training and qualification plan, safeguards contingency plan, or cybersecurity plan made without prior Commission approval under § 53.1565; and (v) Application for amendment of physical security plan, guard training and qualification plan, safeguards contingency plan, or cybersecurity plan under § 53.1510. (5) 
                                    *Emergency plan and related submissions.*
                                     Written communications as defined in paragraphs (b)(5)(i) through (iii) of this section must be submitted to the NRC's Document Control Desk, with a copy to the appropriate Regional Office, and a copy to the appropriate NRC Resident Inspector if one has been assigned to the site of the facility. If the communication is on paper, the submission to the Document Control Desk must be the signed original. Submissions should include the following as appropriate: (i) Emergency plan; (ii) Change to an emergency plan under § 53.1565; and (iii) Emergency implementing procedures under § 53.855. (6) 
                                    *Updated Final Safety Analysis Report.*
                                     An Updated Final Safety Analysis Report or replacement pages under § 53.1545 must be submitted to the NRC's Document Control Desk, with a copy to the appropriate Regional Office, and a copy to the appropriate NRC Resident Inspector if one has been assigned to the site of the facility or the place of manufacture of a reactor licensed under this part. Paper copy submissions may be made using replacement pages; however, if a licensee chooses to use electronic submission, all subsequent updates or submissions must be performed electronically on a total replacement basis. If the communication is on paper, the submission to the Document Control Desk must be the signed original. If the communications are submitted electronically, see Guidance for Electronic Submissions to the Commission. (7) 
                                    *Quality assurance related submissions.*
                                     (i) A change to the Safety Analysis Report QA program description under § 53.1565, or a change to a licensee's NRC-accepted QA topical report under § 53.1565, must be submitted to the NRC's Document Control Desk, with a copy to the appropriate Regional Office, and a copy to the appropriate NRC Resident Inspector if one has been assigned to the site of the facility or the place of manufacture of a reactor licensed under this part. If the communication is on paper, the submission to the Document Control Desk must be the signed original. (ii) A change to an NRC-accepted QA topical report from non-licensees (
                                    *i.e.,*
                                     architect/engineers, nuclear steam supply system suppliers, fuel suppliers, constructors, etc.) must be submitted to the NRC's Document Control Desk. If the communication is on paper, the signed original must be sent. (8) 
                                    *Certification of permanent cessation of operations.*
                                     The licensee's certification of permanent cessation of operations, under subpart G of this part, must state the date on which operations have ceased or will cease, and must be submitted to the NRC's Document Control Desk. This submission must be under oath or affirmation. (9) 
                                    *Certification of permanent fuel removal.*
                                     The licensee's certification of permanent fuel removal, under subpart G of this part, must state the date on which the fuel was removed from the reactor vessel and the disposition of the fuel, and must be submitted to the NRC's Document Control Desk. This submission must be under oath or affirmation. (c) 
                                    *Form of communications.*
                                     All paper copies submitted to demonstrate compliance with the requirements set forth in paragraph (b) of this section must be typewritten, printed, or otherwise reproduced in permanent form on unglazed paper. Exceptions to these requirements imposed on paper submissions may be granted for the submission of micrographic, photographic, or similar forms. (d) 
                                    *Regulation governing submission.*
                                     Licensees, applicants, and holders of standard design approvals submitting correspondence, reports, and other written communications under the regulations of this part are requested but not required to cite whenever practical, in the upper right corner of the first page of the submission, the specific regulation or other basis requiring submission.
 § 53.050
  (a) Any licensee or applicant for a license; holder of or applicant for a standard design approval; applicant for a standard design certification; employee of a licensee, holder of a standard design approval, or applicant for a license, standard design approval, or standard design certification; or any contractor (including a supplier or consultant), subcontractor, employee of a contractor or subcontractor of any licensee or applicant for a license, holder of or applicant for a standard design approval, or applicant for a standard design certification, who knowingly provides to any licensee, applicant, contractor, or subcontractor, any components, equipment, materials, or other goods or services that relate to a licensee's or applicant's activities in this part, may not— (1) Engage in deliberate misconduct that causes or would have caused, if not detected, a licensee or applicant to be in violation of any rule, regulation, or order; or any term, condition, or 
                                    ( print page 87048)
                                    limitation of any license issued by the Commission; or (2) Deliberately submit to the NRC, a licensee, an applicant, or a licensee's or applicant's contractor or subcontractor, information that the person submitting the information knows to be incomplete or inaccurate in some respect material to the NRC. (b) A person who violates paragraph (a)(1) or (2) of this section may be subject to enforcement action in accordance with the procedures in subpart B of 10 CFR part 2. (c) For the purposes of paragraph (a)(1) of this section, deliberate misconduct by a person means an intentional act or omission that the person knows— (1) Would cause a licensee or applicant to be in violation of any rule, regulation, or order; or any term, condition, or limitation, of any license issued by the Commission; or (2) Constitutes a violation of a requirement, procedure, instruction, contract, purchase order, or policy of a licensee, applicant, contractor, or subcontractor.
 § 53.060
  (a) Discrimination by a Commission licensee, holder of a standard design approval, an applicant for a license, standard design certification, or standard design approval, a contractor or subcontractor of a Commission licensee, holder of a standard design approval, applicant for a license, standard design certification, or standard design approval, against an employee for engaging in certain protected activities is prohibited. Discrimination includes discharge and other actions that relate to compensation, terms, conditions, or privileges of employment. The protected activities are established in section 211 of the Energy Reorganization Act of 1974, as amended, and in general are related to the administration or enforcement of a requirement imposed under the Act or the Energy Reorganization Act of 1974, as amended. (1) The protected activities include but are not limited to— (i) Providing the Commission or his or her employer information about alleged violations of either of the statutes named in paragraph (a) of this section or possible violations of requirements imposed under either of those statutes; (ii) Refusing to engage in any practice made unlawful under either of the statutes named in paragraph (a) of this section or under these requirements if the employee has identified the alleged illegality to the employer; (iii) Requesting the NRC to institute action against his or her employer for the administration or enforcement of these requirements; (iv) Testifying in any Commission proceeding, or before Congress, or at any Federal or State proceeding regarding any provision (or proposed provision) of either of the statutes named in paragraph (a) of this section; and (v) Assisting or participating in, or being about to assist or participate in, these activities. (2) These activities are protected even if no formal proceeding is actually initiated as a result of the employee assistance or participation. (3) This section has no application to any employee alleging discrimination prohibited by this section who, acting without direction from his or her employer (or the employer's agent), deliberately causes a violation of any requirement of the Energy Reorganization Act of 1974, as amended, or the Act. (b) Any employee who believes that they have been discharged or otherwise discriminated against by any person for engaging in protected activities specified in paragraph (a)(1) of this section may seek a remedy for the discharge or discrimination through an administrative proceeding in the Department of Labor. The administrative proceeding must be initiated within 180 days after an alleged violation occurs. The employee may do this by filing a complaint alleging the violation with the Department of Labor, Wage and Hour Division. The Department of Labor may order reinstatement, back pay, and compensatory damages. (c) A violation of paragraph (a), (e), or (f) of this section by a Commission licensee, a holder of a standard design approval, an applicant for a Commission license, standard design certification, or a standard design approval, or a contractor or subcontractor of a Commission licensee, holder of a standard design approval, or any applicant may be grounds for— (1) Denial, revocation, or suspension of the license or standard design approval; (2) Withdrawal or revocation of a proposed or final standard design certification; (3) Imposition of a civil penalty on the licensee, holder of a standard design approval, or applicant (including an applicant for a standard design certification under this part following Commission adoption of final design certification rule) or a contractor or subcontractor of the licensee, holder of a standard design approval, or applicant; or (4) Other enforcement action. (d) Actions taken by an employer, or others, which adversely affect an employee may be predicated upon nondiscriminatory grounds. The prohibition applies when the adverse action occurs because the employee has engaged in protected activities. An employee's engagement in protected activities does not automatically render him or her immune from discharge or discipline for legitimate reasons or from adverse action dictated by nonprohibited considerations. (e)(1) Each holder or applicant for a license or design approval, must prominently post the revision of NRC Form 3, “Notice to Employees,” referenced in § 19.11(e)(1) of this chapter. This form must be posted at locations sufficient to permit employees protected by this section to observe a copy on the way to or from their place of work. Premises must be posted no later than 30 days after an application is docketed and remain posted while the application is pending before the Commission, during the term of the license, and for 30 days following license termination. (2) Copies of NRC Form 3 may be obtained by writing to the Regional Administrator of the appropriate NRC Regional Office listed in appendix D to 10 CFR part 20, via email to 
                                    *Forms.Resource@nrc.gov,*
                                     or by visiting the NRC's online library at 
                                    *http://www.nrc.gov/​reading-rm/​doc-collections/​forms/​.* (f) No agreement affecting the compensation, terms, conditions, or privileges of employment, including an agreement to settle a complaint filed by an employee with the Department of Labor pursuant to section 211 of the Energy Reorganization Act of 1974, as amended, may contain any provision which would prohibit, restrict, or otherwise discourage an employee from participating in protected activity as defined in paragraph (a)(1) of this section including, but not limited to, providing information to the NRC or to his or her employer on potential violations or other matters within NRC's regulatory responsibilities. (g) Part 19 of 10 CFR sets forth requirements and regulatory provisions applicable to licensees, holders of a standard design approval, applicants for a license, standard design certification, or standard design approval, and contractors or subcontractors of a Commission licensee, or holder of a standard design approval, and are in addition to the requirements in this section.
 § 53.070
  (a) Information provided to the Commission by a holder of a license, permit, design certification, or standard design approval under this part or an applicant for a license, permit, design certification, or standard design approval under this part, and information required by statute or by the Commission's regulations, orders, license conditions, or terms and conditions of a standard design approval to be maintained by the applicant or the licensee must be complete and accurate in all material respects. (b) Each applicant or licensee, each holder of a standard design approval under this part, and each applicant for a standard design certification under this part following Commission adoption of a final design certification regulation, must notify the Commission of information identified by the applicant or licensee as having for the regulated activity a significant implication for public health and safety or common defense and security. An applicant, licensee, or holder violates this paragraph only if the applicant, licensee, or holder fails to notify the Commission of information that the applicant, licensee, or holder has identified as having a significant implication for public health and safety or common defense and security. Notification must be provided to the Administrator of the appropriate Regional Office within 2 working days of identifying the information. This requirement is not applicable to information which is already required to be provided to the Commission by other reporting or updating requirements.
 § 53.080
  (a) The Commission may, upon application by any interested person or upon its own initiative, grant exemptions from the requirements of the regulations of this part, which are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security. (b) The Commission will not consider granting an exemption unless special circumstances are present. Special circumstances are present whenever— (1) Application of the regulation in the particular circumstances conflicts with other rules or requirements of the Commission; (2) Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule; (3) Compliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated; (4) The exemption would result in benefit to the public health and safety that compensates for any decrease in safety that may result from the grant of the exemption; (5) The exemption would provide only temporary relief from the applicable regulation and the licensee or applicant has made good faith efforts to comply with the regulation; or (6) There is present any other material circumstance not considered when the regulation was adopted for which it would be in the public interest to grant an exemption. If such condition is relied on exclusively for demonstrating compliance with paragraph (b) of this section, the exemption may not be granted until the Executive Director for Operations has consulted with the Commission. (c) Any person may request an exemption permitting the conduct of construction activities prior to the issuance of a CP. The Commission may grant such an exemption upon considering and balancing the following factors: (1) Whether conduct of the proposed activities will give rise to a significant adverse impact on the environment and the nature and extent of such impact, if any; (2) Whether redress of any adverse environment impact from conduct of the proposed activities can reasonably be effective should such redress be necessary; (3) Whether conduct of the proposed activities would foreclose subsequent adoption of alternatives; and (4) The effect of delay in conducting such activities on the public interest, including whether the power needs to be used by the proposed facility, the availability of alternative sources, if any, to meet those needs on a timely basis and delay costs to the applicant and to consumers. (d) Issuance of such an exemption must not be deemed to constitute a commitment to issue a CP. During the period of any exemption granted pursuant to paragraph (c) of this section, any activities conducted must be carried out in such a manner as will minimize or reduce their environmental impact. (e) The Commission's consideration of requests for exemptions from requirements of the regulations of other parts in this chapter that are applicable by virtue of this part must be governed by the exemption requirements of those parts.
 § 53.090
  (a) 
                                    *Common standards.*
                                     In determining that a CP, OL, early site permit, COL, or ML under this part will be issued to an applicant, the Commission will be guided by the following considerations: (1) Except for an early site permit or ML, the processes to be performed, the operating procedures, the facility and equipment, the use of the facility, and other technical specifications, or the proposals, in regard to any of the foregoing, collectively provide reasonable assurance that the applicant will comply with the regulations in this chapter, including the regulations in 10 CFR part 20, and that the health and safety of the public will not be endangered. (2) The applicant for a CP, OL, COL, or ML is technically and financially qualified to engage in the proposed activities in accordance with the regulations in this chapter. However, no consideration of financial qualification is necessary for an electric utility applicant for an OL for a utilization facility of the type described in paragraph (d) of this section or for an applicant for an ML. (3) The issuance of a CP, OL, early site permit, COL, or ML to the applicant will not, in the opinion of the Commission, be inimical to the common defense and security or to the health and safety of the public. (4) Any applicable requirements of subpart A of 10 CFR part 51 have been satisfied. (b) 
                                    *Additional standards for licenses.*
                                     In determining whether a license will be issued to an applicant, the Commission will, in addition to applying the standards set forth in paragraph (a) of this section, consider whether the proposed activities will serve a useful purpose proportionate to the quantities of SNM or source material to be utilized. (c) 
                                    *Additional standards and provisions affecting licenses for commercial power.*
                                     In addition to applying the standards set forth in paragraphs (a) and (b) of this section, paragraphs (c)(1) through (c)(4) of this section apply in the case of a license for a facility for the generation of commercial power. (1) The NRC will— (i) Give notice in writing of each application to the regulatory agency or State as may have jurisdiction over the rates and services incident to the proposed activity; (ii) Publish notice of the application in trade or news publications as it 
                                    ( print page 87050)
                                    deems appropriate to give reasonable notice to municipalities, private utilities, public bodies, and cooperatives which might have a potential interest in the utilization or production facility; and (iii) Publish notice of the application once each week for four consecutive weeks in the 
                                    **Federal Register**
                                    . No license will be issued by the NRC prior to the giving of these notices and until four weeks after the last notice is published in the 
                                    **Federal Register**
                                    . (2) If there are conflicting applications for a limited opportunity for such license, the Commission will give preferred consideration in the following order: first, to applications submitted by public or cooperative bodies for facilities to be located in high cost power areas in the United States; second, to applications submitted by others for facilities to be located in such areas; third, to applications submitted by public or cooperative bodies for facilities to be located in areas other than high cost power areas; and, fourth, to all other applicants. (3) The licensee who transmits electric energy in interstate commerce, or sells it at wholesale in interstate commerce, must be subject to the regulatory provisions of the Federal Power Act. (4) Nothing shall preclude any government agency, now or hereafter authorized by law to engage in the production, marketing, or distribution of electric energy, if otherwise qualified, from obtaining a CP, OL, or COL under this part for a utilization facility for the primary purpose of producing electric energy for disposition for ultimate public consumption. (d) 
                                    *Licenses for commercial nuclear plants.*
                                     A license will be issued, to an applicant who qualifies, for any one or more of the following: to transfer or receive in interstate commerce, or manufacture, produce, transfer, acquire, possess, or use a utilization facility for industrial or commercial purposes.
 § 53.100
  No permit, license, standard design approval, or standard design certification under this part shall be deemed to have been issued for activities that are not under or within the jurisdiction of the United States.
 § 53.110
  Licensees, applicants for licenses, permits, certifications, and design approvals, and applicants for an amendment to any license, permit, certification, or design approval under this part are not required to provide for design features or other measures for the specific purpose of protection against the effects of— (a) Attacks and destructive acts, including sabotage, directed against the facility by an enemy of the United States, whether a foreign government or other person; or (b) Use or deployment of weapons incident to U.S. defense activities.
 § 53.115
  (a) No right to the SNM will be conferred by a license issued under this part except as may be defined by the license. (b) Neither a license issued under this part, nor any right thereunder, nor any right to utilize or produce SNM may be transferred, assigned, or disposed of in any manner, either voluntarily or involuntarily, directly or indirectly, through transfer of control of the license to any person, unless the Commission, after securing full information, finds that the transfer is in accordance with the provisions of the Act and gives its consent in writing.
 § 53.117
  Any license issued under this part must be subject to suspension and to the rights of recapture of the material or control of the facility reserved to the Commission under section 108 of the Act in a state of war or national emergency declared by Congress.
 § 53.120
  (a) The NRC has submitted the information collection requirements contained in this part to the Office of Management and Budget (OMB) for approval as required by the Paperwork Reduction Act (44 U.S.C. 3501 
                                    *et seq.*). The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number. OMB has approved the information collection requirements contained in this part under control number 3150-XXXX. (b) The approved information collection requirements contained in this part appear in §§ 53.070, 53.080, 53.240, 53.410, 53.420, 53.425, 53.430, 53.440, 53.450, 53.480, 53.500, 53.540, 53.605, 53.610, 53.620, 53.700, 53.710, 53.715, 53.720, 53.730, 53.780, 53.785, 53.805, 53.810, 53.815, 53.830, 53.850, 53.855, 53.865, 53.870, 53.875, 53.880, 53.910, 53.1010, 53.1020, 53.1030, 53.1045, 53.1060, 53.1070, 53.1075, 53.1080, 53.1100, 53.1109, 53.1115, 53.1130, 53.1140, 53.1144, 53.1146, 53.1173, 53. 1182, 53.1188, 53.1200, 53.1206, 53.1209, 53.1210, 53.1221, 53.1230, 53.1236, 53.1239, 53.1241, 53.1254, 53.1257, 53,1263, 53.1270, 53.1276, 53.1279, 53.1282, 53.1288, 53.1295, 53.1300, 53.1306, 53.1309, 53.1312, 53.1327, 53.1330, 53.1333, 53.1336, 53.1348, 53.1360, 53.1366, 53.1369, 53.1372, 53.1384, 53.1410, 53.1413, 53.1416, 53.1419, 53.1437, 53.1449, 53.1452, 53.1458, 53.1470, 53.1505, 53.1510, 53.1515, 53.1525, 53.1530, 53.1535, 53.1540, 53.1545, 53.1550, 53.1560, 53.1565, 53.1570, 53.1575, 53.1580, 53.1620, 53.1630, 53.1645, 53.1680, 53.1690, 53.1720. (c) This part contains information collection requirements in addition to those approved under the control number specified in paragraph (a) of this section. The information collection requirement and the control numbers under which it is approved are as follows: (1) In §§ 53.765, 53.770, 53.780, and 53.795, NRC Form 396 is approved under control number 3150-0024. (2) In §§ 53.775 and 53.795, NRC Form 398 is approved under control number 3150-0090. (3) In § 53.1640, NRC Form 366 is approved under control number 3150-0104. (4) In § 53.1630, NRC Form 361 is approved under control number 3150-0238. (5) In § 53.1650, International Atomic Energy Agency Design Information Questionnaire forms are approved under control number 3150-0056. (6) In § 53.1650, DOC/NRC Form AP-A and associated forms are approved under control numbers 0694-0135.
 Subpart B—Technology-Inclusive Safety Requirements § 53.210
  Design features and programmatic controls must be provided for each commercial nuclear plant such that identification and analyses of design-basis accidents (DBAs) in accordance with § 53.240 demonstrate the following: (a) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release would not receive a radiation dose in excess of 25 rem (250 millisieverts) total effective dose equivalent (TEDE); and (b) An individual located at any point on the outer boundary of the low-population zone who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a radiation dose in 
                                    ( print page 87051)
                                    excess of 25 rem (250 millisieverts) TEDE.1 1
                                         The use of 25 rem TEDE is not intended to imply that this number constitutes an acceptable limit for an emergency dose to the public under accident conditions. Rather, this dose value has been set forth in this section as a reference value, which can be used in the evaluation of plant design features with respect to postulated reactor accidents, to assure that these designs provide assurance of low risk of public exposure to radiation, in the event of an accident.
 § 53.220
  Design features and programmatic controls must be provided for each commercial nuclear plant such that identification and analysis of licensing-basis events (LBEs) other than DBAs in accordance with § 53.240 demonstrate the following: (a) Plant SSCs, personnel, and programs provide the necessary capabilities and maintain the necessary reliability to address LBEs other than DBAs in accordance with §§ 53.240 and 53.450(e), and provide measures for defense in depth in accordance with § 53.250; and (b) The analysis of risks to public health and safety resulting from LBEs other than DBAs under § 53.450(e) includes comprehensive risk metrics that satisfy associated risk performance objectives that are acceptable to the NRC and provide an appropriate level of safety.
 § 53.230
  (a) The primary safety function is limiting the release of radioactive materials from the facility and must be maintained during normal operation and for LBEs over the life of the plant. (b) Additional safety functions needed to support the retention of radioactive materials during LBEs—such as controlling reactivity, heat generation, heat removal, and chemical interactions—must be identified for each commercial nuclear plant. (c) The primary and additional safety functions are required to satisfy the safety criteria defined in §§ 53.210 and 53.220, or more restrictive alternative criteria adopted under § 53.470, and must be fulfilled by the design features, human actions, and programmatic controls specified throughout this part.
 § 53.240
  (a) Licensing-basis events must be identified for each commercial nuclear plant and analyzed under § 53.450 to demonstrate that the safety requirements in this subpart have been satisfied. (b) The identified LBEs, ranging from anticipated event sequences to very unlikely event sequences, must collectively address combinations of malfunctions of plant SSCs, human errors, facility hazards, and the effects of external hazards. (c) The analysis of LBEs must— (1) Include analysis of one or more DBAs under § 53.450(f); (2) Confirm the adequacy of design features and programmatic controls needed to satisfy the safety criteria defined in §§ 53.210 and 53.220, or more restrictive alternative criteria adopted under § 53.470, and (3) Establish related functional requirements for plant SSCs, personnel, and programs.
 § 53.250
  (a) Measures must be taken for each commercial nuclear plant to ensure appropriate defense in depth is provided to compensate for uncertainties in the analysis of the safety criteria such that there is reasonable assurance that the safety criteria in this subpart are met over the life of the plant. (b) The uncertainties that must be addressed under paragraph (a) of this section include those related to the state of knowledge and modeling capabilities, the ability of barriers to limit the release of radioactive materials from the facility during LBEs other than DBAs, the reliability and performance of plant SSCs and personnel, and the effectiveness of programmatic controls. (c) The safety analysis may not rely upon a single engineered design feature, human action, or programmatic control, no matter how robust, to address the range of LBEs other than DBAs.
 § 53.260
  Holders of licenses to operate commercial nuclear plants under this part must control public doses and dose rates in unrestricted areas from normal plant operations to meet the requirements in 10 CFR part 20.
 § 53.270
  Holders of licenses to operate commercial nuclear plants under this part must control occupational doses to meet the requirements in 10 CFR part 20.
 Subpart C—Design and Analysis Requirements § 53.400
  (a) Design features must be provided for each commercial nuclear plant such that, when combined with corresponding human actions and programmatic controls, the plant will satisfy the safety criteria defined in §§ 53.210 and 53.220, or more restrictive alternative criteria adopted under § 53.470. (b) Design features must ensure that the safety functions identified in § 53.230 are fulfilled during licensing-basis events (LBEs).
 § 53.410
  (a) Functional design criteria must be defined for each design feature required by § 53.400 and relied upon to demonstrate compliance with the safety criteria defined in § 53.210. (b) Corresponding human actions and programmatic controls must be identified and implemented in accordance with this and other subparts to achieve and maintain the reliability and capability of structures, systems, and components (SSCs) relied upon to satisfy the defined functional design criteria and the safety criteria required in § 53.210, and to maintain consistency with analyses required by § 53.450(f).
 § 53.415
  Safety-related (SR) SSCs must be protected against or must be designed to withstand the effects of natural phenomena (
                                    *e.g.,*
                                     earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches) and constructed hazards (
                                    *e.g.,*
                                     dams, transportation routes, military and industrial facilities) considering an event severity up to the design-basis external hazard levels as determined under § 53.510 without losing the capability to perform the safety functions identified under § 53.230. Specific requirements for earthquake engineering are included in § 53.480.
 § 53.420
  (a) Functional design criteria must be defined for each design feature required by § 53.400 and relied upon to— (1) Demonstrate compliance with the safety criteria in § 53.220 or more restrictive alternative criteria adopted under § 53.470; and (2) Demonstrate compliance with the evaluation criteria in § 53.450(e) or more restrictive alternative criteria adopted under § 53.470. (b) Corresponding human actions and programmatic controls must be identified and implemented in accordance with this and other subparts to achieve and maintain the reliability and capability of SSCs relied upon to—
                                    ( print page 87052) (1) Satisfy the safety criteria in § 53.220 or more restrictive alternative criteria adopted under § 53.470; and (2) Satisfy the evaluation criteria in § 53.450(e) or more restrictive alternate criteria adopted under § 53.470.
 § 53.425
  (a) Design features must be provided for each commercial nuclear plant to support the Radiation Protection Program required in § 53.850. (b) Functional design criteria must be defined for each design feature relied upon to demonstrate compliance with § 53.850. (c) Functional design criteria, including design objectives for dose to the maximally exposed member of the public, must be defined for design features to show that plant design features and corresponding programmatic controls, including monitoring programs, control liquid, gaseous, and solid wastes, as required under part 20 of this chapter.1 1
                                         A guide for keeping doses to the public as low as is reasonably achievable is that the estimated annual dose to the maximally exposed member of the public does not exceed 10 mrem total effective dose equivalent. A design objective of maintaining doses below 10 mrem/year should not be construed as a radiation protection standard.
 § 53.430
  (a) Design features must be provided for each commercial nuclear plant such that, when combined with corresponding programmatic controls, the requirements in § 53.270 can be met. (b) Functional design criteria must be defined for each design feature relied upon to demonstrate compliance with § 53.270.
 § 53.440
  (a)(1) Analysis, appropriate test programs, prototype testing, operating experience, or a combination thereof must demonstrate that each design feature required by § 53.400 meets the defined functional design criteria required by §§ 53.410 and 53.420. This demonstration must consider interdependent effects throughout the commercial nuclear plant and the range of conditions under which the design features required by § 53.400 must function throughout the plant's lifetime. (2) The design processes for SR and non-safety-related but safety-significant (NSRSS) SSCs under this part must include administrative procedures for evaluating operating, design, and construction experience and for considering applicable important industry experiences in the design of those SSCs. (b) The design features required by § 53.400 must, wherever applicable, be designed using generally accepted consensus codes and standards that have been endorsed or otherwise found acceptable by the U.S. Nuclear Regulatory Commission (NRC). (c) The materials used for each SR and NSRSS SSC must be qualified for their service conditions over the design life of the SSC. (d) Possible degradation mechanisms related to aging, fatigue, chemical interactions, operating temperatures, effects of irradiation, and other environmental factors that may affect the performance of SR and NSRSS SSCs must be evaluated and used to inform the design and the development of integrity assessment programs under § 53.870. (e)(1) Safety-related and NSRSS SSCs must be designed and located to minimize, consistent with other safety requirements in this part, the probability and effect of fires and explosions. (2) Noncombustible and fire-resistant materials must be used wherever practical throughout the facility, particularly in locations with SR and NSRSS SSCs. (3) Fire detection and fire suppression systems of appropriate capacity and capability must be provided and designed to minimize the adverse effects of fires on SR and NSRSS SSCs. (4) Fire suppression systems must be designed to ensure that their rupture or inadvertent operation does not significantly impair the ability of SR and NSRSS SSCs to perform their safety functions to satisfy § 53.230. (f) Safety and security must be considered together in the design process such that, where possible, security issues are effectively resolved through design and engineered security features. (g) The reactor system and waste stores for each commercial nuclear plant must be capable of achieving and maintaining a subcritical condition during normal operations and following any LBE identified in accordance with § 53.240. (h) Each commercial nuclear plant must have a capability to provide long-term cooling of the reactor fuel and waste stores during normal operations and following any LBE identified in accordance with § 53.240. (i) The design, analysis, staffing, and programmatic controls for each commercial nuclear plant must consider the number of reactors, waste stores, and other significant inventories of radioactive materials and the associated operating configurations, common systems, system interfaces, and system interactions. (j)(1) Design features must be provided and related functional design criteria defined such that, with limited use of operator actions, one or more physical barriers are maintained to limit the release of radionuclides from reactor systems, waste stores, or other significant inventories of radioactive materials assuming the impact of a large, commercial aircraft. (2) The functional design criteria for those design features provided to address the requirements in paragraph (j)(1) of this section must be based on an assessment of the impact of a large, commercial aircraft used for long distance flights in the United States, with aviation fuel loading typically used in such flights, and an impact speed and angle of impact considering the ability of both experienced and inexperienced pilots to control large, commercial aircraft at low altitude representative of a commercial nuclear plant's low profile.1 1
                                         Changes to the detailed parameters on aircraft impact characteristics set forth in guidance must be approved by the Commission. (k) Design features and related functional design criteria must be defined such that analyses demonstrate a low risk of permanent injury to the public due to the health effects of the chemical hazards of licensed material. (l) Measures must be taken during the design of commercial nuclear plants to minimize, to the extent practicable, contamination of the facility and the environment, facilitate eventual decommissioning, and minimize, to the extent practicable, the generation of radioactive waste in accordance with § 20.1406 of this chapter. (m)(1) Each commercial nuclear plant must include criticality monitoring capabilities meeting the requirements of either § 70.24 of this chapter or paragraph (m)(2) of this section. (2) In lieu of maintaining a monitoring system capable of detecting criticality as described in § 70.24 of this chapter, criticality accident requirements may be satisfied by— (i) Demonstrating the sub-criticality of special nuclear material, except when it is inside the reactor and the reactor is being operated, by maintaining k-effective below 0.95 at a 95 percent probability, 95 percent confidence level, under conditions that maximize reactivity for the applicable storage and handling configurations, and (ii) Providing radiation monitors for fuel storage and associated handling 
                                    ( print page 87053)
                                    areas when fuel is present to detect excessive radiation levels and to support initiating appropriate safety actions. (3) While a spent fuel transportation package approved under 10 CFR part 71 of this chapter or spent fuel storage cask approved under 10 CFR part 72 is in the special nuclear material handing or storage area, the requirements in 10 CFR parts 71 or 72, as applicable, and the requirements of the certificate of compliance for that package or cask, are the applicable requirements for the fuel within that package or cask. (n)(1) The design of each commercial nuclear plant must reflect state-of-the-art human factors principles for safe and reliable performance in all locations that human activities are expected for performing or supporting the continued availability of plant safety or emergency response functions. (2) The design must provide for the capabilities described in § 53.730(b) to ensure the plant staff are able to monitor plant conditions and respond to events. (3) The means by which the design and human actions together will achieve the safety requirements of subpart B of this part must be evaluated and used to inform the design and the development of the concept of operations required by § 53.730(c). (4) A functional requirements analysis and function allocation must be used to ensure that plant design features address how safety functions and functional safety criteria are satisfied, and how the safety functions will be assigned to appropriate combinations of human action, automation, active safety features, passive safety features, or inherent safety characteristics.
 § 53.450
  (a) 
                                    *Requirement to have a probabilistic risk assessment (PRA).*
                                     A PRA of each commercial nuclear plant must be performed to identify potential failures, susceptibility to internal and external hazards, and other contributing factors to event sequences that might challenge the safety functions identified in § 53.230 and to support demonstrating that each commercial nuclear plant meets the safety criteria of § 53.220, or more restrictive alternative criteria adopted under § 53.470. (b) 
                                    *Specific uses of analyses.*
                                     The PRA in combination with other generally accepted approaches for systematically evaluating engineered systems must be used— (1) In informing the selection of the LBEs, as described in § 53.240, which must be considered in the design to determine compliance with the safety criteria in subpart B of this part. (2) For informing the classification of SSCs according to their safety significance in accordance with § 53.460 and for identifying the environmental conditions under which the SSCs and operating staff must perform their safety functions. (3) In evaluating the adequacy of defense-in-depth measures required in accordance with § 53.250. (4) To identify and assess all plant operating states where there is the potential for the uncontrolled release of radioactive material to the environment. (5) To identify and assess events that challenge plant control and safety systems whose failure could lead to the uncontrolled release of radioactive material to the environment. These include internal events, such as human errors and equipment failures, and external events identified in accordance with subpart D of this part. (c) 
                                    *Maintenance and upgrade of analyses.*
                                     The PRA must be maintained at least every 5 years until the permanent cessation of operations under § 53.1070 and upgraded in conformance with generally accepted methods, standards, and practices that have been endorsed or otherwise found acceptable by the NRC. (d) 
                                    *Qualification of analytical codes.*
                                     The analytical codes used in modeling plant behavior in analyses of licensing-basis events (including but not limited to thermodynamics, reactor physics, fuel performance, and mechanistic source term codes) must be qualified for the range of conditions for which they are to be used. (e) 
                                    *Analyses of licensing-basis events other than design-basis accidents.* (1) Analyses must be performed for LBEs other than design-basis accidents (DBAs). These LBEs must be identified using insights from a PRA in combination with other generally accepted approaches for systematically evaluating engineered systems to identify and analyze equipment failures and human errors. (2) The analysis of LBEs other than DBAs must include definition of evaluation criteria for each event or specific categories of LBEs to determine the acceptability of the plant response to the challenges posed by internal and external hazards to provide an appropriate level of safety. (3) The analyses of LBEs other than DBAs must address event sequences from initiation to a defined end state and be used in combination with other engineering analyses to demonstrate that the functional design criteria required by § 53.420 provide sufficient barriers to the unplanned release of radionuclides to satisfy the evaluation criteria defined for each LBE other than DBAs, to satisfy the safety criteria specified in accordance with § 53.220 and provide defense in depth as required by § 53.250. (4) The methodology used to identify, categorize, and analyze LBEs must include a means to identify event sequences deemed significant for controlling the risks posed to public health and safety. (f) 
                                    *Analysis of design-basis accidents.*
                                     (1) The analysis of LBEs required by § 53.240 must include analysis of DBAs that address possible challenges to the safety functions identified under § 53.230. The events selected as DBAs must be those that, if not terminated, have the potential for exceeding the safety criteria in § 53.210. (2) The DBAs selected must be analyzed using deterministic methods that address event sequences from initiation to a safe stable end state and assume only the SR SSCs identified under § 53.460 and human actions addressed by the requirements of subpart F of this part are available to perform the safety functions identified in accordance with § 53.230. (3) The analysis must conservatively demonstrate compliance with the safety criteria in § 53.210. (g) 
                                    *Other required analyses.*
                                     Analyses must be performed to assess— (1) 
                                    *Fire protection.*
                                     Fire protection measures to demonstrate, through inclusion of fires in the analysis of LBEs or by separate analyses, that a fire or explosion in any plant area would not— (i) Prevent equipment from fulfilling the safety functions identified in accordance with § 53.230, or (ii) Challenge the safety criteria in §§ 53.210 and 53.220. (2) 
                                    *Aircraft impact.*
                                     Measures provided to protect against aircraft impacts under § 53.440(j). (3) 
                                    *Dose to members of the public.*
                                     Measures taken under § 53.425, including estimating— (i) The quantity of each of the principal radionuclides expected to be released annually to unrestricted areas in liquid effluents produced during normal reactor operations and the dose to the maximally exposed member of the public in unrestricted areas. (ii) The quantities of each of the principal radionuclides of the gases, halides, and particulates expected to be released annually to unrestricted areas in gaseous effluents produced during normal reactor operations and the dose to the maximally exposed member of the public in unrestricted areas. (iii) The annual external radiation dose in unrestricted areas and the maximally exposed member of the 
                                    ( print page 87054)
                                    public in unrestricted areas due to direct radiation from contained radiation sources from the commercial nuclear plant during normal reactor operations.
 § 53.460
  (a) Structures, systems, and components must be classified according to their safety significance. The SSC categories must include “Safety-Related,” “Non-Safety-Related but Safety-Significant,” and “Non-Safety-Significant,” as defined in subpart A of this part. (b) For SR and NSRSS SSCs, the conditions under which they must perform their safety function in § 53.230 must be identified. Special treatments must be established in accordance with this and other subparts to provide confidence that the SSCs will perform under the service conditions and with reliability consistent with the analysis performed under § 53.450 to demonstrate meeting the safety criteria in §§ 53.210 and 53.220, or more restrictive alternative criteria adopted under § 53.470. (1) The special treatments for SR SSCs must include meeting the applicable quality assurance requirements from appendix B of part 50 of this chapter. (2) The special treatments for NSRSS SSCs and special treatments for SR SSCs beyond those required under (b)(1) of this section may include meeting selected quality assurance requirements from appendix B of part 50 of this chapter when such treatment is needed to address performance requirements, equipment reliability, or uncertainties. (c) Human actions needed to prevent or mitigate LBEs must be identified, be able to be performed reliably under the postulated environmental conditions, and be addressed by programs established in accordance with subpart F of this part to provide confidence that those actions will be performed as assumed in the analysis performed in accordance with § 53.450 to demonstrate meeting the criteria in §§ 53.210, 53.220, and 53.450(e), or more restrictive alternative criteria adopted under § 53.470.
 § 53.470
  Where an applicant or licensee so chooses, alternative criteria more restrictive than those defined in §§ 53.220 and 53.450(e) may be adopted to support operational flexibilities. In such cases, applicants and licensees must ensure that the functional design criteria of § 53.420, the analysis requirements of § 53.450(e), and identification of special treatment of SSCs and human actions under § 53.460 reflect and support the use of alternative criteria to justify operational flexibilities. Licensees must ensure that measures taken to provide the analytical margins supporting operational flexibilities are incorporated into design features and programmatic controls and are maintained within programs required in other subparts.
 § 53.480
  (a) 
                                    *Effects of earthquakes.*
                                     Structures, systems, and components classified as SR or NSRSS must be able to withstand the effects of earthquakes, commensurate with the safety significance of the SSC, without loss of capability to perform their role in fulfilling the safety functions required by § 53.230. (b) 
                                    *Definitions.*
                                     For the purpose of this section— *Design-Basis Ground Motions (DBGMs)*
                                     are the vibratory ground motions for which certain SSCs must be designed to remain functional. *Operating basis earthquake (OBE) ground motion*
                                     is the vibratory ground motion for which those features of the commercial nuclear plant necessary for continued operation without undue risk to the health and safety of the public are designed to remain functional. The OBE ground motion is used in § 53.720. *Response spectrum*
                                     is a plot of the maximum responses (acceleration, velocity, or displacement) of idealized single-degree-of-freedom oscillators as a function of the natural frequencies of the oscillators for a given damping value. The response spectrum is calculated for a specified vibratory motion input at the oscillators' supports. *Surface deformation*
                                     is the distortion of geologic strata on or near the ground surface that occurs because of tectonic forces that result from earthquakes. (c) 
                                    *Design considerations—*
                                    (1) 
                                    *Design-Basis Ground Motions.*
                                     (i) The DBGMs must be derived from the Site Ground Motion Response Spectra developed in accordance with § 53.510(c), by taking into consideration the functional design criteria of SSCs in accordance with §§ 53.410 and 53.420. The horizontal component of the DBGM(s) in the free-field at the foundation level of the structures must be an appropriate response spectrum that is determined based on the risk-significance of SSCs and their safety functions. In view of the limited data available on vibratory ground motion of strong earthquakes, it is acceptable that the design response spectra be smoothed spectra. (ii) The commercial nuclear plant must be designed so that, if the DBGMs occur, the following SSCs remain functional and within applicable stress, strain, and deformation limits: (A) Structures, systems, and components for which functional design criteria are established in accordance with § 53.410 or § 53.420; and (B) Structures, systems, and components classified as SR or NSRSS commensurate with safety significance in accordance with § 53.460. (iii) In addition to seismic loads, applicable concurrent normal operating, functional, and accident-induced loads must be taken into account in the design of the SR SSCs and, commensurate with safety significance, NSRSS SSCs. (iv) The design of the commercial nuclear plant must take into account the possible effects of seismic-induced ground disruption, such as fissuring, lateral spreads, differential settlement, liquefaction, and landsliding, on the facility foundations. (v) The SSCs fulfilling the safety functions required by § 53.230 must be demonstrated through design, testing, or qualification methods to be able to fulfill those safety functions during and after the vibratory ground motion associated with the DBGMs. (vi) The evaluation of SSCs required by this section to show they are able to function during and after earthquake ground motion must take into account soil-structure interaction effects and the expected duration of vibratory motion. It is permissible to design for strain limits in excess of yield strain in some of these SSCs during the DBGMs and under the postulated concurrent loads, provided the necessary safety functions are maintained. (2) 
                                    *OBE Ground Motion.*
                                     The OBE Ground Motion must be characterized by response spectra. The value of the OBE Ground Motion must be set to one-third or less of the DBGMs response spectra. (3) [Reserved] (4) 
                                    *Required seismic instrumentation.*
                                     Suitable instrumentation must be provided so that the seismic response of commercial nuclear plant SR SSCs or NSRSS SSCs can be evaluated promptly after an earthquake. (d) 
                                    *Surface deformation.*
                                     (1) The potential for surface deformation must be taken into account in the design of the commercial nuclear plant by providing reasonable assurance that in the event of deformation, SSCs classified as SR or NSRSS in accordance with § 53.460 will remain functional. (2) In addition to surface deformation induced loads, the design of SSCs must take into account, commensurate with 
                                    ( print page 87055)
                                    safety significance, seismic loads and applicable concurrent functional and accident-induced loads. (3) The design provisions for surface deformation must be based on its postulated occurrence in any direction and azimuth and under any part of the commercial nuclear plant, unless evidence indicates this assumption is not appropriate, and must take into account the estimated rate at which the surface deformation may occur. (e) 
                                    *Seismically induced floods and water waves and other design conditions.*
                                     Seismically induced floods and water waves from either locally or distantly generated seismic activity and other design conditions determined pursuant to subpart D of this part must be taken into account in the design of the commercial nuclear plant so as to prevent undue risk to the health and safety of the public. (f) 
                                    *Analysis.*
                                     The analyses required by § 53.450 must address seismic hazards and related SSC responses in determining that the safety criteria defined in § 53.220 will be met. (g) 
                                    *Design criteria, human actions, and programmatic controls.*
                                     Functional design criteria, human actions, and programmatic controls needed to address seismic events must be identified and implemented in accordance with this and other subparts to achieve and maintain the performance of SSCs relied upon to satisfy the safety criteria in § 53.220 and to maintain consistency with analyses required by § 53.450 when accounting for the site-specific frequencies and magnitudes of earthquakes for a commercial nuclear plant.
 Subpart D—Siting Requirements § 53.500
  (a) The siting of each commercial nuclear plant must be supported by assessments of proposed sites such that the design, including design features and programmatic controls corresponding to the site characteristics, satisfies the safety criteria defined in §§ 53.210 and 53.220 or more restrictive alternative criteria adopted under § 53.470. The siting assessment must ensure that site characteristics that might contribute to the initiation, progression, or consequences of licensing-basis events (LBEs) analyzed under §§ 53.450 and 53.480 are identified and mitigated by design features or programmatic controls. The siting assessment must take into consideration the potential adverse impacts that a commercial nuclear plant may have on nearby populations as a result of normal operations or LBEs. (b) Activities performed to identify site characteristics or otherwise needed to determine site-specific contributors to functional design criteria or analysis assumptions under subpart C of this part must satisfy the applicable special treatment requirements of § 53.460, including, where applicable, the quality assurance requirements from appendix B of part 50 of this chapter.
 § 53.510
  (a) 
                                    *General external hazard requirements.*
                                     The design-basis external hazard level for the relevant external hazards for a site must be identified and characterized based on site-specific assessments of natural and constructed hazards with the potential to adversely affect plant functions. The external hazard frequencies and magnitudes determined from the site-specific assessments must take into account uncertainties and variabilities in data, models, and methods relied on to characterize the external hazards. (b) 
                                    *Definitions.*
                                     For the purpose of this section, the following terms mean: *Geological siting factors*
                                     are geological and seismic factors that may affect the design and operation of the proposed commercial nuclear plant. *Ground Motion Response Spectra (GMRS)*
                                     are the site-specific GMRS resulting from the geologic investigations and evaluations of the site vicinity and region and used to determine design-basis ground motions for structures, systems, and components under § 53.480. *Probabilistic seismic hazard analysis*
                                     is an analytical methodology that incorporates uncertainty into estimates of an annual frequency of exceedance for a certain ground motion parameter (
                                    *e.g.,*
                                     peak ground acceleration, peak ground velocity, response spectral values) at a site. (c) 
                                    *Geological investigations.*
                                     The GMRS for the site must be determined based on the results of investigations of the geological, seismological, and engineering characteristics of the site and its environs and must be characterized by both horizontal and vertical free-field GMRS at the free ground surface. The size of the region to be investigated and the type of data pertinent to the investigations must be determined based on the nature of the region surrounding the site. Data on vibratory ground motion, earthquake recurrence rates, fault geometry and slip rates, and site subsurface material properties must be obtained by reviewing pertinent literature and carrying out field investigations. Uncertainties are inherent in the parameters and models used to estimate the GMRS for the site. The site assessment must reflect these uncertainties through an appropriate analysis, such as a probabilistic seismic hazard analysis. (d) 
                                    *Geologic and seismic siting factors.*
                                     The geologic and seismic siting factors considered for design under §§ 53.415 and 53.480 must include, but are not limited to, determination of the potential for surface tectonic and nontectonic deformations, the size and character of seismically induced floods and water waves that could affect a site from either locally or distantly generated seismic activity, soil and rock stability, liquefaction potential, and natural and artificial slope stability.
 § 53.520
  Site characteristics that might contribute to the initiation, progression, or consequences of LBEs analyzed under § 53.450 must be identified, assessed, and considered in the design and analyses required by subpart C of this part.
 § 53.530
  Every site must have an exclusion area, a low-population zone, and a population center distance as defined in § 53.020. (a) The offsite radiological consequences estimated by the analyses required by § 53.450(f) must be used to confirm that— (1) An individual located at any point on the boundary of the exclusion area for any 2-hour period following onset of the postulated fission product release would not receive a radiation dose in excess of 25 rem (250 millisieverts) total effective dose equivalent. (2) An individual located at any point on the outer boundary of the low-population zone who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a radiation dose in excess of 25 rem (250 millisieverts) total effective dose equivalent. (b) The population center distance must be at least one and one-third times the distance from the reactor to the outer boundary of the low-population zone. The boundary of the population center must be determined upon consideration of population distribution. Political boundaries are not controlling in the calculation of population center distance. (c) Reactor sites should be located away from very densely populated centers. Areas of low-population density 
                                    ( print page 87056)
                                    are, generally, preferred. However, in determining the acceptability of a particular site located away from a very densely populated center but not in an area of low-population density, consideration will be given to safety, environmental, economic, or other factors, which may result in the site being found acceptable.
 § 53.540
  Site characteristics must be addressed by the design features, programmatic controls, and supporting analyses used to demonstrate that the safety criteria in §§ 53.210 and 53.220 are met for each commercial nuclear plant. Site characteristics must be such that adequate emergency plans and security plans can be developed and maintained.
 Subpart E—Construction and Manufacturing Requirements § 53.600
  This subpart applies to those construction and manufacturing activities authorized by a construction permit (CP), combined license (COL), manufacturing license (ML), or limited work authorization (LWA) issued under this part.
 § 53.605
  Each CP and ML issued under this part is subject to the terms and conditions in this section, and each COL issued under this part is subject to the terms and conditions in this section until the date that the Commission makes the finding under § 53.1452(g). (a) 
                                    *Definitions.*
                                     The definitions in § 21.3 of this chapter apply to this section. (b) 
                                    *Posting requirements.*
                                     (1) Each individual, partnership, corporation, dedicating entity, or other entity subject to the regulations in this section must post current copies of this section and the regulations in 10 CFR part 21; section 206 of the Energy Reorganization Act of 1974, as amended; and procedures adopted under these regulations. These documents must be posted in a conspicuous position on any premises within the United States where the activities subject to the license are conducted. (2) If posting of these regulations or the procedures adopted under them is not practical, the licensee may, in addition to posting section 206 of the Energy Reorganization Act of 1974, as amended, post a notice that describes the regulations/procedures, including the name of the individual to whom reports may be made, and states where they may be examined. (c) 
                                    *Procedures.*
                                     The holder of a CP, COL, or ML subject to this section must adopt appropriate procedures to— (1) Evaluate deviations and failures to comply to identify defects and failures to comply associated with substantial safety hazards as soon as practicable, and, except as provided in paragraph (c)(2) of this section, in all cases within 60 days of discovery, to identify a reportable defect or failure to comply that could create a substantial safety hazard, were it to remain uncorrected. (2) Ensure that if an evaluation of an identified deviation or failure to comply potentially associated with a substantial safety hazard cannot be completed within 60 days from the discovery of the deviation or failure to comply, an interim report is prepared and submitted to the Commission through a director or responsible officer, or designated person as discussed in paragraph (d)(5) of this section. The interim report should describe the deviation or failure to comply that is being evaluated and should also state when the evaluation will be completed. This interim report must be submitted in writing within 60 days of discovery of the deviation or failure to comply. (3) Ensure that a director or responsible officer of the holder of a CP, COL, or ML subject to this section is informed as soon as practicable, and, in all cases, within the 5 working days after completion of the evaluation described in paragraph (c)(1) or (c)(2) of this section, if the construction or manufacture of a facility or activity, or a basic component supplied for such a facility or activity— (i) Fails to comply with the Atomic Energy Act of 1954, as amended, or any applicable regulation, order, or license of the Commission relating to a substantial safety hazard; (ii) Contains a defect; or (iii) Underwent any significant breakdown in any portion of the quality assurance program (QAP) conducted under the requirements of appendix B to part 50 of this chapter that could have produced a defect in a basic component. These breakdowns in the QAP are reportable whether or not the breakdown actually resulted in a defect in a design approved and released for construction, installation, or manufacture. (d) 
                                    *Reporting defects and noncompliance.*
                                     (1) The holder of a CP, COL, or ML subject to this section that obtains information reasonably indicating that the facility or manufactured reactors fails to comply with the Atomic Energy Act of 1954, as amended, or any applicable regulation, order, or license of the Commission relating to a substantial safety hazard must notify the Commission of the failure to comply through a director, responsible officer, or designated person as discussed in paragraph (d)(5) of this section. (2) The holder of a CP, COL, or ML subject to this section that obtains information reasonably indicating the existence of any defect found in the construction or manufacture, or any defect found in the final design of a facility as approved and released for construction or manufacture, must notify the Commission of the defect through a director, responsible officer, or designated person as discussed in paragraph (d)(5) of this section. (3) The holder of a CP, COL, or ML subject to this part, who obtains information reasonably indicating that the QAP has undergone any significant breakdown discussed in paragraph (c)(3)(iii) of this section must notify the Commission of the breakdown in the QAP through a director, responsible officer, or designated person as discussed in paragraph (d)(5) of this section. (4) When acting as a dedicating entity, the holder of a CP, COL, or ML subject to this section is responsible for identifying and evaluating deviations; reporting defects and failures to comply associated with substantial safety hazards for dedicated items; and maintaining auditable records for the dedication process. (5) The notification requirements of this paragraph apply to all defects and failures to comply associated with a substantial safety hazard regardless of whether extensive evaluation, redesign, or repair is required to conform to the criteria and bases stated in the Safety Analysis Report, CP, COL, or ML. Evaluation of potential defects and failures to comply and reporting of defects and failures to comply under this section satisfies the CP holder's, COL holder's, and ML holder's evaluation and notification obligations under 10 CFR part 21, and satisfies the responsibility of individual directors or responsible officers or holders of a CP, COL, or ML subject to this section to report defects, and failures to comply associated with substantial safety hazards under section 206 of the Energy Reorganization Act of 1974, as amended. The director or responsible officer may authorize an individual to provide the notification required by this section. However, this does not relieve the director or responsible officer of his or her responsibility under this section.
                                    ( print page 87057) (e) 
                                    *Notification—timing and where sent.*
                                     The notification required by paragraph (d) of this section must consist of— (1) Initial notification by telephone, facsimile, or email identified in appendix A to 10 CFR part 73 to the U.S. Nuclear Regulatory Commission (NRC) Operations Center within 2 days following receipt of information by the director or responsible corporate officer under paragraph (c)(3) of this section, on the identification of a defect or a failure to comply. If the CP, COL, or ML holder elects to use facsimile, verification that the facsimile has been received should be made by calling the NRC Operations Center. This paragraph does not apply to interim reports described in paragraph (c)(2) of this section. (2) Written notification submitted to the NRC Document Control Desk by an appropriate method listed in § 53.040, with a copy to the appropriate NRC Regional Administrator at the address specified in appendix D to 10 CFR part 20 and a copy to the appropriate NRC resident inspector, if applicable, within 30 days following receipt of information by the director or responsible corporate officer under paragraph (c)(3) of this section, on the identification of a defect or failure to comply. (f) 
                                    *Content of notification.*
                                     The written notification required by paragraph (e)(2) of this section must clearly indicate that the written notification is being submitted under this section and include the following information, to the extent known. (1) Name and address of the individual or individuals informing the Commission. (2) Identification of the facility, the activity, or the basic component supplied for the facility or the activity within the United States which contains a defect or fails to comply. (3) Identification of the firm constructing or manufacturing the facility or supplying the basic component which fails to comply or contains a defect. (4) Nature of the defect or failure to comply and the safety hazard which is created or could be created by the defect or failure to comply. (5) The date on which the information of a defect or failure to comply was obtained. (6) In the case of a basic component that contains a defect or failure to comply, the number and location of these components in use at the facility subject to the regulations in this part. (7) In the case of a completed reactor manufactured under this part, the entities to which the reactor was supplied. (8) The corrective action which has been, is being, or will be taken; the name of the individual or organization responsible for the action; and the length of time that has been or will be taken to complete the action. (9) Any advice related to the defect or failure to comply about the facility, activity, or basic component that has been, is being, or will be given to other entities. (g) 
                                    *Procurement documents.*
                                     Each holder of a CP, COL, or ML subject to this section must ensure that each procurement document for a facility or a basic component specifies the provisions of 10 CFR part 21 or this section that apply, as applicable. (h) 
                                    *Coordination with 10 CFR part 21.*
                                     The requirements of this section are satisfied when the defect or failure to comply associated with a substantial safety hazard has been previously reported under 10 CFR part 21, under § 73.1205 of this chapter, under this section, or under § 53.1640. (i) 
                                    *Records retention.*
                                     The holder of a CP, COL, or ML subject to this section must prepare and maintain records necessary to accomplish the purposes of this section, specifically— (1) Retain procurement documents, which define the requirements that facilities or basic components must satisfy in order to be considered acceptable, for the lifetime of the facility or basic component. (2) Retain records of evaluations of all deviations and failures to comply under paragraph (c)(1) of this section for the longest of— (i) Ten years from the date of the evaluation; (ii) Five years from the date that an early site permit is referenced in an application for a COL; or (iii) Five years from the date of delivery of a manufactured reactor. (3) Retain records of all interim reports to the Commission made under paragraph (c)(2) of this section, or notifications to the Commission made under paragraph (d) of this section for the minimum time periods stated in paragraph (i)(2) of this section; (4) Suppliers of basic components must retain records of— (i) All notifications sent to affected licensees or purchasers under paragraph (d)(4) of this section for a minimum of 10 years following the date of the notification; (ii) The facilities or other purchasers to whom the basic components or associated services were supplied for a minimum of 15 years from the delivery of the basic component or associated services. (5) Maintaining reports in accordance with this section satisfies the recordkeeping obligations under 10 CFR part 21 of the entities, including directors or responsible officers thereof, subject to this section.
 § 53.610
  (a) 
                                    *Management and control.*
                                     Licensees must ensure that the following plans, programs, and organizational units are developed and implemented to manage and control the construction activities: (1) Programs to ensure that the construction of a commercial nuclear plant supports the eventual compliance with the design and analysis requirements in subpart C of this part. (2) An organization, headed by qualified personnel, responsible for managing, controlling, and evaluating the adequacy of the construction activities. (3) Procedures describing the qualifications for personnel in key positions in the licensee's management and control organization and the organizational responsibilities, authority, and interfaces with other parts of the licensee's organization. (4) Procedures to evaluate the applicability of other national and international construction experience to the planned and ongoing construction activities and to ensure the applicable experience will be provided to those constructing the plant. (5) A fitness-for-duty program, under 10 CFR part 26. (6)(i) A QAP meeting the requirements of appendix B of part 50 of this chapter as required by § 53.460(b). (ii) Appropriate programmatic controls to provide special treatment for non-safety-related but safety-significant structures, systems, and components (SSCs). (7) A radiation protection program, in accordance with 10 CFR part 20, that includes measures for monitoring the dose to individuals working with radioactive materials brought onto the site, as applicable. (8) An information security program in accordance with §§ 73.21, 73.22, and 73.23 of this chapter, as applicable. (b) 
                                    *Construction activities.*
                                     No person may begin the construction of a commercial nuclear plant on a site on which the facility is to be operated under this part until that person has been issued either a CP or COL, an early site permit authorizing activities under § 53.1130, or an LWA under this part. (1) Licensees must satisfy the following requirements:
                                    ( print page 87058) (i) As appropriate, considering the types and quantities of radioactive materials being brought onto the site— (A) The licensee must maintain and follow a special nuclear material (SNM) material control and accounting program, a measurement control program, and other material control procedures that include corresponding record management requirements as required by the provisions of § 70.32 of this chapter. Prior to initial receipt of SNM onsite, the licensee must implement an SNM material control and accounting program in accordance with 10 CFR part 74. (B) Procedures must be in place to receive, possess, use, and store source, byproduct, and SNM in accordance with applicable portions of 10 CFR parts 30, 40, and 70. (C) A plant staff training program associated with the receipt of radioactive material must be approved and implemented prior to initial receipt of byproduct, source or SNM (excluding exempt quantities as described in § 30.18 of this chapter). (ii) For construction of a commercial nuclear plant involving multiple reactor units, plans and procedures must be in place to prevent or mitigate potential hazards to the SSCs of operating units resulting from construction activities, including the managerial and administrative controls to be used to provide assurance that the limiting conditions for operation of the operating units are not exceeded as a result of construction activities. (iii) Procedures must be in place prior to the start of construction activities that describe how construction will be controlled so as not to impact other features important to the design, such as dewatering, slope stability, backfill, compaction, and seepage. (iv) For LWA holders, a plan must be developed for redress of activities performed under the LWA should one of the following situations arise: (A) LWA work activities are terminated by the holder of the LWA; (B) The LWA is revoked by the NRC; or (C) The Commission denies the associated CP or COL application. (2)(i) Onsite fresh fuel must be protected and stored in compliance with § 73.67 of this chapter. (ii) Before initial fuel load into the reactor (or, for a fueled manufactured reactor, before initiating the physical removal of any one of the independent physical mechanisms to prevent criticality required under § 53.620(d)(1)), a cybersecurity program that meets the requirements of §§ 73.54 or 73.110 of this chapter, a physical security program that meets the requirements of §§ 73.55 or 73.100 of this chapter, and an access authorization program that meets the requirements of §§ 73.56 or 73.120 of this chapter must be established, as applicable. (iii) Fire protection measures must be implemented for work and storage areas (including adjacent fire areas that could affect the work or storage area) before initial receipt of byproduct, source, or non-fuel SNM (excluding exempt quantities as described in § 30.18 of this chapter). The fire protection measures for areas associated with new fuel (including all fuel handling, fuel storage, and adjacent fire areas that could affect the new fuel) must be implemented before receipt of fuel. Prior to the receipt of fuel, a formal letter of agreement must be in place with the local fire department specifying the nature of arrangements in support of the fire protection program. (c) 
                                    *Inspection and acceptance.*
                                     (1) The licensee must have a process for accepting individual or groups of SSCs upon completion of construction and protecting them from damage or tampering as other construction activities continue. (2) The post construction acceptance process must address the inspections, tests, analyses, and acceptance criteria specified in the COL under § 53.1440 or the equivalent verifications needed to support the issuance of an operating license under § 53.1387.
 § 53.620
  (a) 
                                    *Management and control.*
                                     Holders of MLs must ensure that the following plans, programs, and organizational units are developed and implemented to manage and control the manufacturing activities within the scope of the ML: (1) Programs to ensure that the manufacturing of a manufactured reactor or portions of a manufactured reactor complies with the design and analysis requirements in subpart C of this part. The entity with design authority for the manufactured reactor covered by the ML must be identified in the license. (2) An organizational and management structure responsible for managing, controlling, and evaluating the adequacy of the reactor design and manufacturing activities. (3) Procedures describing the qualifications for personnel in key positions in the licensee's management and control organization and the organizational responsibilities, authority, and interfaces with other parts of the licensee's organization. (4) A program to evaluate the applicability of other national and international design and manufacturing experience to the planned and ongoing manufacturing activities. (5) A fitness-for-duty program, in accordance with 10 CFR part 26. (6)(i) A QAP meeting the requirements of appendix B to part 50 of this chapter, to be applied to the design, fabrication, construction, and testing of the SSCs of the manufactured reactor. (ii) Appropriate programmatic controls to provide special treatment measures for non-safety-related but safety-significant SSCs. (7) A radiation protection program, in accordance with 10 CFR part 20, that includes measures for monitoring the dose to individuals if the manufacturing activities include working with radioactive materials. (8) An information security program in accordance with §§ 73.21, 73.22 and 73.23 of this chapter, as applicable. (b) 
                                    *Manufacturing activities.*
                                     Holders of MLs must satisfy the following requirements: (1) The manufacturing process must be conducted within facilities for which the ML holder has the authority to establish controls on any activity that might affect manufacturing. The licensee must establish access controls to the portions of each facility involved in the manufacturing processes governed by the ML. (2) Manufacturing processes must be performed in accordance with the ML and the referenced codes and standards that have been endorsed or otherwise found acceptable by the NRC. (3) A post-manufacturing inspection and acceptance process must be established and implemented before transporting a manufactured reactor or portions of a manufactured reactor for installation at a commercial nuclear plant. The process must consider the results of inspections, tests, and analyses that have been performed and the acceptance criteria that are necessary and sufficient to conclude that manufacturing activities have been completed in accordance with the ML. (c) 
                                    *Control of radioactive materials.*
                                     As appropriate considering the types and quantities of radioactive materials being brought into the manufacturing facility— (1) Procedures must be in place to receive, transfer, possess, and use source, byproduct, and SNM in accordance with the applicable portions of 10 CFR parts 30, 40 and 70. (2) A fire protection program must be established and implemented before the initial receipt of byproduct, source, or 
                                    ( print page 87059)
                                    non-fuel SNM (excluding exempt quantities as described in § 30.18 of this chapter). (3) An emergency plan appropriate for responding to the facility-specific hazards of an accidental release of radioactive material and to limit the health effects of the associated chemical hazards of licensed material must be approved and implemented prior to the receipt of byproduct, source, or SNM (excluding exempt quantities as described in § 30.18 of this chapter). (4) A plant staff training program associated with the receipt of radioactive material must be approved and implemented before initial receipt of byproduct, source, or SNM (excluding exempt quantities as described in § 30.18 of this chapter). (5) Security requirements must be implemented for the protection of SNM based on the type, enrichment, and quantity in accordance with 10 CFR part 73, as applicable, and for the protection of Category 1 and Category 2 quantities of radioactive material in accordance with 10 CFR part 37, as applicable. (d) 
                                    *Fuel loading.*
                                     (1)(i) An ML may authorize possession of a manufactured reactor into which the licensee has loaded fresh (unirradiated) fuel pursuant to a license issued under part 70 of this chapter only if the manufactured reactor is configured during its loading, storage, and transport with at least two independent physical mechanisms in place, each of which is sufficient to prevent criticality assuming optimum neutron moderation and neutron reflection conditions. (ii) The ML applicant may file a separate, subsequent application for the 10 CFR part 70 license or combine the application for the 10 CFR part 70 license with the application for an ML. (iii) The Commission has determined that any such fueled manufactured reactor in which the independent physical mechanisms to prevent criticality have been installed is not in operation. (iv) Upon installation of the fueled manufactured reactor in its place of operation and a Commission finding that the acceptance criteria in the COL that authorized reactor construction are met under § 53.1452(g), the independent physical mechanisms to prevent criticality may be removed. Upon initiating the physical removal of any one of the independent physical mechanisms to prevent criticality, the fueled manufactured reactor has commenced operation. (2) Holders of 10 CFR part 70 licenses authorizing the possession and loading of fresh fuel into manufactured reactors must comply with the requirements of 10 CFR part 70 for the facilities and activities related to the storage, movement, and loading of fresh fuel in the manufactured reactor. Holders of these 10 CFR part 70 licenses must comply with the requirements of Subpart H to 10 CFR part 70, regardless of whether their proposed activities meet the applicability criteria found in 10 CFR 70.60. Procedures, equipment, and personnel required by the 10 CFR part 70 license, must be in place before the receipt of SNM at the manufacturing facility. (i) Before the receipt of SNM, the licensee must have security programs in place that meet the performance objectives of 10 CFR 73.67, with the following additions and exceptions: (A) A physical security plan describing the physical security program must be maintained and a cybersecurity program must be established for the possession and loading of fresh fuel into a manufactured reactor authorized by a 10 CFR part 70 license, regardless of fuel type, enrichment, and quantity. (B) The physical security program must be designed to prevent unintended and uncontrolled criticality events. (C) The cybersecurity program must provide reasonable assurance that a cyberattack would not adversely impact the functions performed by digital assets used by the licensee for implementing the physical security requirements of this section, or the radiation monitoring and criticality requirements in this section or in 10 CFR part 70. (D) All holders of a part 70 license that authorizes loading of fresh fuel into a manufactured reactor must perform the screening required in § 73.67(d)(4) of this chapter to confirm the identity, trustworthiness, and reliability of individuals prior to granting unescorted access to special nuclear material; these determinations must be documented. (ii) [Reserved] (3) The loading or unloading of fresh fuel into or from a manufactured reactor and any changes to the configuration of reactivity control and prevention systems for the fueled manufactured reactor must be performed by a certified fuel handler meeting the requirements in subpart F of this part. (e) 
                                    *Transportation.*
                                     (1) A holder of an ML may not transport or allow to be removed from the places of manufacture the manufactured reactor or portions thereof as defined in the ML except for transport to a site for which the Commission has issued a COL that references the subject ML. (2) A holder of an ML must include in any contract governing the transport of a manufactured reactor or portions thereof as defined in the ML from the places of manufacture to any other location, a provision requiring that the person transporting the manufactured reactor comply with all shipping requirements in applicable NRC regulations, certificates of compliance, and NRC-issued licenses. (3) Procedures governing the preparation of the manufactured reactor or portions thereof as defined in the ML for transport and the conduct of the transport must be issued prior to transport. The procedures must implement the protective measures and restrictions described in NRC regulations and NRC-issued licenses to protect the reactor from potential conditions that would adversely affect the safe operation of a commercial nuclear plant. (4) For a manufactured reactor that is to be loaded with fresh fuel before transport to the place of operation, the ML must specify that transportation will be in accordance with parts 71 and 73 of this chapter. (f) 
                                    *Acceptance and installation at the site for which the Commission has issued a COL that references the subject ML.*
                                     (1) Installation at the site for which the Commission has issued a COL that references the subject ML must follow the regulations in § 53.610. (2) Upon arrival at the site, the manufactured reactor or portions of a manufactured reactor may not be installed in its place of operation unless the COL holder performs inspections sufficient to verify the reactor is in compliance with the ML and has not been damaged in transit. The COL holder must perform these inspections in accordance with documented procedures subject to quality assurance measures commensurate with their importance to safety. In addition, inspections must confirm that the interface requirements between the manufactured reactor or portions of a manufactured reactor and the remaining portions of the commercial nuclear plant are met.
 Subpart F—Requirements for Operation § 53.700
  (a) Each holder of an operating license (OL) or combined license (COL) under this part must develop, implement, and maintain controls for plant structures, systems, and components (SSCs), responsibilities of plant personnel, and plant programs during the operating life of each commercial nuclear plant such that the requirements defined in subpart B are satisfied. More specifically:
                                    ( print page 87060) (1) Each holder of an OL or COL under this part must maintain the capabilities, availability, and reliability of plant SSCs to ensure that the safety functions identified in § 53.230 will be performed if called upon during licensing-basis events (LBEs). (2) Each holder of an OL or COL under this part must ensure that plant personnel have adequate knowledge and skills to perform their assigned duties that support the performance of the safety functions identified in § 53.230. (3) Each holder of an OL or COL under this part must implement plant programs sufficient to ensure that the safety functions identified in § 53.230 will be performed if called upon during normal operations and LBEs. (b) [Reserved]
 § 53.710
  Controls must be provided for each commercial nuclear plant licensed under this part such that the capabilities, availability, and reliability of plant SSCs, when combined with corresponding programmatic controls and human actions, provide that the safety criteria defined in §§ 53.210 and 53.220 will be met. (a) Technical specifications must be developed, implemented, and maintained that define conditions or limitations on plant operations that are necessary to ensure that safety-related (SR) SSCs can fulfill the safety functions identified under § 53.230 and support meeting the safety criteria of § 53.210. The technical specifications must describe the following requirements: (1) Limits on the inventory of radioactive materials within the reactor system and supporting systems with the potential, individually or collectively, to cause a release exceeding the safety criteria in § 53.210 as a result of a design-basis accident analyzed in accordance with § 53.450(f). (2) Operating limits for the facility that if exceeded could lead to a failure to perform a required safety function necessary to demonstrate compliance with the safety criteria in § 53.210. (3) For each SSC classified as SR in accordance with § 53.460, technical specifications must define— (i) 
                                    *Limiting conditions for operation.*
                                     Limiting conditions for operation are the lowest functional capability or performance levels of SR SSCs required to ensure that the design-basis accidents analyzed in accordance with § 53.450(f) satisfy the safety criteria of § 53.210. When a limiting condition for operation is not met, the licensee must shut down the plant or follow any remedial action permitted by the technical specifications until the condition can be met. (ii) 
                                    *Surveillance requirements.*
                                     Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that the limiting conditions for operation will be met. (4) Design elements to be included are those elements of the plant such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (a)(1) through (3) of this section. (5) Administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the plant in a safe manner. Each licensee must submit any reports to the Commission pursuant to approved technical specifications under § 53.040. (b) Controls on plant operations, including availability controls, must be developed and implemented to ensure that the configurations and special treatments for SR SSCs and non-safety-related but safety-significant (NSRSS) SSCs provide the capabilities, availability, and reliability required to demonstrate compliance with the criteria of §§ 53.220 and 53.450(e).1
                                     The controls must— 1
                                         The comprehensive risk metrics and related risk performance objectives established under § 53.220 involve assessing and averaging the risks over a defined period (
                                        *e.g.,*
                                         plant year) and do not constitute a real-time requirement that must be continuously demonstrated by the licensee. (1)(i) Identify who within the commercial nuclear plant has authority to make configuration changes; (ii) Establish processes to make configuration changes to NSRSS SSCs; and (iii) Establish processes to ensure that all organizations of the commercial nuclear plant affected by the configuration changes are formally notified and approve of the change. (2) Describe how the special treatments for each NSRSS SSC and special treatments for SR SSCs beyond those under paragraph (a) of this section will be established and maintained over the operating life of the commercial nuclear plant.
 § 53.715
  (a) A program to control maintenance activities and monitor the performance or condition of SR and NSRSS SSCs must be developed, implemented, and maintained. (b) Whenever a licensee determines through activities related to maintenance, repair, and inspection of SSCs, the activities under § 53.710, or otherwise that the performance or condition of an SR or NSRSS SSC does not demonstrate compliance with established special treatments or performance goals related to capabilities, availability, or reliability, the licensee must take appropriate corrective action. (c) Performance and condition monitoring activities and associated goals and preventive maintenance activities must be evaluated at least every 24 months. The evaluations must take into account, where practical, industry-wide operating experience. Adjustments must be made where necessary to ensure that the objective of preventing failures of SSCs through maintenance is appropriately balanced against the objective of minimizing unavailability of SSCs due to monitoring or preventive maintenance. (d) Before performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance), the licensee must assess and manage the increase in risk that may result from the proposed maintenance activities.
 § 53.720
  If vibratory ground motion exceeding that of the operating basis earthquake ground motion or significant plant damage due to vibratory ground motion occurs, the licensee must shut down the commercial nuclear plant. If structures, systems, or components necessary for the safe shutdown of the commercial nuclear plant are not available after the occurrence of this vibratory ground motion, the licensee must consult with the Commission and must propose a plan for the timely, safe shutdown of the commercial nuclear plant. Prior to resuming operations, the licensee must demonstrate to the Commission that those features necessary for continued operation without undue risk to the health and safety of the public or necessary to maintain the licensing basis of the commercial nuclear plant were either not functionally damaged or have been repaired.
 § 53.725
  (a) 
                                    *Two classes of commercial nuclear plants.*
                                     Commercial nuclear plants licensed under this part are either of the 
                                    ( print page 87061)
                                    class, based upon the similarity of operating and technical characteristics of the plants in the class, of self-reliant-mitigation facilities or of interaction-dependent-mitigation facilities. A commercial nuclear plant is a self-reliant-mitigation facility if the U.S. Nuclear Regulatory Commission (NRC) determined as part of its approval of the OL or COL for that plant that its design demonstrates compliance with the criteria of § 53.800(a)(1) through (a)(5). Otherwise, the commercial nuclear plant is an interaction-dependent-mitigation facility. (b) 
                                    *Purpose and applicability.*
                                     The regulations in §§ 53.725 through 53.830 address areas related to staffing, training, personnel qualifications, and human factors engineering for applicants for or holders of OLs or COLs under this part. These regulations are organized as follows: (1) Sections 53.725 through 53.745 address general requirements for staffing, training, personnel qualifications, and human factors engineering. The regulations within these sections are applicable to all applicants for or holders of OLs or COLs under this part, except where specifically stated otherwise. (2) Sections 53.760 through 53.795 address operator and senior operator licensing requirements. The regulations within these sections are applicable to those applicants for or holders of OLs or COLs under this part for interaction-dependent-mitigation facilities that have not yet certified the permanent cessation of operations and permanent removal of fuel from the reactor vessel as described under § 53.1070. (3) Sections 53.800 through 53.820 address generally licensed reactor operator requirements. The regulations within these sections are in lieu of §§ 53.760 through 53.795 for those applicants for or holders of OLs or COLs under this part for self-reliant-mitigation facilities that have not yet certified the permanent cessation of operations and permanent removal of fuel from the reactor vessel as described under § 53.1070. (4) Section 53.830 provides general personnel training requirements. The regulations within this section are applicable to all applicants for or holders of OLs or COLs under this part. (c) 
                                    *Definitions.*
                                     When used in §§ 53.725 through 53.830: *Applicant*
                                     refers to an applicant for an operator or senior operator license; 
                                    *licensee*
                                     refers to the holder of an operator, senior operator, or generally licensed reactor operator license; and 
                                    *facility licensee*
                                     refers to the licensee for the commercial nuclear plant where the applicant would be licensed or the licensee is licensed. *Automation*
                                     means a device or system that accomplishes (partially or fully) a function or task. *Auxiliary operator*
                                     means any individual who operates components of a commercial nuclear plant but does not manipulate controls or direct the manipulation of controls of the plant and is not required to be licensed under the provisions of this part. *Controls*
                                     when used with respect to a nuclear reactor means apparatus and mechanisms, the manipulation of which directly affects the reactivity or power level of the reactor. *Generally licensed reactor operator*
                                     means any individual licensed under the provisions of § 53.810 to manipulate controls of a self-reliant-mitigation facility and to direct the licensed activities of generally licensed reactor operators. *Interaction-dependent-mitigation facility*
                                     means a commercial nuclear plant design other than one that demonstrates compliance with the operating and technical characteristics defined under § 53.800. *Load following*
                                     means a commercial nuclear plant automatically changing its output to match expected demand in response to externally originated instructions or signals. *Operator*
                                     means any individual licensed under the provisions of §§ 53.760 through 53.795 to manipulate controls of an interaction-dependent-mitigation facility. *Performance testing*
                                     means testing conducted to verify a simulation facility's performance as compared to actual or predicted reference plant performance. *Reference plant*
                                     means the specific commercial nuclear plant on which a simulation facility's configuration, system control arrangement, and design data are based. The reference plant may or may not be constructed. *Self-reliant-mitigation facility*
                                     means a commercial nuclear plant design that demonstrates compliance with the operating and technical characteristics defined under § 53.800. *Senior operator*
                                     means any individual licensed under the provisions of §§ 53.760 through 53.795 to manipulate controls of an interaction-dependent-mitigation facility and to direct the licensed activities of operators. *Simulation facility*
                                     means an interface designed to provide a realistic imitation of the operation of a commercial nuclear plant used for the administration of examinations, for training, and/or to demonstrate compliance with experience requirements for applicants or licensees. A simulation facility may rely, in whole or part, upon the physical utilization of the reference plant itself. *Systems approach to training*
                                     means a training program that includes the following five elements: (1) Systematic analysis of the jobs to be performed. (2) Learning objectives derived from the analysis which describe desired performance after training. (3) Training design and implementation based on the learning objectives. (4) Evaluation of trainee mastery of the objectives during training. (5) Evaluation and revision of the training based on the performance of trained personnel in the job setting.
 § 53.726
  (a) An applicant or licensee or facility licensee must submit any communication or report required by the regulations contained within §§ 53.725 through 53.830 and must submit any application filed under these regulations to the Commission. (b) Each licensee that is required to comply with the requirements of §§ 53.760 through 53.795 (
                                    *i.e.,*
                                     interaction-dependent-mitigation facilities) must notify the appropriate NRC contact within 30 days of the following in regard to a licensed operator or senior operator: (1) Permanent reassignment from the position for which the licensee has certified the need for a licensed operator or senior operator under § 53.775(a)(1); (2) Termination of any operator or senior operator; or (3) Permanent disability or illness as required under § 55.770 of this chapter.
 § 53.728
  Information provided to the Commission by an applicant for an operator or senior operator license or by a licensee or information required by statute or by the Commission's regulations, orders, or license conditions to be maintained by the applicant or the licensee must be complete and accurate in all material respects.
 § 53.730
  Each applicant for or holder of an OL or COL for a commercial nuclear plant under this part must comply with the following: (a) 
                                    *Human factors engineering design requirements.*
                                     The plant design must reflect state-of-the-art human factors engineering principles for safe and reliable performance in all locations that 
                                    ( print page 87062)
                                    human activities are expected for performing or supporting the continued availability of plant safety or emergency response functions. (b) 
                                    *Human system interface design requirements.*
                                     The plant design must provide for the following to support operating personnel in monitoring plant conditions and responding to plant events: (1) Features for displaying to operating personnel a minimum set of parameters that define the safety status of the plant and are capable of displaying both the full range of important plant parameters and data trends on demand, as well as indicating when process limits are being approached or exceeded; (2) Automatic indication of the bypassed and operable status of safety systems; (3) Direct indication of SSC status that relates to the ability of the SSC to perform its safety function, such as relief and safety valve position (
                                    *i.e.,*
                                     open or closed) for barriers important to fulfilling safety functions of with such devices, and ultimate heat sink and cooling system status and availability; (4) Instrumentation to measure, record, and display key plant parameters related to the performance of SSCs and the integrity of barriers important to fulfilling safety functions to support operators in monitoring plant conditions and responding to plant events. Examples include temperatures and pressures within important systems or structures, core or fuel system conditions (including possible damage states), temperatures and levels associated with cooling functions, combustible gas concentrations, radiation levels in systems and within structures, and radioactive effluent releases; (5) Leakage control and detection in the design of systems that pass through barriers important to fulfilling safety functions for the release of radionuclides. An example is an SSC that penetrates a containment structure that might contain radioactive materials that could contribute to the source term during an accident; (6) Monitoring of in-plant radiation and airborne radioactivity as appropriate for a broad range of normal operating and accident conditions; and (7) For self-reliant-mitigation facilities, the plant design must also provide the generally licensed reactor operators with the capability to do the following: (i) Receive plant operating data, including reactor parameters and information needed for the evaluation of emergency conditions. (ii) Immediately initiate a reactor shutdown from their location. (iii) Promptly dispatch operations and maintenance personnel. (iv) Immediately implement responsibilities under the facility emergency plan, as applicable. (c) 
                                    *Concept of operations.*
                                     A concept of operations that is of sufficient scope and detail to address the following must be provided: (1) Plant goals; (2) The roles and responsibilities of operating personnel and automation (or any combination thereof) that are responsible for completing plant functions; (3) Staffing, qualifications, and training; (4) The management of normal operations; (5) The management of off-normal conditions and emergencies; (6) The management of maintenance and modifications; and (7) The management of tests, inspections, and surveillances. (d) 
                                    *Functional requirements analysis and function allocation.*
                                     A functional requirements analysis and a function allocation must be provided that are sufficient to demonstrate compliance with the following: (1) The functional requirements analysis must address how safety functions and functional safety criteria are satisfied, and (2) The function allocation must describe how the safety functions will be assigned to human action, automation, active safety features, passive safety features, and/or inherent safety characteristics. (e) 
                                    *Operating experience.*
                                     A program, during construction and during operation, as applicable, for evaluating and applying operating experience must be developed, implemented, and maintained. (f) 
                                    *Staffing plan.*
                                     A staffing plan must be developed and comply with the following: (1) The staffing plan must include a description of how engineering expertise will be available to the on-shift operating personnel during all plant conditions, to assist if they encounter a situation not covered by procedures or training. Engineering expertise includes familiarity with the operation of the plant for which the expertise is provided and one of the following: (i) A bachelor's degree in engineering, engineering technology, or physical science from an institution accredited by a U.S. government recognized accrediting body or equivalent; or (ii) A Professional Engineer's license from a U.S. State or territory. (2) Applicants for or holders of OLs or COLs for interaction-dependent-mitigation facilities must include within their staffing plans a description of how the proposed numbers, positions, and qualifications of operators and senior operators across all modes of plant operations will be sufficient to ensure that plant safety functions will be maintained. This description must be supported by human factors engineering analyses and assessments. (3) Applicants for or holders of OLs or COLs for self-reliant-mitigation facilities must include within their staffing plans a description of how generally licensed reactor operator staffing that is both sufficient to continually monitor the operations of fueled reactors and to provide for a continuity of responsibility for facility operations at all times during the operating phase will be maintained. (4) Applicants for or holders of OLs or COLs under this part must include within their staffing plans a description of how the numbers, positions, and responsibilities of personnel contained within those plans will adequately support all necessary functions within areas such as plant operations, equipment surveillance and maintenance, radiological protection, chemistry control, fire brigades, engineering, security, and emergency response. (5) The staffing plan must be approved by the NRC as part of its approval of the OL or COL for the plant. The approved staffing plan is subject to the requirements of § 53.1565. (g) 
                                    *Training, examination, and proficiency programs.*
                                     Develop, implement, and maintain programs that comply with the following requirements. These programs must be approved by the NRC as part of its approval of the OL or COL for the plant: (1) For those applicants for or holders of OLs or COLs for interaction-dependent-mitigation facilities: (i) The operator licensing initial training program required under § 53.780(a); (ii) The operator licensing initial examination program required under § 53.780(b); (iii) The operator licensing requalification program required under § 53.780(c); and (iv) The operator proficiency program required under § 53.780(g). (2) For those applicants for or holders of OLs or COLs for self-reliant-mitigation facilities, the generally licensed reactor operator training, examination, and proficiency programs required under § 53.815.
                                    ( print page 87063) (3) The operator licensing requalification programs required under § 53.780(c) or § 53.815(b) must be implemented upon commencing the administration of initial examinations under the operator licensing examination program required under § 53.780(b) or § 53.815(b), respectively.
 § 53.735
  The regulations in §§ 53.725 through 53.830 do not require a license for an individual who— (a) Under the direction and in the presence of an operator or senior operator or a generally licensed reactor operator, as appropriate, manipulates the controls of a commercial nuclear plant as a part of the individual's training in a facility licensee's training program as approved by the Commission to qualify for an operator or senior operator license or a generally licensed reactor operator license there, as appropriate, under these regulations; or (b) Under the direction and in the presence of a senior operator or generally licensed reactor operator, as appropriate, manipulates the controls of a commercial nuclear plant to load or unload the fuel into, out of, or within the reactor vessel while the reactor is not operating.
 § 53.740
  (a) Facility licensees must demonstrate compliance with the requirements of either §§ 53.760 through 53.795 for interaction-dependent-mitigation facilities or §§ 53.800 through 53.820 for self-reliant-mitigation facilities. (b) The facility licensee must maintain the staffing complement described under its approved facility staffing plan until such time as the permanent cessation of operations and permanent removal of fuel from the reactor vessel has been certified as described under § 53.1070. The approved staffing plan is subject to the requirements of § 53.1565. (c) Except as provided under § 53.735, the facility licensee may not permit the manipulation of the controls of a commercial nuclear plant by anyone who is not an operator or senior operator or generally licensed reactor operator, as appropriate. (d) Facility licensees for interaction-dependent-mitigation facilities that have not yet certified the permanent cessation of operations and permanent removal of fuel from the reactor vessel as described under § 53.1070 must designate senior operators to be responsible for supervising the licensed activities of operators. (e) Apparatus and mechanisms other than controls, the operation of which may affect the reactivity or power level of a reactor, must be manipulated only while plant conditions are being monitored by an individual who is an operator or senior operator or a generally licensed reactor operator, as appropriate. (f)(1) Load following is permitted if at least one of the following is immediately capable of refusing demands when they could challenge the safe operation of the plant or when precluded by the plant equipment conditions: (i) The actuation of an automatic protection system that utilizes setpoints more conservative than those otherwise credited for the purposes of reactor protection; or (ii) An automated control system; or (iii) An operator or senior operator or a generally licensed reactor operator, as appropriate. (2) The provisions of paragraph (e) of this section do not apply during load following operations. (g)(1) Facility licensees for interaction-dependent-mitigation facilities must have present during alteration of the core (including fuel loading or transfer) an individual holding a senior operator license, or a senior operator license limited to fuel handling to directly supervise the activity and, during this time, the facility licensee must not assign other duties to this person. (2) Facility licensees for self-reliant-mitigation facilities must have present during alteration of the core (including fuel loading or transfer) an individual holding a generally licensed reactor operator license to directly supervise the activity and, during this time, the facility licensee must not assign other duties to this person. (3) The provisions of paragraphs (g)(1) and (2) of this section do not apply to core alterations performed as part of refueling operations while a facility that is capable of online refueling is operating at power. (h) Facility licensees may take reasonable action that departs from a license condition or a technical specification (contained in a license issued under this part) in an emergency when this action is immediately needed to protect the public health and safety and no action consistent with license conditions and technical specifications that can provide adequate or equivalent protection is immediately apparent. Such facility licensee action must be approved, as a minimum, by a senior operator or a generally licensed reactor operator, as applicable, or, after certifying the permanent cessation of operations and permanent removal of fuel from the reactor vessel as described under § 53.1070 by a certified fuel handler, senior operator, or generally licensed reactor operator, as applicable, prior to taking the action.
 § 53.745
  A person must be authorized by a license issued by the Commission to perform the function of an operator, senior operator, or generally licensed reactor operator as defined in this part.
 § 53.760
  (a) 
                                    *Applicability.*
                                     Sections 53.760 through 53.795 address operator and senior operator licensing requirements. The regulations within these sections are applicable to those applicants for or holders of OLs or COLs under this part for interaction-dependent-mitigation facilities that have not yet certified the permanent cessation of operations and permanent removal of fuel from the reactor vessel as described under § 53.1070. (b) Reserved.
 § 53.765
  (a) An applicant for an operator or senior operator license must have a medical examination by a physician. An operator or senior operator must have a medical examination by a physician every 2 years. (b) To certify the medical fitness of an applicant for an operator or senior operator license, an authorized representative of the facility licensee must complete and sign NRC Form 396, “Certification of Medical Examination by Facility Licensee,” which can be obtained by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, by calling 301-415-7232, or by visiting the NRC's website at 
                                    *https://www.nrc.gov*
                                     and selecting forms from the index found on the home page, or by other means provided by the NRC. (1) Form NRC 396 must certify that a physician has conducted the medical examination of the applicant as required in paragraph (a) of this section. (2) When the medical certification requests a conditional license based on medical evidence, the medical evidence must be submitted on NRC Form 396 to the Commission to enable the Commission to make a determination in accordance with § 53.775(b). (c) The facility licensee must document and maintain the results of medical qualifications data, test results, 
                                    ( print page 87064)
                                    and each operator's or senior operator's medical history for the current license period and provide the documentation to the Commission upon request. The facility licensee must retain this documentation while an individual performs the functions of an operator or senior operator.
 § 53.770
  If, during the term of the operator or senior operator license, the licensee develops a permanent physical or mental condition that causes the licensee to fail to demonstrate compliance with the requirements of § 53.775(b)(1)(i), the facility licensee must notify the Commission within 30 days of learning of the diagnosis. For conditions for which a conditional license (as described in § 53.775(b)) is requested, the facility licensee must provide medical certification on Form NRC 396 to the Commission (as described in § 53.765(b)).
 § 53.775
  (a) 
                                    *How to apply.*
                                     (1) The applicant for an operator or senior operator license must— (i) Complete NRC Form 398, “Personal Qualification Statement— Licensee,” which can be obtained by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, by calling 301-415-5877, or by visiting the NRC's website at 
                                    *https://www.nrc.gov*
                                     and selecting forms from the index found on the home page, or by other means provided by the NRC; (ii) File an original of NRC Form 398, or an equivalent electronic submittal, together with the information required in paragraphs (a)(1)(iii) and (a)(1)(iv) of this section, with the appropriate Regional Administrator. (iii) Provide evidence that the applicant, as a trainee, has successfully demonstrated competence in manipulating the controls of either the facility for which a license is sought or a simulation facility that demonstrates compliance with the requirements of § 53.780(e). For operators applying for a senior operator license, certification that the operator has successfully operated the controls of the facility as an operator will be accepted; and (iv) Provide certification by the facility licensee of medical condition and general health on Form NRC 396, to comply with § 53.765. (2) The Commission may at any time after the application has been filed, and before the license has expired, require further information under oath or affirmation to enable it to determine whether to grant or deny the application or whether to revoke, modify, or suspend the license. (3) An applicant whose application has been denied because of a medical condition or their general health may submit a further medical report at any time as a supplement to the application. (4) Each application and statement must contain complete and accurate disclosure as to all matters required to be disclosed. The applicant must sign statements required by paragraphs (a)(1)(i) and (a)(1)(ii) of this section. (b) 
                                    *Disposition of an initial application.*
                                     (1) 
                                    *License approval.*
                                     The Commission will approve an initial application if it finds that the following criteria are met: (i) 
                                    *Health.*
                                     The applicant's medical condition and general health will not adversely affect the performance of assigned operator or senior operator job duties or cause operational errors endangering public health and safety. The Commission will base its finding upon the certification by the facility licensee as detailed in § 53.765(b). (ii) 
                                    *Examination.*
                                     The applicant has passed the requisite examination in accordance with § 53.780(b). The examination determines whether the applicant for an operator's or senior operator's license has learned to operate a facility competently and safely, and additionally, in the case of a senior operator, whether the applicant has learned to supervise the licensed activities of operators competently and safely. (2) 
                                    *Conditional license.*
                                     If an applicant's general medical condition does not demonstrate compliance with the minimum standards under § 53.775(b)(1)(i) of this section, the Commission may approve the application and include conditions in the license to accommodate the medical condition. The Commission will consider the recommendations and supporting evidence of the facility licensee and of the examining physician (provided on Form NRC 396) in arriving at its decision. (c) 
                                    *Re-applications.*
                                     (1) An applicant whose application for a license has been denied because of failure to pass the examination may file a new application. The application must be submitted on Form NRC 398 and include a statement signed by an authorized representative of the facility licensee by whom the applicant will be employed that states in detail the extent of the applicant's additional training and remediation since the denial and certifies that the applicant is ready for re-examination. (2) An applicant who has passed a portion of the examination and failed another may request in a new application on Form NRC 398 to be excused from re-examination on the portions of the examination that the applicant has passed. The Commission may in its discretion grant the request if it determines that sufficient justification is presented.
 § 53.780
  (a) 
                                    *Operator licensing initial training program.*
                                     (1) A program that is based upon a systems approach to training, as defined by § 53.725(b), must be utilized for the training of applicants for operator and senior operator licenses. The program must ensure that applicants at the facility will possess the knowledge, skills, and abilities necessary to protect the public health and maintain those plant safety functions specific to the facility design. The program must be approved by the Commission prior to its use for training applicants, as described under § 53.730(g). The approved operator licensing initial training program is subject to the requirements of § 53.1565. (2) The facility licensee must maintain operator licensing initial training program records documenting the initial operator licensing training administered and completed by each applicant. The facility licensee must retain these records during the period in which any trainees subsequently remain licensed as operators or senior operators at the facility. (b) 
                                    *Operator licensing initial examination program.*
                                     (1) The facility licensee must establish and implement an examination program for testing a representative sample of the knowledge, skills, and abilities needed to safely perform operator and senior operator duties, to include both the examination methods and criteria to be used to assess passing performance. The program must provide for valid and reliable examinations and be approved by the Commission prior to its use for examining applicants, as described under § 53.730(g). The approved operator licensing initial examination program is subject to the requirements of § 53.1565. (2) The facility licensee must submit prepared examinations to the Commission for review and approval in advance of their administration. (3) The Commission will either administer an approved examination or allow the facility licensee to administer the examination. The facility licensee must ensure that sufficient advance notification is provided to the 
                                    ( print page 87065)
                                    Commission to either administer the examination or allow for a representative of the Commission to be afforded the opportunity to be present when the facility licensee administers the examination. (4) Graded examination documentation for each applicant must be promptly provided to the Commission for review in making operator licensing decisions. (5) The facility licensee must maintain operator licensing initial examination program records documenting the participation of each operator and senior operator applicant in the initial examination. The records must contain copies of examinations administered, the answers given by the applicant, documentation of the grading of examinations, and documentation of any additional training administered in areas in which an applicant exhibited deficiencies. The facility licensee must retain these records during the period in which the associated operators or senior operators remain licensed at the facility. (c) 
                                    *Operator licensing requalification program.*
                                     (1) A program based upon a systems approach to training, as defined by § 53.725(b), must be utilized for the continuing training of operators and senior operators. (i) The program must ensure that operators and senior operators at the facility maintain the knowledge, skills, and abilities necessary to protect the public health and maintain those plant safety functions specific to the facility design. The program must be conducted for a continuous period not to exceed 24 months in duration. (ii) The program must be approved by the Commission prior to its use for continuing training, as described under § 53.730(g). The approved operator licensing requalification program is subject to the requirements of § 53.1565. (2) The following requirements apply to operator licensing requalification programs: (i) The facility licensee must propose a requalification examination program for testing, for each requalification period, a sample of the topics included under the systems approach to training, to include both the examination methods and criteria to be used to assess passing performance. The program must provide for valid and reliable examinations and be approved by the Commission prior to its use for examining operators and senior operators, as described under § 53.730(g). The approved requalification examination program is subject to the requirements of § 53.1565. (ii) The following requirements apply to the requalification examination program: (A) The facility licensee must make prepared requalification examinations available to the Commission for review. (B) The facility licensee must ensure that a representative of the Commission is afforded the opportunity to be present during requalification examination administration. (C) The facility licensee must ensure that each operator and senior operator is administered a complete requalification examination on a periodicity not to exceed 24 months. Additionally, the facility licensee must ensure that any licensed operator or senior licensed operator who either demonstrates unsatisfactory performance on, or fails to complete, the biennial requalification examination is removed from the performance of licensed operator and senior licensed operator duties until such time that any necessary remedial training has been completed and a retake examination has been passed. (D) The facility licensee must promptly provide a summary of examination results for each operator and senior operator following the completion of the requalification examination. (3) The facility licensee must maintain operator licensing requalification program records documenting the participation of each operator and senior operator in the requalification program. The records must contain copies of examinations administered, the answers given by the operator or senior operator, documentation of the grading of examinations, and documentation of any additional training administered in areas in which an operator or senior operator exhibited deficiencies. The facility licensee must retain these records until the operator's or senior operator's license is renewed. (d) 
                                    *Examination integrity.*
                                     Applicants, operators and senior operators, and facility licensees must not engage in any activity that compromises the integrity of any application or examination required by §§ 53.760 through 53.795. The integrity of an examination is considered compromised if any activity, regardless of intent, affected, or, but for detection, could have affected the equitable and consistent administration of the examination. This includes activities related to the preparation and certification of applications and all activities related to the preparation, administration, and grading of examinations required by §§ 53.760 through 53.795. (e) 
                                    *Simulation facilities.*
                                     (1) This section addresses the use of a simulation facility for the administration of examinations, for training, or to demonstrate compliance with experience requirements for applicants for operator and senior operator licenses. (2) Simulation facilities used for training purposes, for demonstrating compliance with experience requirements, or for the conduct of examinations under § 53.780(b) and (c) must demonstrate compliance with the following criteria as they relate to the facility licensee's reference plant: (i) The simulation facility must be of sufficient scope and fidelity for individuals to acquire and demonstrate the necessary knowledge, skills, and abilities to safely perform operator and senior operator duties. (ii) The simulation facility must utilize models relating to nuclear, thermal-hydraulic, and other applicable design-specific characteristics that either replicate the most recent fuel load in the reference commercial nuclear plant or, prior to initial fuel load (or, for a fueled manufactured reactor, prior to initiating the physical removal of any one of the independent physical mechanisms to prevent criticality required under § 53.620(d)(1)), replicate the intended initial fuel load for the reference commercial nuclear plant, with the exception of those portions of the simulation facility that utilize the reference plant itself. (iii) Simulation facility fidelity must be demonstrated so that significant control manipulations are completed without procedural exceptions, simulator performance exceptions, or deviation from the approved training scenario sequence. (3) Facility licensees that maintain a simulation facility that has been approved by the Commission for training purposes, demonstrating compliance with experience requirements, or the conduct of examinations under § 53.780(b) and (c) for the facility licensee's reference plant must: (i) Conduct performance testing throughout the life of the simulation facility in a manner sufficient to ensure that paragraph (e)(2) of this section is met; (ii) Retain the results of performance testing for 4 years after the completion of each performance test or until superseded by updated test results; (iii) Promptly correct modeling and hardware discrepancies and discrepancies identified from scenario validation and from performance testing or provide justification as to why the 
                                    ( print page 87066)
                                    presence of such discrepancies will not adversely affect simulator performance with respect to the criteria of paragraph (e)(2) of this section; (iv) Make the results of any uncorrected performance test failures that may exist at the time of the initial license examination or requalification examination available for NRC review, prior to or concurrent with preparations for each initial license examination or requalification examination; and (v) Maintain the provisions for license application and examination integrity consistent with § 53.780(d). (4) A simulation facility must demonstrate compliance with the requirements of paragraphs (e)(2) and (e)(3) of this section for the Commission to accept the simulation facility for conducting initial examinations as described in § 53.780(b), requalification training as described in § 53.780(c), or performing control manipulations that affect reactivity to establish eligibility for an operator or senior operator license as described in § 53.775(a). (f) 
                                    *Waiver of examination requirement.*
                                     On application, the Commission may waive any or all of the requirements for an examination if it finds that the applicant has demonstrated the required knowledge, skills, and abilities to safely operate the plant, and is capable of continuing to do so. The Commission may make such a finding based on demonstration of the following: (1) Operating experience at a comparable facility; (2) Proof of the applicant's past competent and safe performance; and (3) Proof of the applicant's current qualifications. (g) 
                                    *Proficiency.*
                                     The facility licensee must develop, implement, and maintain a proficiency program to ensure that operators and senior operators will actively perform the functions of an operator or senior operator, respectively, as needed to maintain proficiency with on-shift duties and familiarity with plant status. This program must include those steps that will be taken to re-establish proficiency when it cannot be maintained. This program must be approved by the Commission as part of its approval of the OL or COL for the plant. The approved proficiency program is subject to the requirements of § 53.1565. (h) 
                                    *Records.*
                                     Each record required by this section must be legible throughout the retention period specified by each Commission regulation. The record may be the original, a reproduced copy, or an electronic copy provided that the copy is authenticated by authorized personnel.
 § 53.785
  Each operator and senior operator license contains and is subject to the following conditions whether stated in the license or not: (a) Neither the license nor any right under the license may be assigned or otherwise transferred. (b) The license is limited to the facility for which it is issued. (c) The license is limited to those controls of the facility or facilities specified in the license. (d) The license is subject to, and the licensee must observe, all applicable rules, regulations, and orders of the Commission. (e) The licensee must maintain or re-establish proficiency in accordance with the facility licensee's Commission-approved proficiency program required under § 53.780(g). (f) The licensee must be subject to the facility's Commission-approved operator licensing requalification and requalification examination programs required under § 53.780(c). (g) The licensee must have a biennial medical examination as described by § 53.765. (h) The licensee must notify the Commission within 30 days about a conviction for a felony. (i) The licensee must not consume or ingest alcoholic beverages within the protected area of commercial nuclear plants. The licensee must not use, possess, or sell any illegal drugs. The licensee must not perform activities authorized by a license issued under this part while under the influence of alcohol or any prescription, over-the-counter, or illegal substance that could adversely affect his or her ability to safely and competently perform his or her licensed duties. For the purpose of this paragraph, with respect to alcoholic beverages and drugs, the term “under the influence” means the licensee exceeded, as evidenced by a confirmed test result, the lower of the cutoff levels for drugs or alcohol contained in 10 CFR part 26, or as established by the facility licensee. The term “under the influence” also means the licensee could be mentally or physically impaired as a result of substance use including prescription and over-the-counter drugs, as determined under the provisions, policies, and procedures established by the facility licensee for its fitness-for-duty program, in such a manner as to adversely affect his or her ability to safely and competently perform licensed duties. (j) Each licensee must participate in the drug and alcohol testing programs as required under 10 CFR part 26. (k) The licensee must comply with any other conditions that the Commission may impose to protect health or to minimize danger to life or property.
 § 53.790
  (a) 
                                    *Issuance of operator and senior operator licenses.*
                                     If the Commission determines that an applicant for an operator license or a senior operator license demonstrates compliance with the requirements of the Atomic Energy Act of 1954, as amended, (the Act) and its regulations, it will issue a license in the form and containing any conditions and limitations it considers appropriate and necessary. (b) 
                                    *Modification and revocation of operator and senior operator licenses.*
                                     (1) The terms and conditions of all operator and senior operator licenses are subject to amendment, revision, or modification by reason of rules, regulations, or orders issued in accordance with the Act or any amendments thereto. (2) Any license may be revoked, suspended, or modified, in whole or in part— (i) For any material false statement in the application or in any statement of fact required under section 182 of the Act; (ii) Because of conditions revealed by the application or statement of fact or any report, record, inspection, or other means that would warrant the Commission to refuse to grant a license on an original application; (iii) For willful violation of, or failure to observe, any of the terms and conditions of the Act or the license, or of any rule, regulation, or order of the Commission; (iv) For any conduct determined by the Commission to be a hazard to safe operation of the facility; or (v) For the sale, use, or possession of illegal drugs, or refusal to participate in the facility drug and alcohol testing program, or a confirmed positive test for drugs, drug metabolites, or alcohol in violation of the conditions and cutoff levels established by § 53.785(i) or the consumption of alcoholic beverages within the protected area of commercial nuclear plants, or a determination of unfitness for scheduled work as a result of the consumption of alcoholic beverages.
 § 53.795
  (a) 
                                    *Expiration.*
                                     (1) Each operator license and senior operator license expires 6 years after the date of 
                                    ( print page 87067)
                                    issuance, upon termination of employment with the facility licensee, or upon determination by the facility licensee that the licensed individual no longer needs to maintain a license. (2) If a licensee files an application for renewal or an upgrade of an existing license on Form NRC 398 at least 30 days before the expiration of the existing license, it does not expire until disposition of the application for renewal or for an upgraded license has been finally determined by the Commission. Filing by mail will be deemed to be complete at the time the application is postmarked (b) 
                                    *Renewal.*
                                     (1) The applicant for renewal of an operator license or senior operator license must— (i) Complete and sign Form NRC 398 and include the number of the license for which renewal is sought. (ii) File an original of NRC Form 398 as specified in § 53.775. (iii) Provide written evidence of the applicant's experience under the existing license and the approximate number of hours that the licensee has operated the facility. (iv) Provide a statement by an authorized representative of the facility licensee that during the effective term of the current license the applicant has satisfactorily completed the requalification program for the facility for which operator or senior operator license renewal is sought. (v) Provide evidence that the applicant has discharged the license responsibilities competently and safely. The Commission may accept as evidence of the applicant's having met this requirement a certificate of an authorized representative of the facility licensee or holder of an authorization by which the licensee has been employed. (vi) Provide certification by the facility licensee of medical condition and general health on Form NRC 396, to comply with § 53.765. (2) The license will be renewed if the Commission finds that— (i) The medical condition and the general health of the licensee continue to be such as not to cause operational errors that endanger public health and safety. The Commission will base this finding upon the certification by the facility licensee as described in § 53.765(b). (ii) The licensee— (A) Is capable of continuing to competently and safely assume licensed duties; (B) Has successfully completed a requalification program that has been approved by the Commission as required by § 53.780(c); and (C) Has passed the requalification examinations as required by § 53.780(c). (iii) There is a continued need for an operator to operate or for a senior operator to supervise operators at the facility designated in the application. (iv) The past performance of the licensee has been satisfactory to the Commission. In making its finding, the Commission will include in its evaluation information such as notices of violations or letters of reprimand in the licensee's docket.
 § 53.800
  (a) A commercial nuclear plant is a self-reliant-mitigation facility if the NRC determined as part of its approval of the OL or COL for that plant that its design demonstrates compliance with criteria (a)(1) though (a)(5) of this section. A self-reliant-mitigation facility is of a class, based upon the similarity of operating and technical characteristics of the plants in the class, such that its licensee must comply with the requirements of §§ 53.800 through 53.820 in lieu of those in §§ 53.760 through 53.795. (1) The safety performance criteria of §§ 53.210 and 53.220 and, if applicable, any alternative criteria used in accordance with § 53.470, must be met without reliance upon human action for credited event mitigation. (2) The results of a probabilistic risk analysis must demonstrate that the evaluation criteria for the events analyzed in accordance with § 53.450 will be met without reliance on human actions to achieve acceptable event mitigation. (3) The functional requirements analysis and function allocation performed under § 53.730(d) must demonstrate that functions required for safety are not reliant upon credited human action. (4) The plant response to events analyzed under § 53.450 must rely exclusively on safety features and characteristics that will neither be rendered unavailable by credible human errors of commission or omission nor credibly require manual human operation in response to equipment failures. Compliance with this paragraph may be achieved through the use of SSCs that function through inherent characteristics or that have engineered protections against human failures. (5) The plant design must provide for a layered defense-in-depth approach that is not dependent upon any single barrier or credited human action. (b) [Reserved]
 § 53.805
  (a) Licensees for self-reliant-mitigation facilities that have not yet certified the permanent cessation of operations and permanent removal of fuel from the reactor vessel as described under § 53.1070 must demonstrate compliance with the following requirements: (1) Ensure that, in addition to being qualified to perform those items identified by the facility-specific systems approach to training conducted under § 53.815, generally licensed reactor operators are qualified to safely and competently— (i) Perform administrative tasks, including compliance with technical specifications, and perform operability determinations; (ii) Implement maintenance and configuration controls; (iii) Comply with radioactive release limitations; (iv) Understand plant operating data, including reactor parameters, and evaluate emergency conditions; (v) Initiate a reactor shutdown from necessary locations; (vi) Dispatch and direct operations and maintenance personnel; (vii) Implement any applicable responsibilities under the facility emergency plan; and (viii) Make required notifications to local, State, participating Tribal and Federal authorities. (2) Develop, implement, and maintain facility technical specifications that provide the necessary administrative controls to ensure the implementation of these requirements. (3) Develop, implement, and maintain the generally licensed reactor operator training, examination, and proficiency programs required under § 53.815. (4) Ensure that generally licensed reactor operators are subject to the facility's generally licensed reactor operator training, examination, and proficiency programs required under § 53.815. Ensure that generally licensed reactor operators are subject to and comply with the applicable programmatic requirements for plant personnel required under 10 CFR parts 26 and 73. An individual that is not in compliance with any of these programs is not qualified to be in a position that may involve the manipulation of the controls of the commercial nuclear plant. (5) Report annually to the NRC the identity of all generally licensed reactor operators at the commercial nuclear plant, including all additions and deletions since the previous report.
                                    ( print page 87068) (6) Ensure that the facility design continues to meet the criteria of § 53.800. (b) [Reserved]
 § 53.810
  (a) A general license to manipulate the controls of a self-reliant-mitigation facility and to direct the licensed activities of generally licensed reactor operators is hereby issued to any individual employed in a position that may involve the manipulation of the controls of that self-reliant-mitigation facility and who observes the restrictions of this section. (b) A generally licensed reactor operator must comply with the operating procedures and other conditions specified in the license authorizing operation of the facility. (c) The general license is limited to the facility or facilities at which the operator is employed. (d) The Commission will suspend the general license on an individual basis for violations of any provision of the Act or any rule or regulation issued thereunder whenever the Commission deems such suspension desirable, including— (1) For willful violation of, or failure to observe, any of the terms and conditions of the Act or the general license, or of any rule, regulation, or order of the Commission; (2) For any conduct determined by the Commission to be a hazard to safe operation of the facility; or (3) For the sale, use, or possession of illegal drugs, or refusal to participate in the facility drug and alcohol testing program, or a confirmed positive test for drugs, drug metabolites, or alcohol in violation of the conditions and cutoff levels established by § 53.810(f) or the consumption of alcoholic beverages within the protected area of commercial nuclear plants, or a determination of unfitness for scheduled work as a result of the consumption of alcoholic beverages. (e) The Commission may require information from a generally licensed reactor operator to determine whether a general license should be revoked or suspended with respect to that operator. (f) The generally licensed reactor operator must not consume or ingest alcoholic beverages within the protected area of commercial nuclear plants. The generally licensed reactor operator must not use, possess, or sell any illegal drugs. The generally licensed reactor operator must not perform activities requiring a general license while under the influence of alcohol or any prescription, over-the-counter, or illegal substance that could adversely affect his or her ability to safely and competently perform these activities. For the purpose of this paragraph, with respect to alcoholic beverages and drugs, the term “under the influence” means the generally licensed reactor operator exceeded, as evidenced by a confirmed test result, the lower of the cutoff levels for drugs or alcohol contained in 10 CFR part 26, or as established by the facility licensee. The term “under the influence” also means the generally licensed reactor operator could be mentally or physically impaired as a result of substance use including prescription and over-the-counter drugs, as determined under the provisions, policies, and procedures established by the facility licensee for its fitness-for-duty program, in such a manner as to adversely affect his or her ability to safely and competently perform generally licensed reactor operator duties. (g) The generally licensed reactor operator must notify the Commission within 30 days about a conviction for a felony.
 § 53.815
  (a) 
                                    *Applicability.*
                                     The requirements of this section apply to each licensee of a self-reliant-mitigation facility that has not yet certified the permanent cessation of operations and permanent removal of fuel from the reactor vessel as described under § 53.1070. (b) 
                                    *Requirements.*
                                     (1) The licensee must develop, implement, and maintain training and examination programs that demonstrate compliance with the requirements of paragraphs (b)(2) and (3) of this section. (2) The training program must provide for both the initial and continuing training of generally licensed reactor operators and be derived from a systems approach to training as defined in this part. (3)(i) The training program must incorporate the instructional requirements necessary to provide qualified generally licensed reactor operators to operate and maintain the facility in a safe manner in all modes of operation. The training program must comply with the facility license, including all technical specifications and applicable regulations. The facility licensee must periodically evaluate and revise the training program as appropriate to reflect industry experience and relevant changes, including changes to the facility, procedures, regulations, and quality assurance (QA) requirements. Facility licensee management must periodically review the training program for effectiveness. (ii) The training program must ensure that generally licensed reactor operators have and maintain the necessary knowledge, skills, and abilities. (iii) The training program must include the generally licensed reactor operator manipulating the controls of either the facility or a simulation facility that demonstrates compliance with the requirements of § 53.815(e). (iv) The training program must include an initial examination program for testing a representative sample of the knowledge, skills, and abilities needed to safely perform generally licensed reactor operator duties, to include both the examination methods and criteria to be used to assess passing performance. The facility licensee must provide the opportunity for a representative of the Commission to be present during initial examination administration. (v) The training program must include a requalification examination program for testing a sample of the topics included under the systems approach to training, to include the examination methods and criteria to be used to assess passing performance. The requalification examination program must specify an appropriate periodicity for administering a complete requalification examination to each generally licensed reactor operator, and the facility licensee must provide the opportunity for a representative of the Commission to be present during requalification examination administration. (A) The facility licensee must ensure that any generally licensed reactor operator who either demonstrates unsatisfactory performance on, or fails to complete, the requalification examination is removed from the performance of generally licensed reactor operator duties until such time that any necessary remedial training has been completed and a retake examination has been passed. (B) [Reserved] (vi) The training program must be approved by the Commission prior to its use, as described under § 53.730(g). The examination program must provide for valid and reliable examinations and must be approved by the Commission prior to their use, as described under § 53.730(g). The approved programs are subject to the requirements of § 53.1565. (c) 
                                    *Records.*
                                     The following is required regarding the documentation of the generally licensed reactor operator training and examination programs: (1) Sufficient records must be maintained by the facility licensee to 
                                    ( print page 87069)
                                    maintain the integrity of the programs and kept available for NRC inspection to verify the adequacy of the programs. (2) The facility licensee must maintain records documenting the participation of each generally licensed reactor operator in the training and examination programs. The records must contain copies of examinations administered, the answers given by the generally licensed reactor operator, documentation of the grading of examinations, and documentation of any additional training administered in areas in which a generally licensed reactor operator exhibited deficiencies. The facility licensee must retain these records while the associated generally licensed reactor operators remain employed at the facility. (3) Each record required by this part must be legible throughout the retention period. The record may be the original, a reproduced copy, or an electronic copy provided that the copy is authenticated by authorized personnel. (d) 
                                    *Examination integrity.*
                                     Generally licensed reactor operators and facility licensees must not engage in any activity that compromises the integrity of any examination conducted under the generally licensed reactor operator training and examination programs. The integrity of an examination is considered compromised if any activity, regardless of intent, affected, or, but for detection, could have affected the equitable and consistent administration of the examination. This includes all activities related to the preparation, administration, and grading of examinations. (e) 
                                    *Simulation facilities.*
                                     (1) Simulation facilities used for training purposes, for maintaining proficiency, or for the conduct of examinations must demonstrate compliance with the following criteria as they relate to the facility licensee's reference plant: (i) The simulation facility must be of sufficient scope and fidelity for individuals to acquire and demonstrate the necessary knowledge, skills, and abilities to safely perform generally licensed reactor operator duties. (ii) The simulation facility must utilize models relating to nuclear, thermal-hydraulic, and other applicable design-specific characteristics that either replicate the most recent fuel load in the reference commercial nuclear plant or, prior to initial fuel load (or, for a fueled manufactured reactor, prior to initiating the physical removal of any one of the independent physical mechanisms to prevent criticality required under § 53.620(d)(1)), replicate the intended initial fuel load for the reference commercial nuclear plant, with the exception of those portions of the simulation facility that utilize the reference plant itself. (iii) Simulator fidelity must be demonstrated so that significant control manipulations are completed without procedural exceptions, simulator performance exceptions, or deviation from the approved training scenario sequence. (2) Facility licensees that maintain a simulation facility for training purposes, for maintaining proficiency, or for the conduct of examinations must— (i) Conduct performance testing throughout the life of the simulation facility in a manner sufficient to ensure that paragraph (e)(1) of this section is met; (ii) Retain the results of performance testing for 4 years after the completion of each performance test or until superseded by updated test results; (iii) Promptly correct modeling and hardware discrepancies and discrepancies identified from scenario validation and from performance testing or provide justification for why the presence of such discrepancies will not adversely affect the criteria of paragraph (e)(1) of this section; (iv) Make the results of any uncorrected performance test failures that may exist at the time of an inspection available for NRC review; and (v) Maintain the provisions for examination integrity consistent with § 53.815(d). (f) 
                                    *Waiver of examination requirement.*
                                     The facility licensee may waive any or all of the requirements for an examination in accordance with the facility licensee's Commission-approved generally licensed reactor operator training and examination programs. (g) 
                                    *Proficiency.*
                                     The facility licensee must develop, implement, and maintain a proficiency program to allow generally licensed reactor operators to maintain proficiency regarding position functions and familiarity with plant status. This program must include those steps that will be taken in order to re-establish proficiency when it cannot be maintained.
 § 53.820
  The general license ceases to be applicable on an individual basis once a generally licensed reactor operator is no longer being employed in a position that may involve the manipulation of the controls of the self-reliant mitigation facility.
 § 53.830
  (a) This section addresses personnel training requirements. The regulations within this section are applicable to all applicants for or holders of OLs or COLs under this part. (b) Prior to initial fuel load (or, for a fueled manufactured reactor, prior to initiating the physical removal of any one of the independent physical mechanisms to prevent criticality required under § 53.620(d)(1)), each holder of an operating or COL under this part must, with sufficient time to provide trained and qualified personnel to operate the facility, establish, implement, and maintain a training program that demonstrates compliance with the requirements of paragraphs (c) and (d) of this section. (c) The training program must be derived from a systems approach to training as defined in this part and must provide, at a minimum, for the training and qualification of the following categories of commercial nuclear plant personnel: (1) Supervisors (
                                    *e.g.,*
                                     shift supervisors); (2) Technicians (
                                    *e.g.,*
                                     maintenance, chemistry, and radiological); and (3) Other appropriate operating personnel (
                                    *e.g.,*
                                     auxiliary operators, certified fuel handlers, and individuals who provide engineering expertise to on-shift operating personnel). (d) The training program must incorporate the instructional requirements necessary to provide qualified personnel to operate components of a commercial nuclear plant and maintain the facility in a safe manner in all modes of operation. The training program must be developed to be in compliance with the facility license, including all technical specifications and applicable regulations. (1) The training program must be periodically evaluated and revised as appropriate to reflect industry experience and relevant changes, including changes to the facility, procedures, regulations, and QA requirements. The training program must be periodically reviewed by facility licensee management for effectiveness. (2) Sufficient records must be maintained by the facility licensee to maintain program integrity and kept available for NRC inspection to verify the adequacy of the training program.
 § 53.845
  (a) The required plant programs under this part must include but are not necessarily limited to the programs 
                                    ( print page 87070)
                                    described in the following sections of this subpart. Licensees may combine, separate, and otherwise organize programs and related documents as appropriate for the technologies and organizations associated with the commercial nuclear plant. (b) In addition to the programs described in the following sections, programs must be provided for each commercial nuclear plant, if necessary, to ensure that the performance of design features and human actions are consistent with the analyses performed under §§ 53.450 and 53.730 and that the plant will demonstrate compliance with the safety criteria defined in §§ 53.210 and 53.220.
 § 53.850
  (a) Each holder of an OL or COL under this part must develop, implement, and maintain a Radiation Protection Program for operations that is commensurate with the scope and extent of licensed activities under this part and includes measures for limiting and monitoring radioactive plant effluents and limiting and monitoring the dose to individuals working with radioactive materials in accordance with 10 CFR part 20. (b) Each holder of an OL or COL under this part must develop, implement, and maintain a program for the control of radioactive effluents and for keeping the doses to members of the public from radioactive effluents as low as is reasonably achievable and for environmental monitoring. The program must be contained in an Offsite Dose Calculations Manual, must be implemented by procedures, and must include remedial actions to be taken whenever the program limits are exceeded. The Offsite Dose Calculations Manual must— (1) Contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and (2) Contain the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by § 53.1645. (c) Each holder of an OL or COL under this part must develop, implement, and maintain a Process Control Program that identifies the administrative and operational controls for solid radioactive waste processing, process parameters, and surveillance requirements sufficient to ensure compliance with the requirements of 10 CFR part 20, 10 CFR part 61, and 10 CFR part 71.
 § 53.855
  (a) Each holder of an OL or COL under this part must have an emergency response plan that must contain information needed to demonstrate compliance with either the requirements in § 50.160 of this chapter or the requirements in appendix E to part 50 and the planning standards of § 50.47(b) of this chapter. (b) No initial OL, initial COL, or early site permit that includes complete and integrated emergency plans will be issued under this part unless a finding is made by the NRC, in accordance with § 50.47 of this chapter, that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency.
 § 53.860
  (a) 
                                    *Physical Protection Program.*
                                     Each holder of an OL or COL under this part must develop, implement, and maintain a physical protection program under the following requirements: (1) The licensee must implement security requirements for the protection of special nuclear material based on the type, enrichment, and quantity in accordance with 10 CFR part 73, as applicable, and implement security requirements for the protection of Category 1 and Category 2 quantities of radioactive material in accordance with 10 CFR part 37, as applicable; and (2) The licensee must demonstrate compliance with the provisions set forth in either §§ 73.55 or 73.100 of this chapter, unless the licensee demonstrates compliance with the following criterion: (i) The radiological consequences from a design-basis threat-initiated event involving the loss of engineered systems for decay heat removal and possible breaches in physical structures surrounding the reactor, spent fuel, and other inventories of radioactive materials result in offsite doses below the values in § 53.210. (ii) The applicant must perform a site-specific analysis, including identification of target sets, to demonstrate that the criterion in § 53.860(a)(2)(i) is satisfied. The analysis must assume that licensee mitigation and recovery actions, including any operator actions, are unavailable or ineffective. The licensee must maintain the analysis until the permanent cessation of operations and permanent removal of fuel from the reactor vessel as described under § 53.1070. (b) 
                                    *Fitness for Duty.*
                                     Each holder of an OL or COL under this part must develop, implement, and maintain a fitness for duty program under 10 CFR part 26. (c) 
                                    *Access Authorization.*
                                     Each holder of an OL or COL under this part must develop, implement, and maintain an access authorization program under § 73.120 of this chapter if the criterion in § 53.860(a)(2)(i) is satisfied, or the requirements in § 73.56 of this chapter if the criterion is not satisfied. (d) 
                                    *Cybersecurity.*
                                     Each holder of an OL or COL under this part must develop, implement, and maintain a cybersecurity program under §§ 73.54 or 73.110 of this chapter. (e) 
                                    *Information Security.*
                                     Each holder of an OL or COL under this part must develop, implement, and maintain an information protection system under §§ 73.21, 73.22, and 73.23 of this chapter, as applicable.
 § 53.865
  Each holder of an OL or COL under this part must develop, implement, and maintain a quality assurance program in accordance with appendix B of part 50 of this chapter. A written quality assurance program manual must be developed and used to guide the conduct of the program in accordance with generally accepted consensus codes and standards that have been endorsed or otherwise found acceptable by the NRC.
 § 53.870
  Each holder of an OL or COL under this part must develop, implement, and maintain an integrity assessment program to monitor, evaluate, and manage— (a) The effects of plant aging on SR and NSRSS SSCs. The program may refer to surveillances, tests, and inspections conducted for specific SSCs in accordance with other requirements in this part or conducted in accordance with applicable consensus codes and standards endorsed or otherwise found acceptable by the NRC; (b) Cyclic or transient load limits to ensure that SR and NSRSS SSCs are maintained within the applicable design limits; and (c) Degradation mechanisms related to chemical interactions, operating temperatures, effects of irradiation, and other environmental factors to ensure that the capabilities, availability, and reliability of SR and NSRSS SSCs demonstrate compliance with the 
                                    ( print page 87071)
                                    functional design criteria of §§ 53.410 and 53.420.
 § 53.875
  (a)(1) Each holder of an OL or COL under this part must have a fire protection plan that describes the overall fire protection program for the facility; identifies the various positions within the licensee's organization that are responsible for the program; states the authorities that are delegated to each of these positions to implement those responsibilities; and outlines the plans for fire protection, fire detection and suppression capability; and limitation of fire damage. (2) The fire protection plan must also describe specific features necessary to implement the program described in paragraph (a)(1) of this section such as the following: administrative controls and personnel requirements for fire prevention and manual fire suppression activities; automatic and manually operated fire detection and suppression systems; and the means to limit fire damage to SSCs so that the capability to demonstrate compliance with the requirements of § 53.210 is ensured. (b)(1) Each holder of an OL or COL under this part must develop a performance-based or deterministic fire protection program that demonstrates compliance with the safety criteria outlined in §§ 53.210 and 53.220, related safety functions outlined in § 53.230, and defense in depth as outlined in § 53.250 with specific fire protection measures related to fire prevention, fire detection, and fire suppression. (2) The fire protection program must comply with the following: (i) Safety-related and NSRSS SSCs must be designed, located, and maintained to minimize, consistent with other safety requirements, the probability and effect of fires and explosions. (ii) Noncombustible and fire-resistant materials must be used wherever practical throughout the facility, particularly in locations with SR and NSRSS SSCs. (iii) Fire detection and fire suppression systems of appropriate capacity and capability must be provided and designed and maintained to minimize the adverse effects of fires on SR and NSRSS SSCs. (iv) Fire suppression systems must be designed and maintained to ensure that their rupture or inadvertent operation does not significantly impair the ability of SR and NSRSS SSCs to perform their safety functions to satisfy § 53.230.
 § 53.880
  (a) Each holder of an OL or COL under this part must develop, implement, and maintain a program for inservice inspection (ISI) and inservice testing (IST) prior to receiving an OL or COL. The ISI/IST programs must, wherever applicable, be in accordance with generally accepted consensus codes and standards that have been endorsed or otherwise found acceptable by the NRC. The ISI/IST program must include all inspections and tests required by the codes and standards used in the design and be supplemented by risk insights that identify the most important SSCs to plant safety. The types of testing and inspections and their frequency should be informed by risk insights to maintain the reliability and performance of SSCs consistent with the associated design and analyses activities involving those SSCs. Risk insights must also be used to determine when to conduct the inspections and tests (
                                    *e.g.,*
                                     full power, shutdown, refueling) to minimize risk to the plant workers and the public. The ISI/IST program must be documented in a written manual and managed by qualified personnel reporting to the Plant Manager. (b) Prior to plant operation, baseline inspections and testing must be performed using the same techniques as will be used for future inspections and testing. The results of these inspections and testing must be used as benchmarks for evaluating the results of future inspections and testing. Sufficient room and support must be provided to accommodate the personnel, ISI/IST equipment, and shielding necessary to perform the inspections and testing. Acceptance criteria for determining whether corrective action is needed must be developed (or taken from the codes and standards used in the design) for evaluating the results of the inspections and testing. The results of the inspections and testing must be provided to the Plant Manager who is responsible for determining what, if any, corrective action is needed and when it should be taken. The ISI/IST results and corrective actions must be documented and the documentation retained for the life of the plant.
 § 53.910
  (a) Each holder of an OL or COL under this part must have a program for developing, implementing, and maintaining an integrated set of procedures, guidelines, and related supporting activities to support normal operations and respond to possible unplanned events. (b) The program required by paragraph (a) of this section must include but is not limited to development, implementation, maintenance, and supporting activities of procedures and guidelines for the following: (1) Plant operations; (2) Maintenance activities under § 53.715; (3) Program requirements under this subpart F of this part; (4) Emergency operating procedures, if developed to address the role of human actions in responding to LBEs; (5) Accident management guidelines, if developed to address the role of human actions in responding to LBEs; (6) Procedures for each area in which licensed special nuclear material is handled, used, or stored to protect personnel upon the sounding of a criticality alarm required by § 53.440(m); and (7) Procedures that describe how the licensee will address the following areas if the licensee is notified of a potential aircraft threat: (i) Verification of the authenticity of threat notifications; (ii) Maintenance of continuous communication with threat notification sources; (iii) Contacting all onsite personnel and applicable offsite response organizations; (iv) Onsite actions necessary to enhance the capability of the facility to mitigate the consequences of an aircraft impact; (v) Measures to reduce visual discrimination of the site relative to its surroundings or individual buildings within the protected area; (vi) Dispersal of equipment and personnel, as well as rapid entry into site protected areas for essential onsite personnel and offsite responders who are necessary to mitigate the event; and (vii) Recall of site personnel.
 Subpart G—Decommissioning Requirements § 53.1000
  This subpart defines the requirements related to decommissioning for applicants for, or holders of, an operating license (OL) or combined license (COL). The requirements related to maintaining financial assurance for decommissioning are in §§ 53.1010 through 53.1060. The requirements for transitioning from operations to decommissioning and for the release of property and termination of the license are in §§ 53.1070 through 53.1080.
 § 53.1010
  (a) This section establishes requirements for indicating to the U.S. Nuclear Regulatory Commission (NRC) how an applicant for or holder of an OL or COL under this part will provide reasonable assurance that funds will be available for the decommissioning process. Reasonable assurance consists of a series of steps as provided in paragraph (b) of this section and §§ 53.1020, 53.1030 and 53.1040. Funding for the decommissioning of commercial nuclear plants may also be subject to the regulation of Federal or State government agencies (
                                    *e.g.,*
                                     Federal Energy Regulatory Commission (FERC) and State Public Utility Commissions) that have jurisdiction over rate regulation. The requirements of this subpart, in particular § 53.1020, are in addition to, and not a substitution for, other requirements, and are not intended to be used by themselves or by other agencies to establish rates. (b) Each applicant for an OL or COL under this part must prepare a plan and an associated decommissioning report that ensures and documents that adequate funding will be available to decommission the facility. Each holder of an OL or COL must implement and maintain the plan. (1)(i) Before the Commission issues an OL under this part, the applicant must update the decommissioning report to certify that it has provided financial assurance for decommissioning in the amount proposed in the application and approved by the NRC under § 53.1020. (ii) No later than 30 days after the Commission issues the notice of intended operation under § 53.1452 for a COL under this part, the licensee must update the decommissioning report to certify that it has provided financial assurance for decommissioning in the amount proposed in the application and approved by the NRC under § 53.1020. (2) The amount of financial assurance for decommissioning to be provided must be based on a site-specific cost estimate for decommissioning the facility under § 53.1020. (3) The amount of financial assurance for decommissioning to be provided must be adjusted annually using a rate at least equal to that stated in § 53.1030. (4) The amount of financial assurance for decommissioning to be provided must be covered by one or more of the methods described in § 53.1040 as acceptable to the NRC. A copy of the financial instrument obtained to satisfy the requirements of § 53.1040 must be submitted to the NRC as part of the application for an OL under this part; however, an applicant for or holder of a COL need not obtain such financial instrument or submit a copy to the Commission except as provided in § 53.1060(b).
 § 53.1020
  Cost estimates for decommissioning must be site-specific. Site-specific decommissioning cost estimates (DCEs) must account for the engineering, labor, equipment, transportation, disposal, and related charges needed to support termination of the license. They must include the costs for decontaminating structures, systems, and components and the site environs; removal of contaminated components and materials from the plant and the site environs; disposal of removed components and materials in appropriate facilities; and any other activities supporting the release of the property and termination of the license. They must also address the approach to annual adjustments required by § 53.1030. Finally, site-specific DCEs must include plans for adjusting levels of funds assured for decommissioning to demonstrate that a reasonable level of assurance will be provided that funds will be available when needed to cover the cost of decommissioning.
 § 53.1030
  Each holder of an OL or COL under this part must annually adjust the cost estimate for decommissioning to account for escalation in labor, energy, and waste burial costs. Licensees may elect to use either a site-specific adjustment factor, approved as part of the plan and associated decommissioning report required by § 53.1010, in paragraph (a) of this section or the generic adjustment factor in paragraph (b) of this section. (a) A site-specific adjustment factor must address the estimated contributions and escalation of costs for the following aspects of decommissioning: (1) Labor, materials, and services; (2) Energy and waste transportation; and (3) Radioactive waste burial or other disposition. (b) A generic adjustment factor must be at least equal to 0.65 L + 0.13 E + 0.22 B, where L and E are escalation factors for labor and energy, respectively, and are to be taken from regional data of U.S. Department of Labor Bureau of Labor Statistics and B is an escalation factor for waste burial and is to be taken from NRC report NUREG-1307, “Report on Waste Burial Charges.”
 § 53.1040
  Financial assurance for decommissioning is to be provided by the following methods. (a) 
                                    *Prepayment.*
                                     Prepayment is the deposit made preceding the start of operation or the transfer of a license under § 53.1570 into an account segregated from licensee assets and outside the administrative control of the licensee and its subsidiaries or affiliates of cash or liquid assets such that the amount of funds would be sufficient to pay decommissioning costs. Prepayment may be in the form of a trust, escrow account, or Government fund with payment by certificate of deposit, deposit of government or other securities, or other method acceptable to the NRC. This trust, escrow account, Government fund, or other type of agreement must be established in writing and maintained at all times in the United States with an entity that is an appropriate State or Federal Government agency, or an entity whose operations in which the prepayment deposit is managed are regulated and examined by a Federal or State agency. A licensee that has prepaid funds based on a site-specific cost estimate under § 53.1020 may take credit for projected earnings on the prepaid decommissioning trust funds, using up to a 2 percent annual real rate of return through the time of termination of the license. A licensee may use a credit of greater than 2 percent if the licensee's rate-setting authority has specifically authorized a higher rate. Actual earnings on existing funds may be used to calculate future fund needs. (b) 
                                    *External sinking fund.*
                                     An external sinking fund is a fund established and maintained by setting funds aside periodically in an account segregated from licensee assets and outside the administrative control of the licensee and its subsidiaries or affiliates in which the total amount of funds would be sufficient to pay decommissioning costs. An external sinking fund may be in the form of a trust, escrow account, or Government fund, with payment by certificate of deposit, deposit of Government or other securities, or other method acceptable to the NRC. This trust, escrow account, Government fund, or other type of agreement must be established in writing and maintained at all times in the United States with an entity that is an appropriate State or Federal Government agency, or an entity whose operations in which the external sinking fund is managed are regulated and examined by a Federal or State agency. A licensee that has collected funds based on a site-specific cost 
                                    ( print page 87073)
                                    estimate under § 53.1020 may take credit for projected earnings on the external sinking funds using up to a 2 percent annual real rate of return from the time of future funds' collection through the time of termination of the license. A licensee may use a credit of greater than 2 percent if the licensee's rate-setting authority has specifically authorized a higher rate. Actual earnings on existing funds may be used to calculate future fund needs. A licensee whose rates for decommissioning costs cover only a portion of these costs may make use of this method only for the portion of these costs that are collected in one of the manners described in this paragraph. This method may be used as the exclusive mechanism relied upon for providing financial assurance for decommissioning in the following circumstances: (1) By a licensee that recovers, either directly or indirectly, the estimated total cost of decommissioning through rates established by “cost of service” or similar ratemaking regulation. Public utility districts, municipalities, rural electric cooperatives, and State and Federal agencies, including associations of any of the foregoing, that establish their own rates and are able to recover their cost of service allocable to decommissioning, are deemed to satisfy this condition. (2) By a licensee whose source of revenues for its external sinking fund is a “non-bypassable charge,” the total amount of which will provide funds estimated to be needed for decommissioning pursuant to §§ 53.1020, 53.1060, or 53.1575. (c) 
                                    *A surety method, insurance, or other guarantee method.*
                                     (1) These methods guarantee that decommissioning costs will be paid. A surety method may be in the form of a surety bond, or letter of credit. Any surety method or insurance used to provide financial assurance for decommissioning must contain the following conditions: (i) The surety method or insurance must be open-ended, or, if written for a specified term, such as 5 years, must be renewed automatically, unless 90 days or more prior to the renewal day the issuer notifies the NRC, the beneficiary, and the licensee of its intention not to renew. The surety or insurance must also provide that the full-face amount be paid to the beneficiary automatically prior to the expiration without proof of forfeiture if the licensee fails to provide a replacement acceptable to the NRC within 30 days after receipt of notification of cancellation. (ii) The surety or insurance must be payable to a trust established for decommissioning costs. The trustee and trust must be acceptable to the NRC. An acceptable trustee includes an appropriate State or Federal Government agency or an entity that has the authority to act as a trustee and whose trust operations are regulated and examined by a Federal or State agency. (2) A parent company guarantee of funds for decommissioning costs based on a financial test may be used if the guarantee and test are as contained in appendix A to 10 CFR part 30. (3) For commercial companies that issue bonds, a guarantee of funds by the applicant or licensee for decommissioning costs based on a financial test may be used if the guarantee and test are as contained in appendix C to 10 CFR part 30. For commercial companies that do not issue bonds, a guarantee of funds by the applicant or licensee for decommissioning costs may be used if the guarantee and test are as contained in appendix D to 10 CFR part 30. A guarantee by the applicant or licensee may not be used in any situation in which the applicant or licensee has a parent company holding majority control of voting stock of the company. (d) 
                                    *Funding method for Federal licensees.*
                                     For a Federal licensee, a statement of intent containing a cost estimate for decommissioning and indicating that funds for decommissioning will be obtained when necessary. (e) 
                                    *Contractual funding method.*
                                     Contractual obligation(s) on the part of a licensee's customer(s), the total amount of which over the duration of the contract(s) will provide the licensee's total share of uncollected funds estimated to be needed for decommissioning pursuant to §§ 53.1020, 53.1060, or 53.1575. To be acceptable to the NRC as a method of decommissioning funding assurance, the terms of the contract(s) must include provisions that the buyer(s) of electricity or other products will pay for the decommissioning obligations specified in the contract(s), notwithstanding the operational status either of the licensed plant to which the contract(s) pertains or force majeure provisions. All proceeds from the contract(s) for decommissioning funding will be deposited to the external sinking fund. The NRC reserves the right to evaluate the terms of any contract(s) and the financial qualifications of the contracting entity or entities offered as assurance for decommissioning funding. (f) 
                                    *Other funding mechanisms.*
                                     Any other mechanism, or combination of mechanisms, that provides, as determined by the NRC upon its evaluation of the specific circumstances of each licensee submittal, assurance of decommissioning funding equivalent to that provided by the mechanisms specified in paragraphs (a) through (e) of this section. Licensees who do not have sources of funding described in paragraph (b) of this section may use an external sinking fund in combination with a guarantee mechanism, as specified in paragraph (c) of this section, provided that the total amount of funds estimated to be necessary for decommissioning is assured.
 § 53.1045
  (a)(1) Decommissioning trust funds may be used by licensees if— (i) The withdrawals are for expenses for decommissioning activities consistent with the definition of decommission or decommissioning in § 53.020; (ii) The expenditure would not reduce the value of the decommissioning trust below an amount necessary to place and maintain the reactor in a safe storage condition if unforeseen conditions or expenses arise; and (iii) The withdrawals would not inhibit the ability of the licensee to complete funding of any shortfalls in the decommissioning trust needed to ensure the availability of funds to ultimately release the site and terminate the license. (2) Initially, 3 percent of the amount determined in accordance with § 53.1020 may be used for decommissioning planning. For licensees that have submitted the certifications required under § 53.1070 and commencing 90 days after the NRC has received the post-shutdown decommissioning activities report (PSDAR) required by § 53.1060, an additional 20 percent may be used. An updated site-specific DCE must be submitted to the NRC prior to the licensee using any funding in excess of these amounts. (b) Licensees that are not “electric utilities” as defined in § 53.020 that use prepayment or an external sinking fund to provide financial assurance must provide in the terms of the arrangements governing the trust, escrow account, or Government fund, used to segregate and manage the funds that— (1) The trustee, manager, investment advisor, or other person directing investment of the funds— (i) Is prohibited from investing the funds in securities or other obligations of the licensee or any other owner or operator of any commercial nuclear 
                                    ( print page 87074)
                                    plant or their affiliates, subsidiaries, successors or assigns, or in a mutual fund in which at least 50 percent of the fund is invested in the securities of a licensee or parent company whose subsidiary is an owner or operator of a foreign or domestic commercial nuclear plant. However, the funds may be invested in securities tied to market indices or other non-nuclear sector collective, commingled, or mutual funds, provided that no more than 10 percent of trust assets may be indirectly invested in securities of any entity owning or operating one or more commercial nuclear plants. (ii) Is obligated at all times to adhere to a standard of care set forth in the trust, which either shall be the standard of care, whether in investing or otherwise, required by State or Federal law or one or more State or Federal regulatory agencies with jurisdiction over the trust funds, or, in the absence of any such standard of care, whether in investing or otherwise, that a prudent investor would use in the same circumstances. The term “prudent investor,” shall have the same meaning as set forth in FERC's “Regulations Governing Nuclear Plant Decommissioning Trust Funds” at 18 CFR 35.32(a)(3), or any successor regulation. (2) The licensee, its affiliates, and its subsidiaries are prohibited from being engaged as investment manager for the funds or from giving day-to-day management direction of the funds' investments or direction on individual investments by the funds, except in the case of passive fund management of trust funds where management is limited to investments tracking market indices. (3) The trust, escrow account, Government fund, or other account used to segregate and manage the funds may not be amended in any material respect without written notification to the Director, Office of Nuclear Reactor Regulation, or Director, Office of Nuclear Material Safety and Safeguards, as applicable, at least 30 working days before the proposed effective date of the amendment. The licensee must provide the text of the proposed amendment and a statement of the reason for the proposed amendment. The trust, escrow account, Government fund, or other account may not be amended if the person responsible for managing the trust, escrow account, Government fund, or other account receives written notice of objection from the Director, Office of Nuclear Reactor Regulation, or Director, Office of Nuclear Material Safety and Safeguards, as applicable, within the notice period. (4) Except for withdrawals being made under paragraph (a) of this section or for payments of ordinary administrative costs (including taxes) and other incidental expenses of the fund (including legal, accounting, actuarial, and trustee expenses) in connection with the operation of the fund, no disbursement or payment may be made from the trust, escrow account, Government fund, or other account used to segregate and manage the funds until written notice of the intention to make a disbursement or payment has been given to the Director, Office of Nuclear Reactor Regulation, or Director, Office of Nuclear Material Safety and Safeguards, as applicable, at least 30 working days before the date of the intended disbursement or payment. The disbursement or payment from the trust, escrow account, Government fund or other account may be made following the 30 working day notice period if the person responsible for managing the trust, escrow account, Government fund, or other account does not receive written notice of objection from the Director, Office of Nuclear Reactor Regulation, or Director, Office of Nuclear Material Safety and Safeguards, as applicable, within the notice period. Disbursements or payments from the trust, escrow account, Government fund, or other account used to segregate and manage the funds, other than for payment of ordinary administrative costs (including taxes) and other incidental expenses of the fund (including legal, accounting, actuarial, and trustee expenses) in connection with the operation of the fund, are restricted to decommissioning expenses or transfer to another financial assurance method acceptable under § 53.1040 until final decommissioning has been completed. After decommissioning has begun and withdrawals from the decommissioning fund are made under paragraph (a) of this section, no further notification need be made to the NRC. (c) Licensees that are “electric utilities” under § 53.020 that use prepayment or an external sinking fund to provide financial assurance must include a provision in the terms of the trust, escrow account, Government fund, or other account used to segregate and manage funds that except for withdrawals being made under paragraph (a) of this section or for payments of ordinary administrative costs (including taxes) and other incidental expenses of the fund (including legal, accounting, actuarial, and trustee expenses) in connection with the operation of the fund, no disbursement or payment may be made from the trust, escrow account, Government fund, or other account used to segregate and manage the funds until written notice of the intention to make a disbursement or payment has been given the Director, Office of Nuclear Reactor Regulation, or Director, Office of Nuclear Material Safety and Safeguards, as applicable, at least 30 working days before the date of the intended disbursement or payment. The disbursement or payment from the trust, escrow account, Government fund or other account may be made following the 30 working day notice period if the person responsible for managing the trust, escrow account, Government fund, or other account does not receive written notice of objection from the Director, Office of Nuclear Reactor Regulation, or Director, Office of Nuclear Material Safety and Safeguards, as applicable, within the notice period. Disbursements or payments from the trust, escrow account, Government fund, or other account used to segregate and manage the funds, other than for payment of ordinary administrative costs (including taxes) and other incidental expenses of the fund (including legal, accounting, actuarial, and trustee expenses) in connection with the operation of the fund, are restricted to decommissioning expenses or transfer to another financial assurance method acceptable under § 53.1040 until final decommissioning has been completed. After decommissioning has begun and withdrawals from the decommissioning fund are made under paragraph (a) of this section, no further notification need be made to the NRC. (d) A licensee that is not an “electric utility” under § 53.020 and using a surety method, insurance, or other guarantee method to provide financial assurance must provide that the trust established for decommissioning costs to which the surety or insurance is payable contains in its terms the requirements in § 53.1045(b)(1) through (4).
 § 53.1050
  The NRC reserves the right to take the following steps in order to ensure a licensee's adequate accumulation of decommissioning funds: review, as needed, the rate of accumulation of decommissioning funds and, either independently or in cooperation with FERC and the licensee's State Public Utility Commission, take additional actions as appropriate on a case-by-case basis, including modification of a licensee's schedule for the accumulation of decommissioning funds.
 § 53.1060
  (a) Each holder of an OL under this part or holder of a COL under this part after the date that the Commission has made the finding under § 53.1452(g) must report, at least once every 2 years, by March 31, on the status of its certification of decommissioning funding for each commercial nuclear reactor or part of a commercial nuclear reactor that it owns. The information in this report must include, at a minimum, the amount of decommissioning funds estimated to be required under §§ 53.1020 and 53.1030; the amount of decommissioning funds accumulated to the end of the calendar year preceding the date of the report; a schedule of the annual amounts remaining to be collected; the assumptions used regarding rates of escalation in decommissioning costs, rates of earnings on decommissioning funds, and rates of other factors used in funding projections; any contracts upon which the licensee is relying under § 53.1040(e); any modifications occurring to a licensee's method of providing financial assurance since the last submitted report; and any material changes to trust agreements. If any of the preceding items is not applicable, the licensee should so state in its report. Any licensee for a plant that is within 5 years of the projected end of its operation, or where conditions have changed such that it will close within 5 years (before the end of its licensed life), or that has already closed (before the end of its licensed life), or that is involved in a merger or an acquisition must submit this report annually. (b) Each holder of a COL under this part must, 2 years before and 1 year before the scheduled date for initial loading of fuel (or, for a fueled manufactured reactor, 2 years before and 1 year before the scheduled date for initiating the physical removal of any one of the independent physical mechanisms to prevent criticality required under § 53.620(d)(1)), submit a report to the NRC containing a certification updating the DCEs and a copy of the financial instrument to be used to satisfy § 53.1040. No later than 30 days after the Commission publishes notice in the 
                                    **Federal Register**
                                     under § 53.1452(a), the licensee must submit an updated decommissioning report required under § 53.1010(b)(1)(ii), including a copy of the financial instrument obtained to satisfy § 53.1040. (c) Each licensee must keep records of information important to the safe and effective decommissioning of the facility in an identified location until the license is terminated by the Commission. If records of relevant information are kept for other purposes, reference to these records and their locations may be used. Information the Commission considers important to decommissioning consists of— (1) Records of spills or other unusual occurrences involving the spread of contamination in and around the facility, equipment, or site. These records may be limited to instances when significant contamination remains after any cleanup procedures or when there is reasonable likelihood that contaminants may have spread to inaccessible areas as in the case of possible seepage into porous materials such as concrete. These records must include any known information on identification of involved nuclides, quantities, forms, and concentrations. (2) As-built drawings and modifications of structures and equipment in restricted areas where radioactive materials are used and/or stored and of locations of possible inaccessible contamination such as buried pipes that may be subject to contamination. If required drawings are referenced, each relevant document need not be indexed individually. If drawings are not available, the licensee must substitute appropriate records of available information concerning these areas and locations. (3) Records of the cost estimate performed for the decommissioning funding plan or of the amount certified for decommissioning, and records of the funding method used for assuring funds if either a funding plan or certification is used. (4) Records of— (i) The licensed site area, as originally licensed and any revisions, which must include a site map and any acquisition or use of property outside the originally licensed site area for the purpose of receiving, possessing, or using licensed materials; (ii) The licensed activities carried out on the acquired or used property; and (iii) The release and final disposition of any property recorded in paragraph (c)(4)(i) of this section, the historical site assessment performed for the release, radiation surveys performed to support release of the property, submittals to the NRC made under § 53.1070, and the methods employed to ensure that the property met the radiological criteria of subpart E of 10 CFR part 20 at the time the property was released. (d) Each holder of an OL or COL under this part must at or about 5 years prior to the projected end of operations submit a preliminary DCE which includes an up-to-date assessment of the major factors that could affect the cost to decommission. (e) Prior to or within 2 years following permanent cessation of operations, the licensee must submit a PSDAR to the NRC, and a copy to the affected State(s). The PSDAR must contain a description of the planned decommissioning activities along with a schedule for their accomplishment, a discussion that provides the reasons for concluding that the environmental impacts associated with site-specific decommissioning activities will be bounded by appropriate previously issued environmental impact statements, and a site-specific DCE, including the projected cost of managing irradiated fuel. (f) For decommissioning activities that delay completion of decommissioning by including a period of storage or surveillance, the licensee must provide a means of adjusting cost estimates and associated funding levels over the storage or surveillance period. (g) After submitting its site-specific DCE required by paragraph (e) of this section, and until the licensee has completed its final radiation survey and demonstrated that residual radioactivity has been reduced to a level that permits termination of its license, the licensee must annually submit to the NRC, by March 31, a financial assurance status report. The report must include the following information, current through the end of the previous calendar year: (1) The amount spent on decommissioning, both cumulative and over the previous calendar year, the remaining balance of any decommissioning funds, and the amount provided by other financial assurance methods being relied upon; (2) An estimate of the costs to complete decommissioning, reflecting any difference between actual and estimated costs for work performed during the year, and the decommissioning criteria upon which the estimate is based; (3) Any modifications occurring to a licensee's current method of providing financial assurance since the last submitted report; and (4) Any material changes to trust agreements or financial assurance contracts. (5) If the sum of the balance of any remaining decommissioning funds, plus earnings on such funds calculated at not greater than a 2 percent real rate of return, together with the amount provided by other financial assurance methods being relied upon, does not cover the estimated cost to complete the decommissioning, the financial assurance status report must include 
                                    ( print page 87076)
                                    additional financial assurance to cover the estimated cost of completion. (h) After submitting its site-specific DCE required by paragraph (e) of this section, the licensee must annually submit to the NRC, by March 31, a report on the status of its funding for managing irradiated fuel. The report must include the following information, current through the end of the previous calendar year: (1) The amount of funds accumulated to cover the cost of managing the irradiated fuel; (2) The projected cost of managing irradiated fuel until title to the fuel and possession of the fuel is transferred to the Secretary of Energy; and (3) If the funds accumulated do not cover the projected cost, a plan to obtain additional funds to cover the cost.
 § 53.1070
  For each holder of an OL or COL under this part— (a)(1) When the licensee has determined to permanently cease operations the licensee must, within 30 days, submit a written certification to the NRC, consistent with the requirements of § 53.040(b)(8); (2) When appropriate to support decommissioning activities and the eventual permanent removal of fuel from the reactor vessel, the licensee must develop defueled technical specifications by reviewing the operational technical specifications and determining which specifications no longer apply during decommissioning and which ones should remain applicable. The licensee must make the appropriate submittals to the NRC in accordance with § 53.1510 to request changes to the technical specifications; and (3)(i) Once fuel has been permanently removed from the reactor vessel, the licensee must submit a written certification to the NRC that meets the requirements of § 53.040(b)(9); and (ii) The licensee must establish and maintain staffing consisting of certified fuel handlers, as defined under § 53.020, and other non-licensed personnel with appropriate qualifications, and in sufficient numbers, to ensure support for facility operations and radiological control activities, as required by the facility defueled technical specifications. These personnel must be subject to the training requirements of § 53.830. (b) Upon docketing of the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel, or when a final legally effective order to permanently cease operations has come into effect, the license issued under this part no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel. (c) Decommissioning will be completed within 60 years of permanent cessation of operations. Completion of decommissioning beyond 60 years will be approved by the Commission only when necessary to protect public health and safety. Factors that will be considered by the Commission in evaluating an alternative that provides for completion of decommissioning beyond 60 years of permanent cessation of operations include unavailability of waste disposal capacity and other site-specific factors affecting the licensee's capability to carry out decommissioning, including presence of other nuclear facilities at the site. (d)(1) Prior to or within 2 years following permanent cessation of operations, the licensee must submit a PSDAR and site-specific DCE in accordance with § 53.1060(e). (2) The NRC must notice receipt of the PSDAR and make the PSDAR publicly available and publish notice of its availability for public comment in the 
                                    **Federal Register**
                                    . The NRC must also schedule a public meeting readily accessible to individuals in the vicinity of the licensee's facility. The NRC must publish a notice in the 
                                    **Federal Register**
                                     and in a forum, such as local newspapers, that is readily accessible to individuals in the vicinity of the site, announcing the date, time, and location of the meeting, along with a brief description of the purpose of the meeting. (e) Licensees must not perform any major decommissioning activities, as defined in § 53.020, until 90 days after the NRC has received the licensee's PSDAR submittal and until certifications of permanent cessation of operations and permanent removal of fuel from the reactor vessel, as required under paragraph (a) of this section, have been submitted. (f) Licensees must not perform any decommissioning activities, as defined in § 53.020, that— (1) Foreclose release of the site for possible unrestricted use; (2) Result in significant environmental impacts not previously reviewed; or (3) Result in there no longer being reasonable assurance that adequate funds will be available for decommissioning. (g) In taking actions permitted under § 53.1540 following submittal of the PSDAR, the licensee must notify the NRC in writing, and send a copy to the affected State(s), before performing any decommissioning activity inconsistent with, or making any significant schedule change from, those actions and schedules described in the PSDAR, including changes that increase the decommissioning cost by more than 20 percent from the previously provided DCE. (h) Licensees may use decommissioning trust funds consistent with the limitations of § 53.1045(a). Licensees must report on the status of decommissioning trust funds consistent with the requirements of § 53.1060. (i) Licensees must submit an application for termination of license in accordance with § 53.1070. The application for termination of license must be accompanied or preceded by a license termination plan to be submitted for NRC approval. (1) The license termination plan must be a supplement to the Final Safety Analysis Report or equivalent and must be submitted at least 2 years before termination of the license date. (2) The license termination plan must include— (i) A site characterization; (ii) Identification of remaining dismantlement activities; (iii) Plans for site remediation; (iv) Detailed plans for the final radiation survey; (v) A description of the end use of the site, if restricted; (vi) An updated site-specific estimate of remaining decommissioning costs; (vii) A supplement to the environmental report, pursuant to § 51.53 of this chapter, describing any new information or significant environmental change associated with the licensee's proposed termination activities; and (viii) Identification of parts, if any, of the facility or site that were released for use before approval of the license termination plan. (3) Following receipt of the license termination plan, the NRC must make the license termination plan publicly available and publish notice of its availability for public comment in the 
                                    **Federal Register**
                                    . The NRC must also schedule a public meeting readily accessible to individuals in the vicinity of the licensee's facility upon receipt of the license termination plan. The NRC must publish a notice in the 
                                    **Federal Register**
                                     and in a forum, such as local newspapers, that is readily accessible to individuals in the vicinity of the site, announcing the date, time, and location of the meeting, along with a brief description of the purpose of the meeting. (j) If the license termination plan demonstrates that the remainder of 
                                    ( print page 87077)
                                    decommissioning activities will be performed in accordance with the regulations in this chapter, will not be inimical to the common defense and security or to the health and safety of the public, and will not have a significant effect on the quality of the environment and after notice to interested persons, the Commission will approve the plan, by license amendment, subject to such conditions and limitations as it deems appropriate and necessary and authorize implementation of the license termination plan. (k) The Commission will terminate the license if it determines that— (1) The remaining dismantlement has been performed in accordance with the approved license termination plan, and (2) The final radiation survey and associated documentation, including an assessment of dose contributions associated with parts released for use before approval of the license termination plan, demonstrate that the facility and site have met the criteria for decommissioning in subpart E of 10 CFR part 20.
 § 53.1075
  (a) Licensees that have submitted the certifications required under § 53.1070 must maintain a decommissioning fire protection program to address the potential for fires that could cause the release or spread of radioactive materials. (1) The objectives of the decommissioning fire protection program are to (i) Reasonably prevent these fires from occurring; (ii) Rapidly detect, control, and extinguish those fires that do occur and that could result in a radiological hazard; and (iii) Ensure that the risk of fire-induced radiological hazards to the public, environment, and plant personnel is minimized. (2) The licensee must assess the decommissioning fire protection program on a regular basis. The licensee must revise the decommissioning fire protection program documentation as appropriate throughout the various stages of facility decommissioning. (3) The licensee may make changes to the decommissioning fire protection program without NRC approval if these changes do not reduce the effectiveness of fire protection for structures, systems, and components that could result in a radiological hazard, taking into account the decommissioning plant conditions and activities. (b) [Reserved]
 § 53.1080
  (a) Prior written NRC approval is required to release part of a commercial nuclear plant or site for unrestricted use at any time before receiving approval of a license termination plan. Section 53.1060 specifies recordkeeping requirements associated with partial release. Holders of an OL or COL under this part seeking NRC review and approval must— (1) Evaluate the effect of releasing the property to ensure that— (i) The dose to individual members of the public does not exceed the limits and standards of subpart D of 10 CFR part 20; (ii) There is no reduction in the effectiveness of emergency planning or physical security; (iii) Effluent releases remain within license conditions; (iv) The environmental monitoring program and offsite dose calculation manual are revised to account for the changes; (v) The siting criteria of subpart D of this part continue to be met; and (vi) All other applicable statutory and regulatory requirements continue to be met. (2) Perform a historical site assessment of the part of the commercial nuclear plant or site to be released; and (3) Perform surveys adequate to demonstrate compliance with the radiological criteria for unrestricted use specified in § 20.1402 of this chapter for impacted areas. (b) For release of non-impacted areas, the licensee may submit a written request for NRC review and approval of the release if a license amendment is not otherwise required. The request submittal must include— (1) The results of the evaluations performed in accordance with paragraphs (a)(1) and (a)(2) of this section; (2) A description of the part of the commercial nuclear plant or site to be released; (3) The schedule for release of the property; (4) The results of the evaluations performed in accordance with § 53.1540; and (5) A discussion that provides the reasons for concluding that the environmental impacts associated with the licensee's proposed release of the property will be bounded by appropriate previously issued environmental impact statements. (c) After receiving a request from the licensee for NRC approval of the release of a non-impacted area, the NRC must— (1) Determine whether the licensee has adequately evaluated the effect of releasing the property as required by paragraph (a)(1) of this section; (2) Determine whether the licensee's classification of any release areas as non- impacted is adequately justified; and (3) If determining that the licensee's submittal is adequate, inform the licensee in writing that the release is approved. (d) For release of impacted areas, the licensee must submit an application for amendment of its license for the release of the property. The application must include— (1) The information specified in paragraphs (b)(1) through (b)(3) of this section; (2) The methods used for and results obtained from the radiation surveys required to demonstrate compliance with the radiological criteria for unrestricted use specified in § 20.1402; and (3) A supplement to the environmental report, under § 51.53 of this chapter, describing any new information or significant environmental change associated with the licensee's proposed release of the property. (e) After receiving a license amendment application from the licensee for the release of an impacted area, the NRC must— (1) Determine whether the licensee has adequately evaluated the effect of releasing the property as required by paragraph (a)(1) of this section; (2) Determine whether the licensee's classification of any release areas as non-impacted is adequately justified; (3) Determine whether the licensee's radiation survey for an impacted area is adequate; and (4) If determining that the licensee's submittal is adequate, approve the licensee's amendment application. (f) The NRC must publish notice receipt of the release approval request or license amendment application in the 
                                    **Federal Register**
                                     and make the approval request or license amendment application available for public comment. Before acting on an approval request or license amendment application submitted in accordance with this section, the NRC must conduct a public meeting readily accessible to individuals in the vicinity of the licensee's facility for the purpose of obtaining public comments on the proposed release of part of the commercial nuclear plant or site. The NRC must publish a document in the 
                                    **Federal Register**
                                     and in a forum, such as local newspapers, which is readily accessible to individuals in the vicinity of the site, announcing the date, time, 
                                    ( print page 87078)
                                    and location of the meeting, along with a brief description of the purpose of the meeting. Subpart H—Licenses, Certifications, and Approvals
 § 53.1100
  (a) 
                                    *Serving of applications.*
                                     (1) Each filing of an application for a standard design approval, standard design certification, or license under this part, and any amendments to the applications, must be submitted to the U.S. Nuclear Regulatory Commission (NRC) under § 53.040, as applicable. (2) Each applicant for a construction permit (CP), early site permit, combined license (COL), or manufacturing license (ML) under this part must, upon notification by the presiding officer designated to conduct the public hearing required by the Atomic Energy Act of 1954, as amended, (the Act) update the application and serve the updated copies of the application or parts of it, eliminating all superseded information, together with an index of the updated application, as directed by presiding officer. Any subsequent amendment to the application must be served on those served copies of the application and must be submitted to the NRC as specified in § 53.040, as applicable. (3) The applicant must make a copy of the updated application available at the public hearing for the use of any other parties to the proceeding and must certify that the updated copies of the application contain the current contents of the application submitted in accordance with the requirements under this part. (4) At the time of filing an application, the Commission will make available at the NRC website, 
                                    *http://www.nrc.gov,*
                                     a copy of the application, subsequent amendments, and other records pertinent to the matter that is the subject of the application for public inspection and copying. (5) The serving of copies required by this section must not occur until the application has been docketed under § 2.101(a) of this chapter. Copies must be submitted to the Commission, as specified in § 53.040, as applicable, to enable the Director, Office of Nuclear Reactor Regulation to determine whether the application is sufficiently complete to permit docketing. (b) 
                                    *Oath or affirmation.*
                                     Each application for a standard design approval, standard design certification, or license, including, whenever appropriate, a CP or early site permit, or amendment of it, and each amendment of each application must be executed in a signed original by the applicant or duly authorized officer thereof under oath or affirmation. (c) [Reserved] (d) [Reserved] (e) 
                                    *Filing fees.*
                                     Each application for a standard design approval, standard design certification, or commercial nuclear plant license under this part, including, whenever appropriate, a CP, COL, operating license (OL), ML, or early site permit, other than a license exempted from 10 CFR part 170, must be accompanied by the fee prescribed in 10 CFR part 170. No fee will be required to accompany an application for renewal, amendment, or termination of a CP, OL, COL, or ML, except as provided in § 170.21 of this chapter. (f) 
                                    *Environmental report.*
                                     An application for a CP, OL, early site permit, design certification, COL, or ML for a commercial nuclear plant must be accompanied by an environmental report required under subpart A of 10 CFR part 51.
 § 53.1101
  Except as provided in § 53.1120, no person within the United States may transfer or receive in interstate commerce, manufacture, produce, transfer, acquire, possess, or use any utilization facility except as authorized by a license issued by the Commission.
 § 53.1103
  (a) An applicant may combine several applications in one application for different kinds of licenses under the regulations in this chapter. (b) The Commission may combine in a single license the activities of an applicant which would otherwise be licensed separately.
 § 53.1106
  An applicant may incorporate by reference in its application information contained in previous applications, statements, or reports filed with the Commission, provided, however, that such references are clear and specific.
 § 53.1109
  Each application must include, unless otherwise indicated in this subpart— (a) Name of applicant; (b) Address of applicant; (c) Description of business or occupation of applicant; (d)(1) If applicant is an individual, the citizenship of applicant; (2) If applicant is a partnership, the name, citizenship and address of each partner and the principal location where the partnership does business; (3) If applicant is a corporation or an unincorporated association, the following information: (i) The State where it is incorporated or organized and the principal location where it does business; (ii) The names, addresses and citizenship of its directors and of its principal officers; and (iii) Whether it is owned, controlled, or dominated by an alien, a foreign corporation, or foreign government, and if so, give details; or (4) If the applicant is acting as agent or representative of another person in filing the application, identify the principal and furnish information required under this paragraph with respect to such principal; (e) The class and type of license applied for, the use to which the facility will be put, the period of time for which the license is sought, and a list of other licenses, except operator's licenses, issued or applied for in connection with the proposed facility; (f) [Reserved] (g)(1) Except as provided in paragraph (g)(2) of this section, if the application is for an OL or COL for a commercial nuclear plant, or if the application is for an early site permit for a commercial nuclear plant and contains plans for coping with emergencies under § 53.1146(b)(2)(ii), the applicant must submit the radiological emergency response plans of State, local, and participating Tribal governmental entities in the United States that are wholly or partially within the plume exposure pathway emergency planning zone (EPZ),1
                                     and the plans of State governments wholly or partially within the ingestion pathway EPZ.2
                                     If the application is for an early site permit that, under § 53.1146(b)(2)(i), proposes major features of the emergency plans describing the EPZs, then the descriptions of the EPZs must meet the requirements of this paragraph. Generally, the plume exposure pathway EPZ for a commercial nuclear plant must consist of an area about 10 miles (16 km) in radius and the ingestion pathway EPZ must consist of an area about 50 miles (80 km) in radius. The exact size and configuration of the EPZs surrounding a particular commercial nuclear plant must be determined in relation to the local emergency response needs and capabilities as they are affected by such conditions as demography, topography, land characteristics, access routes, and jurisdictional boundaries. The size of the EPZs also may be determined on a case-by-case basis for gas-cooled reactors and for reactors with an authorized power level less than 250 
                                    ( print page 87079)
                                    megawatt thermal. The plans for the ingestion pathway must focus on such actions as are appropriate to protect the food ingestion pathway. (2) Applicants for commercial nuclear plants consisting of either small modular reactors or non-light-water reactors complying with § 50.160 of this chapter who apply for a CP, an OL, a COL, or an early site permit under this part must submit as part of the application the analysis used to determine whether the criteria in § 53.1109(g)(2)(i)(A) and (B) are met and, if they are met, the size of the plume exposure pathway EPZ. (i) The plume exposure pathway EPZ is the area within which: (A) Public dose, as defined in § 20.1003 of this chapter, is projected to exceed 10 millisieverts (1 rem) total effective dose equivalent over 96 hours from the release of radioactive materials from the facility considering accident likelihood and source term, timing of the accident sequence, and meteorology; and (B) Pre-determined, prompt protective measures are necessary. (ii) If the application is for an OL or COL or if the application is for an early site permit and contains plans for coping with emergencies under § 53.1146(b)(2)(ii), and if the plume exposure pathway EPZ extends beyond the site boundary: (A) The applicant must submit radiological emergency response plans of State, local, and participating Tribal governmental entities in the United States that are wholly or partially within the plume exposure pathway EPZ. (B) The exact configuration of the plume exposure pathway EPZ surrounding the facility shall be determined in relation to the local emergency response needs and capabilities as they are affected by such conditions as demography, topography, land characteristics, access routes, and jurisdictional boundaries. (iii) If the application is for an early site permit that, under § 53.1146(b)(2)(i), proposes major features of the emergency plans and describes the EPZ, and if the EPZ extends beyond the site boundary, then the exact configuration of the plume exposure pathway EPZ surrounding the facility must be determined in relation to the local emergency response needs and capabilities as they are affected by such conditions as demography, topography, land characteristics, access routes, and jurisdictional boundaries. (h) [Reserved] (i) A list of the names and addresses of such regulatory agencies as may have jurisdiction over the rates and services incident to the proposed activity, and a list of trade and news publications which circulate in the area where the proposed activity will be conducted and which are considered appropriate to give reasonable notice of the application to those municipalities, private utilities, public bodies, and cooperatives, which might have a potential interest in the facility; and (j) If the application contains Restricted Data or classified National Security information, confirmation that all Restricted Data and classified National Security information are separated from the unclassified information. 1
                                         EPZs are discussed in NUREG-0396, U.S. Environmental Protection Agency 520/1-78-016, “Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light-Water Nuclear Power Plants,” December 1978. 2
                                         If the State, local, and participating Tribal emergency response plans have been previously provided to the NRC for inclusion in the facility docket, the applicant need only provide the appropriate reference to meet this requirement.
 § 53.1112
  (a) Each CP, early site permit, and COL under this part may include conditions to address environmental issues during construction. These conditions are to be set out in an attachment to the license, which is incorporated in and made a part of the license. These conditions will be derived from information contained in the environmental report submitted pursuant to § 51.50 of this chapter, as analyzed and evaluated in the NRC record of decision and will identify the obligations of the licensee in the environmental area, including, as appropriate, requirements for reporting and keeping records of environmental data, and any conditions and monitoring requirement for the protection of the nonaquatic environment. (b) Each license authorizing operation of a commercial nuclear plant under this part, and each license for a commercial nuclear plant for which the certification of permanent cessation of operations required under § 53.1070 has been submitted may include conditions to address environmental issues during operation and decommissioning. These conditions are to be set out in an attachment to the license, which is incorporated in and made a part of the license. These conditions will be derived from information contained in the environmental report or the supplement to the environmental report submitted under §§ 51.50 and 51.53 of this chapter as analyzed and evaluated in the NRC record of decision, and will identify the obligations of the licensee in the environmental area, including, as appropriate, requirements for reporting and keeping records of environmental data and any conditions and monitoring requirement for the protection of the nonaquatic environment.
 § 53.1115
  As part of its application and in any event before the receipt of Restricted Data or classified National Security Information or the issuance of a license or standard design approval under this part, or before the Commission has adopted a final standard design certification rule under this part, the applicant must agree in writing that it will not permit any individual to have access to or any facility to possess Restricted Data or classified National Security Information until the individual and/or facility has been approved for access under the provisions of 10 CFR parts 25 and/or 95. The agreement of the applicant becomes part of the license or standard design approval.
 § 53.1118
  Any person who is a citizen, national, or agent of a foreign country, or any corporation, or other entity which the Commission knows or has reason to believe is owned, controlled, or dominated by an alien, a foreign corporation, or a foreign government, will be ineligible to apply for and obtain a license.
 § 53.1120
  Nothing in this part must be deemed to require a license for— (a) The manufacture, production, or acquisition by the Department of Defense of any utilization facility authorized pursuant to section 91 of the Act or the use of such facility by the Department of Defense or by a person under contract with and for the account of the Department of Defense; (b) Except to the extent that the Department of Energy facilities of the types subject to licensing pursuant to section 202 of the Energy Reorganization Act of 1974, as amended, are involved— (1)(i) The processing, fabrication or refining of special nuclear material (SNM) or the separation of SNM, or the separation of SNM from other substances by a prime contractor of the Department of Energy under a prime contract for—
                                    ( print page 87080) (A) The performance of work for the Department of Energy at a United States government-owned or controlled site; (B) Research in, or development, manufacture, storage, testing or transportation of, atomic weapons or components thereof; or (C) The use or operation of a utilization facility in a United States owned vehicle or vessel; or (ii) The processing, fabrication or refining of SNM of the separation of SNM, or the separation of SNM from other substances by a prime contractor or subcontractor of the Commission or the Department of Energy under a prime contract or subcontract when the Commission determines that the exemption of the prime contractor or subcontractor is authorized by law; and that, under the terms of the contract or subcontract, there is adequate assurance that the work thereunder can be accomplished without undue risk to the public health and safety; or (2)(i) The construction or operation of a utilization facility for the Department of Energy at a United States government-owned or controlled site, including the transportation of the utilization facility to or from such site and the performance of contract services during temporary interruptions of such transportation; or the construction or operation of a utilization facility for the Department of Energy in the performance of research in, or development, manufacture, storage, testing, or transportation of, atomic weapons or components thereof; or the use or operation of a utilization facility for the Department of Energy in a United States government-owned vehicle or vessel; provided that such activities are conducted by a prime contractor of the Department of Energy under a prime contract with the Department of Energy; or (ii) The construction or operation of a utilization facility by a prime contractor or subcontractor of the Commission or the Department of Energy under his or her prime contract or subcontract when the Commission determines that the exemption of the prime contractor or subcontractor is authorized by law; and that, under the terms of the contract or subcontract, there is adequate assurance that the work thereunder can be accomplished without undue risk to the public health and safety; or (c) The transportation or possession of any utilization facility by a common or contract carrier or warehouse employee in the regular course of carriage for another or storage incident thereto.
 § 53.1121
  Applications and documents submitted to the Commission in connection with applications may be made available for public inspection under the provisions of part 2 of this chapter.
 § 53.1124
  (a) 
                                    *Limited work authorization.*
                                     An application for a limited work authorization (LWA) under this part may be submitted as part of an application for an early site permit, CP, or COL under this part as required in § 53.1130(a)(2). (b) 
                                    *Early site permit.*
                                     (1) A holder of an early site permit may request an LWA. (2) An application for a CP or COL under this part may, but need not, reference an early site permit. (c) 
                                    *Standard design approval.*
                                     An application for a standard design approval under this part may, but need not, reference an OL or custom COL under this part that is essentially the same as the information supporting the standard design for which approval is being requested. (d) 
                                    *Standard design certification.*
                                     An application for a standard design certification under this part may, but need not, reference an OL or custom COL under this part that is essentially the same as the standard design for which certification is being requested. (e) 
                                    *Manufacturing license.*
                                     (1) A manufactured reactor manufactured under an ML issued under this part may only be transported to and installed at a site for which a COL under this part has been issued. (2) An ML applicant under this part may reference a standard design certification or a standard design approval under this part in its application. (f) 
                                    *Construction permit.*
                                     An application for a CP may, but need not, reference a standard design certification or standard design approval issued under this part, respectively, and may also reference an early site permit issued under this part. In the absence of a demonstration that an entity other than the one originally sponsoring a standard design certification is qualified to supply a design, the Commission will entertain an application for a CP that references a standard design certification issued under this part only if the entity that sponsored the certification supplies the design for the applicant's use. (g) 
                                    *Operating license.*
                                     (1) An application for an OL under this part may, but need not, reference an early site permit, standard design certification, or standard design approval issued under this part. In the absence of a demonstration that an entity other than the one originally sponsoring a standard design certification is qualified to supply a design, the Commission will entertain an application for an OL that references a standard design certification issued under this part only if the entity that sponsored the certification supplies the design for the applicant's use. (2) The holder of a CP must, at the time of submission of the Final Safety Analysis Report (FSAR), file an application for an OL. (h) 
                                    *Combined licenses.*
                                     An application for a COL under this part may, but need not, reference an early site permit, standard design certification, standard design approval, or ML issued under this part. In the absence of a demonstration that an entity other than the one originally sponsoring and obtaining a standard design certification is qualified to supply a design, the Commission will entertain an application for a COL that references a standard design certification issued under this part only if the entity that sponsored the certification supplies the design for the applicant's use.
 § 53.1130
  (a) 
                                    *Request for limited work authorization.*
                                     (1) Any person to whom the Commission may otherwise issue either a license or permit related to a commercial nuclear plant may request an LWA allowing that person to perform the driving of piles, subsurface preparation, placement of backfill, concrete, or permanent retaining walls within an excavation, and installation of the foundation, including placement of concrete, any of which are for a structure, system, or component (SSC) of the facility for which either a CP or COL is otherwise required under § 53.610. (2) An application for an LWA may be submitted as part of a complete application for a CP or COL in accordance with § 2.101(a)(1) through (a)(5) of this chapter, or as a partial application in accordance with § 2.101(a)(9) of this chapter. An application for an LWA by the holder of an early site permit must be submitted as a complete application in accordance with § 2.101(a)(1) through (a)(4) of this chapter. (3) The application must include— (i) A Safety Analysis Report required by §§ 53.1146, 53.1309 or 53.1416, as applicable, a description of the activities requested to be performed, and the design and construction information 
                                    ( print page 87081)
                                    otherwise required by the Commission's rules and regulations to be submitted for a CP or COL under this part but limited to those portions of the facility that are within the scope of the LWA. The Safety Analysis Report must demonstrate that activities conducted under the LWA will be conducted in compliance with the technically relevant Commission requirements in 10 CFR chapter I applicable to the design of those portions of the facility within the scope of the LWA; (ii) An environmental report in accordance with § 51.49 of this chapter; and (iii) A plan for redress of activities performed under the LWA, should limited work activities be terminated by the holder, or the LWA be revoked by the NRC or upon effectiveness of the Commission's final decision denying the associated CP or COL application, as applicable. (b) 
                                    *Issuance of limited work authorization.*
                                     (1) The Director, Office of Nuclear Reactor Regulation may issue an LWA only after— (i) The NRC staff issues the final environmental impact statement for the LWA under subpart A of part 51 of this chapter; (ii) The presiding officer makes the finding in §§ 51.105(c) or 51.107(d) of this chapter, as applicable; (iii) The Director determines that the applicable standards and requirements of the Act, and the Commission's regulations applicable to the activities to be conducted under the LWA, have been met, the applicant is technically qualified to engage in the activities authorized, and that issuance of the LWA will provide reasonable assurance of adequate protection to public health and safety and will not be inimical to the common defense and security; and (iv) The presiding officer finds that there are no unresolved safety issues relating to the activities to be conducted under the LWA that would constitute good cause for withholding the authorization. (2) Each LWA will specify the activities that the holder is authorized to perform. (c) 
                                    *Effect of limited work authorization.*
                                     Any activities undertaken under an LWA are entirely at the risk of the applicant and, except as to the matters determined under paragraph (b)(1) of this section, the issuance of the LWA has no bearing on the issuance of a CP or COL with respect to the requirements of the Act and rules, regulations, or orders issued under the Act. The environmental impact statement for a CP or COL application for which an LWA was previously issued will not address, and the presiding officer will not consider, the sunk costs of the holder of the LWA in determining the proposed action (
                                    *i.e.,*
                                     issuance of the CP or COL). (d) 
                                    *Implementation of redress plan.*
                                     If construction is terminated by the holder, the underlying application is withdrawn by the applicant or denied by the NRC, or the LWA is revoked by the NRC, then the holder must begin implementation of the redress plan in a reasonable time. The holder must also complete the redress of the site no later than 18 months after termination of construction, revocation of the LWA, or upon effectiveness of the Commission's final decision denying the associated CP application or the associated COL application, as applicable.
 § 53.1140
  Sections 53.1140 through 53.1188 set out the requirements and procedures applicable to Commission issuance of an early site permit under this part for approval of a site for a commercial nuclear plant separate from the filing of an application for a CP or COL for the facility.
 § 53.1143
  Any person who may apply for a CP or for a COL under this part, may file an application for an early site permit with the Director, Office of Nuclear Reactor Regulation. An application for an early site permit may be filed notwithstanding the fact that an application for a CP or a COL has not been filed in connection with the site for which a permit is sought.
 § 53.1144
  The application must contain all of the information required by § 53.1109(a) through (d) and (j).
 § 53.1146
  (a) The application must contain— (1) A Site Safety Analysis Report that must include the following: (i) The specific number, type, and thermal power level of the facilities, or range of possible facilities, for which the site may be used; (ii) The anticipated maximum levels of radiological and thermal effluents each facility will produce; (iii) The type of cooling systems, including intakes and outflows, where appropriate, that may be associated with each facility; (iv) The boundaries of the site; (v) The proposed general location of each facility on the site; (vi) The external hazards and site characteristics required by this part; (vii) The location and description of any nearby industrial, military, or transportation facilities and routes; (viii) The existing and projected future population profile of the area surrounding the site; (ix) A description and assessment of the site on which a facility is to be located. The assessment must address the requirements of subpart D of this part; (x) Information demonstrating that site characteristics are such that adequate security plans and measures can be developed; and (xi) A description of the quality assurance program (QAP) required by appendix B to part 50 of this chapter applied to site-related activities for the future design, fabrication, construction, and testing of the SSCs of a facility or facilities that may be constructed on the site. (2) A complete environmental report as required by § 51.50(b) of this chapter. (b)(1) The Site Safety Analysis Report must identify physical characteristics of the proposed site, such as egress limitations from the area surrounding the site, that could pose a significant impediment to the development of emergency plans. If physical characteristics are identified that could pose a significant impediment to the development of emergency plans, the application must identify measures that would, when implemented, mitigate or eliminate the significant impediment. (2) The Site Safety Analysis Report may also— (i) Propose major features of the emergency plans, under either § 50.160 or the requirements in appendix E to part 50 and § 50.47(b) of this chapter, as applicable, such as the exact size and configuration of the EPZs, for review and approval by the NRC, in consultation with the Federal Emergency Management Agency (FEMA), as applicable, in the absence of complete and integrated emergency plans; or (ii) Propose complete and integrated emergency plans for review and approval by the NRC, in consultation with FEMA, as applicable, in accordance with either § 50.160 or the requirements in appendix E to part 50 and § 50.47(b) of this chapter. To the extent approval of emergency plans is sought, the application must contain the information required by § 53.1109(g). (3) Emergency plans submitted under paragraph (b)(2)(ii) of this section must include the proposed inspections, tests, and analyses that the holder of a COL referencing the early site permit must perform, and the acceptance criteria that are necessary and sufficient to provide 
                                    ( print page 87082)
                                    reasonable assurance that, if the inspections, tests, and analyses are performed and the acceptance criteria met, the facility has been constructed and will be operated in conformity with the emergency plans, the provisions of the Act, and the Commission's rules and regulations. Major features of an emergency plan submitted under paragraph (b)(2)(i) of this section may include proposed inspections, tests, analyses, and acceptance criteria (ITAAC). (4) Under paragraphs (b)(1) and (b)(2)(i) of this section, the Site Safety Analysis Report must include, where appropriate, a description of contacts and arrangements made with Federal, State, participating Tribal, and local governmental agencies with emergency planning responsibilities. The Site Safety Analysis Report must contain any certifications that have been obtained. If these certifications, where appropriate, cannot be obtained, the Site Safety Analysis Report must contain information, including a utility plan, sufficient to show that the proposed plans provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency at the site. Under the option set forth in paragraph (b)(2)(ii) of this section, the applicant must make good faith efforts, where appropriate, to obtain from the same governmental agencies certifications that— (i) The proposed emergency plans are practicable; (ii) These agencies are committed to participating in any further development of the plans, including any required field demonstrations; and (iii) That these agencies are committed to executing their responsibilities under the plans in the event of an emergency. (c) An applicant may request that an LWA under § 53.1130 be issued in conjunction with the early site permit. The application must include the information otherwise required by § 53.1130. (d) Each applicant for an early site permit under this part must protect safeguards information against unauthorized disclosure in accordance with the requirements in §§ 73.21 and 73.22 of this chapter, as applicable.
 § 53.1149
  (a) 
                                    *Standards for review of applications.*
                                     Applications filed under this part will be reviewed according to the applicable standards set out in this part. In addition, the Commission must prepare an environmental impact statement during review of the application, under the applicable provisions of 10 CFR part 51. The Commission must determine, after consultation with FEMA, as applicable, whether the information required of the applicant by § 53.1146(b)(1) shows that there is no significant impediment to the development of emergency plans that cannot be mitigated or eliminated by measures proposed by the applicant, whether any major features of emergency plans submitted by the applicant under § 53.1146(b)(2)(i) are acceptable under either § 50.160 or appendix E to part 50 and § 50.47(b) of this chapter, and whether any emergency plans submitted by the applicant under § 53.1146(b)(2)(ii) provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. (b) 
                                    *Administrative review of applications; hearings.*
                                     An early site permit application is subject to all procedural requirements in 10 CFR part 2, including the requirements for docketing in § 2.101(a)(1) through (4) of this chapter, and the requirements for issuance of a notice of hearing in § 2.104(a) and (d) of this chapter, provided that the designated sections may not be construed to require that the environmental report, or draft or final environmental impact statement includes an assessment of the benefits of construction and operation of the reactor or reactors, or an analysis of alternative energy sources. The presiding officer in an early site permit hearing must not admit contentions proffered by any party concerning an assessment of the benefits of construction and operation of the reactor or reactors, or an analysis of alternative energy sources if those issues were not addressed by the applicant in the early site permit application. All hearings conducted on applications for early site permits filed under this part are governed by the procedures contained in subparts C, G, L, and N of 10 CFR part 2, as applicable.
 § 53.1155
  The Commission must refer a copy of the application for an early site permit to the Advisory Committee on Reactor Safeguards (ACRS). The ACRS must report on those portions of the application which concern safety.
 § 53.1158
  (a) After conducting a hearing under § 53.1149(b) and receiving the report to be submitted by the ACRS under § 53.1155, the Commission may issue an early site permit, in the form the Commission deems appropriate, if the Commission finds that— (1) An application for an early site permit demonstrates compliance with the applicable standards and requirements of the Act and the Commission's regulations; (2) Notifications, if any, to other agencies or bodies have been duly made; (3) There is reasonable assurance that the site is in conformity with the provisions of the Act and the Commission's regulations; (4) The applicant is technically qualified to engage in any activities authorized; (5) The proposed ITAAC, including any on emergency planning, are necessary and sufficient, within the scope of the early site permit, to provide reasonable assurance that the facility has been constructed and will be operated in conformity with the license, the provisions of the Act, and the Commission's regulations; (6) Issuance of the permit will not be inimical to the common defense and security or to the health and safety of the public; (7) Any significant adverse environmental impact resulting from activities requested under § 53.1146(c) can be redressed; and (8) The findings required by subpart A of 10 CFR part 51 have been made. (b) The early site permit must specify the site characteristics, design parameters, and terms and conditions of the early site permit the Commission deems appropriate. Before issuance of either a CP or COL referencing an early site permit, the Commission must find that any relevant terms and conditions of the early site permit have been met. Any terms or conditions of the early site permit that could not be met by the time of issuance of the CP or COL, must be set forth as terms or conditions of the CP or COL. (c) The early site permit must specify those § 53.1130(b) activities requested under § 53.1146(c) that the permit holder is authorized to perform.
 § 53.1161
  If the activities authorized by § 53.1158(c) are performed and the site is not referenced in an application for a CP or a COL issued under this part while the permit remains valid, then the early site permit remains in effect solely for the purpose of site redress, and the holder of the permit must redress the site under the terms of the site redress plan required by § 53.1146(c). If, before redress is complete, a use not envisaged in the redress plan is found for the site or parts thereof, the holder of the permit 
                                    ( print page 87083)
                                    must carry out the redress plan to the greatest extent possible consistent with the alternate use.
 § 53.1164
  (a) Except as provided in paragraph (b) of this section, an early site permit issued under this subpart may be valid for not less than 10, nor more than 20 years from the date of issuance. (b) An early site permit continues to be valid beyond the date of expiration in any proceeding on a CP application or a COL application that references the early site permit and is docketed before the date of expiration of the early site permit, or, if a timely application for renewal of the permit has been docketed, before the Commission has determined whether to renew the permit. (c) An applicant for a CP or COL may, at its own risk, reference in its application a site for which an early site permit application has been docketed but not granted. (d) Upon issuance of a CP or COL, a referenced early site permit is subsumed, to the extent referenced, into the CP or COL.
 § 53.1167
  A holder of an early site permit may request an LWA under § 53.1130.
 § 53.1170
  An application to transfer an early site permit will be processed under § 53.1570.
 § 53.1173
  (a) Not less than 12, nor more than 36 months before the expiration date stated in the early site permit, or any later renewal period, the permit holder may apply for a renewal of the permit. An application for renewal must contain all information necessary to bring up to date the information and data contained in the previous application. (b) Any person whose interests may be affected by renewal of the permit may request a hearing on the application for renewal. The request for a hearing must comply with § 2.309 of this chapter. If a hearing is granted, notice of the hearing will be published under § 2.309 of this chapter. (c) An early site permit, either original or renewed, for which a timely application for renewal has been filed, remains in effect until the Commission has determined whether to renew the permit. If the permit is not renewed, it continues to be valid in certain proceedings in accordance with the provisions of § 53.1164(b). (d) The Commission must refer a copy of the application for renewal to the ACRS. The ACRS must report on those portions of the application which concern safety and must apply the criteria set forth in § 53.1176.
 § 53.1176
  (a) The Commission must grant the renewal if it determines that— (1) The site complies with the Act, the Commission's regulations, and orders applicable and in effect at the time the site permit was originally issued; and (2) Any new requirements the Commission may wish to impose— (i) Are necessary for adequate protection to public health and safety or common defense and security; (ii) Are necessary for compliance with the Commission's regulations, and orders applicable and in effect at the time the site permit was originally issued; or (iii) Would provide a substantial increase in overall protection of the public health and safety or the common defense and security to be derived from the new requirements, and the direct and indirect costs of implementation of those requirements are justified in view of this increased protection. (b) A denial of renewal under the provisions of § 53.1176(a) does not bar the permit holder or another applicant from filing a new application for the site which proposes changes to the site or the way that it is used to correct the deficiencies cited in the denial of the renewal.
 § 53.1179
  Each renewal of an early site permit may be for not less than 10, nor more than 20 years, plus any remaining years on the early site permit then in effect before renewal.
 § 53.1182
  A site for which an early site permit has been issued under this part may be used for purposes other than those described in the permit, including the location of other types of energy facilities. The permit holder must inform the Director, Office of Nuclear Reactor Regulation (Director), of any significant uses for the site which have not been approved in the early site permit. The information about the activities must be given to the Director at least 30 days in advance of any actual construction or site modification for the activities. The information provided could be the basis for imposing new requirements on the permit, under the provisions of § 53.1188. If the permit holder informs the Director that the holder no longer intends to use the site for a commercial nuclear plant, the Director may terminate the permit.
 § 53.1188
  (a) 
                                    *Commission finality.*
                                     (1) While an early site permit is in effect under § 53.1164 or § 53.1179, the Commission may not change or impose new site characteristics, design parameters, or terms and conditions, including emergency planning requirements, on the early site permit unless the Commission— (i) Determines that a modification is necessary to bring the permit or the site into compliance with the Commission's regulations and orders applicable and in effect at the time the permit was issued; (ii) Determines the modification is necessary to assure adequate protection of the public health and safety or the common defense and security; (iii) Determines that a modification is necessary based on an update under paragraph (b) of this section; or (iv) Issues a variance requested under paragraph (d) of this section. (2) In making the findings required for issuance of a CP, COL, or OL, or the findings required by § 53.1452(g), or in any enforcement hearing other than one initiated by the Commission under paragraph (a)(1) of this section, if the application for the CP, COL, or OL references an early site permit, the Commission must treat as resolved those matters resolved in the proceeding on the application for issuance or renewal of the early site permit, except as provided for in paragraphs (b), (c), and (d) of this section. (i) If the Commission grants a CP application that references an early site permit and an application for an OL references the CP, the Commission must treat as resolved those matters resolved in the proceeding for the issuance or renewal of the early site permit, except as provided for in paragraphs (b), (c), and (d) of this section. (ii) If the early site permit approved an emergency plan (or major features thereof) that is in use by a licensee of a commercial nuclear plant, the Commission must treat as resolved changes to the early site permit emergency plan (or major features thereof) that are identical to changes made to the licensee's emergency plans under § 53.1565 occurring after issuance of the early site permit. (iii) If the early site permit approved an emergency plan (or major features thereof) that is not in use by a licensee of a commercial nuclear plant, the Commission must treat as resolved changes that are equivalent to those that could be made under § 53.1565 without prior NRC approval had the emergency plan been in use by a licensee.
                                    ( print page 87084) (b) 
                                    *Updating of early site permit-emergency preparedness.*
                                     An applicant for a CP, OL, or COL who has filed an application referencing an early site permit issued under this subpart must update the emergency preparedness information that was provided under § 53.1146(b) and discuss whether the updated information materially changes the bases for compliance with applicable NRC requirements. (c) 
                                    *Hearings and petitions.*
                                     (1) In any proceeding for the issuance of a CP, OL, or COL referencing an early site permit, contentions on the following matters may be litigated in the same manner as other issues material to the proceeding: (i) The nuclear reactor proposed to be built does not fit within one or more of the site characteristics or design parameters included in the early site permit; (ii) One or more of the terms and conditions of the early site permit have not been met; (iii) A variance requested under paragraph (d) of this section is unwarranted or should be modified; (iv) New or additional information is provided in the application that substantially alters the bases for a previous NRC conclusion or constitutes a sufficient basis for the Commission to modify or impose new terms and conditions related to emergency preparedness; or (v) Any significant environmental issue that was not resolved in the early site permit proceeding, or any issue involving the impacts of construction and operation of the facility that was resolved in the early site permit proceeding for which significant new information has been identified. (2) Any person may file a petition requesting that the site characteristics, design parameters, or terms and conditions of the early site permit be modified, or that the permit be suspended or revoked. The petition will be considered under § 2.206 of this chapter. Before construction commences, the Commission must consider the petition and determine whether any immediate action is required. If the petition is granted, an appropriate order will be issued. Construction under the CP or COL will not be affected by the granting of the petition unless the order is made immediately effective. Any change required by the Commission in response to the petition must demonstrate compliance with the requirements of paragraph (a)(1) of this section. (d) 
                                    *Variances.*
                                     An applicant for a CP, OL, or COL referencing an early site permit may include in its application a request for a variance from one or more site characteristics, design parameters, or terms and conditions of the early site permit, or from the Site Safety Analysis Report. In determining whether to grant the variance, the Commission must apply the same technically relevant criteria applicable to the application for the original or renewed early site permit. Once a CP or COL referencing an early site permit is issued, variances from the early site permit will not be granted for that CP or COL. (e) 
                                    *Early site permit amendment.*
                                     The holder of an early site permit may not make changes to the early site permit or the Site Safety Analysis Report without prior Commission approval. The request for a change to the early site permit must be in the form of an application for a license amendment and must demonstrate compliance with the requirements of §§ 53.1510 and 53.1520.
 § 53.1200
  Sections 53.1200 through 53.1221 set out procedures for the filing, NRC staff review, and referral to the ACRS of standard designs, or major portions thereof, for a commercial nuclear plant under this part.
 § 53.1203
  Any person may submit a proposed standard design for a commercial nuclear plant to the NRC staff for its review. The submittal may consist of either the final design for the entire facility or the final design for major portions thereof.
 § 53.1206
  The application must contain all of the information required by § 53.1109(a) through (c) and (j).
 § 53.1209
  (a) 
                                    *Major portion of a standard design.*
                                     If the applicant seeks review of a major portion of a standard design, the application need only contain the information required by this section to the extent the requirements are applicable to the major portion of the standard design for which NRC staff approval is sought. If an applicant seeks approval of a major portion of the design, the scope of the application for which approval is sought must include all functional design criteria necessary to demonstrate compliance with the safety criteria in §§ 53.210, 53.220 and 53.450(e), as applicable, for the major portion of the standard design for which NRC staff approval is sought. Such applicants must identify conditions related to interfaces with systems outside the scope of the major portion of the standard design for which NRC staff approval is sought, and functional or physical boundary conditions between the major portion of the standard design for which NRC staff approval is sought and the remainder of the standard design. These conditions must be demonstrated when the standard design approval is incorporated into a subsequent CP, design certification, ML, or COL application. (b) 
                                    *Final Safety Analysis Report.*
                                     The application must contain an FSAR that describes the facility and the limits on its operation, presents a safety analysis of the SSCs and of the facility, or major portions thereof, for which the applicant seeks design approval, and must include the following information: (1) 
                                    *Site Parameters.*
                                     The site parameters postulated for the design under this part, including the design-basis external hazard levels for the relevant external hazards, and an analysis and evaluation of the design in terms of those site parameters. (2) 
                                    *Design information.*
                                     Except as specified in this paragraph, an application for a standard design approval for a commercial nuclear plant must include the design information equivalent to that required for a standard design certification under § 53.1239(a)(2) through (27) for those portions of a commercial nuclear plant included in the standard design approval.
 § 53.1210
  (a) In addition to the FSAR, the application must also include the following: (1) 
                                    *Availability Controls (if not included in the FSAR).*
                                     A description of the controls on plant operations, including availability controls, to provide reasonable confidence that the configurations and special treatments for safety-related (SR) SSCs and non-safety-related but safety-significant (NSRSS) SSCs provide the capabilities and reliabilities required to demonstrate compliance with the safety criteria of § 53.220. (2) 
                                    *Safeguards Information.*
                                     A description of the program to protect Safeguards Information against unauthorized disclosure in accordance with the requirements in §§ 73.21 and 73.22 of this chapter, as applicable. (b) If there are SSCs of the plant which required research and development to confirm the adequacy of their design, provide a report in the application which documents the 
                                    ( print page 87085)
                                    resolution of any safety questions associated with such SSCs. (c) A description of how the performance of each design feature has been demonstrated capable of fulfilling functional design criteria considering interdependent effects through either analysis, appropriate test programs, prototype testing, operating experience, or a combination thereof, in accordance with § 53.440(a).
 § 53.1212
  Applications filed under this part will be reviewed under the standards set out in 10 CFR parts 20, 53, and 73.
 § 53.1215
  The Commission must refer a copy of the application to the ACRS. The ACRS must report on those portions of the application which concern safety.
 § 53.1218
  (a) Upon completion of its review of a submittal under §§ 53.1200 through 53.1221 and receipt of a report by the ACRS under § 53.1215, the NRC staff must publish a determination in the 
                                    **Federal Register**
                                     as to whether or not the design is acceptable, subject to appropriate terms and conditions, and make an analysis of the design in the form of a report available at the NRC website, 
                                    *https://www.nrc.gov.* (b) A standard design approval issued under this section is valid for 15 years from the date of issuance and may not be renewed. A design approval continues to be valid beyond the date of expiration in any proceeding on an application for a CP, OL, COL, or ML under this part that references the design approval and is docketed before the date of expiration of the design approval.
 § 53.1221
  (a) An approved design must be used by and relied upon by the NRC staff and the ACRS in their reviews of any standard design certification or individual facility license application under this part that incorporates by reference a standard design approved under this part unless there exists significant new information that substantially affects the earlier determination or other good cause. (b) The determination and report by the NRC staff do not constitute a commitment to issue a permit or license, or in any way affect the authority of the Commission, Atomic Safety and Licensing Board Panel, or presiding officers in any proceeding under part 2 of this chapter. (c) Except for information requests seeking to verify compliance with the current licensing basis of the standard design approval, information requests to the holder of a standard design approval must be evaluated before issuance to ensure that the burden to be imposed on respondents is justified in view of the potential safety significance of the issue to be addressed in the requested information. Each evaluation performed by the NRC staff must be in accordance with § 53.1580 and must be approved by the Executive Director for Operations or authorized designee before issuance of the request. (d) The Commission will require, before granting a CP, COL, OL, or ML that references a standard design approval, that engineering documents, such as analyses, drawings, procurement specifications, or construction and installation specifications, be completed and available for audit if the more detailed information is necessary for the Commission to verify the information in the application and make its safety determination, including the determination that the application is consistent with the design approval information. This information may be acquired by appropriate arrangements with the design approval applicant.
 § 53.1230
  Sections 53.1230 through 53.1263 set forth the requirements and procedures applicable to the Commission's issuance of rules granting standard design certifications for commercial nuclear plants under this part separate from the filing of an application for a CP or COL for such a facility.
 § 53.1233
  (a) An application for design certification may be filed notwithstanding the fact that an application for a CP, COL, or ML for such a facility has not been filed. (b) The application must comply with the applicable filing requirements of § 53.040 and §§ 2.811 through 2.819 of this chapter.
 § 53.1236
  The application must contain all of the information required by § 53.1109(a) through (c) and (j).
 § 53.1239
  The application must contain a level of design information sufficient to enable the Commission to judge the applicant's proposed means of assuring that construction conforms to the design and to reach a final conclusion on all safety questions associated with the design before the certification is granted. The information submitted for a design certification must include performance requirements and design information sufficiently detailed to permit the preparation of acceptance and inspection requirements by the NRC. The Commission will require, before design certification, that information normally contained in engineering documents, such as analyses, drawings, procurement specifications, or construction and installation specifications, be completed and available for audit if the more detailed information is necessary for the Commission to verify the information in the application and make its safety determination. (a) 
                                    *Final Safety Analysis Report.*
                                     The application must contain an FSAR that describes the facility and the limits on its operation, and presents a safety analysis of the SSCs, and must include the following information: (1) 
                                    *Site Parameters.*
                                     The site parameters postulated for the design under this part, including the design-basis external hazard levels for the relevant external hazards, and an analysis and evaluation of the design in terms of those site parameters. (2) 
                                    *Plant Description and Safety Functions—*
                                    (i) 
                                    *General Plant Description.*
                                     A general description of the commercial nuclear plant including reactor type, the intended use of the reactor, nuclear design (
                                    *e.g.,*
                                     neutron spectrum, reactor control, multi-unit reactor control), overall layout of the plant including significant plant features and SSCs, maximum power level and the nature and inventory of radioactive materials. (ii) 
                                    *Safety functions.*
                                     A description of the primary and additional safety functions required under § 53.230 and a summary of how each safety function is satisfied. (3) 
                                    *Design Features and functional design criteria—licensing-basis events.*
                                     (i) A description of the design features required by § 53.400 and the functional design criteria required by §§ 53.410 and 53.420 that, when combined with corresponding human actions and programmatic controls, demonstrate that the plant will demonstrate compliance with the safety criteria defined in § 53.210 and established in accordance with § 53.220, or more restrictive alternative criteria adopted under § 53.470, during licensing-basis events (LBEs). (ii) A description of how design features demonstrate compliance with 
                                    ( print page 87086)
                                    the requirements of § 53.440(a) through (i) and (k) through (m). (4) 
                                    *Design Features Supporting Normal Operations.*
                                     A description of the design features required by § 53.425 to support the holder of an OL or COL complying with § 53.260 during normal operations. (5) 
                                    *Design Features and Functional Design Criteria—aircraft impact.*
                                     A description of the design features and functional design criteria required to demonstrate compliance with the requirements of § 53.440(j) for addressing the impact of a large, commercial aircraft. (6) 
                                    *Earthquake engineering.*
                                     The information necessary to demonstrate that the commercial nuclear plant complies with the earthquake engineering criteria in § 53.480. (7) 
                                    *Programmatic Controls and Interfaces.*
                                     (i) A description of the corresponding programmatic controls and interfaces necessary to achieve and maintain the reliability and capability of SSCs relied upon to demonstrate compliance with the functional design criteria required by §§ 53.410 and 53.420 and the safety criteria in §§ 53.210 and 53.220, or more restrictive alternative criteria adopted under § 53.470, and necessary to maintain consistency with analyses required by § 53.450. (ii) For an application for a multi-unit commercial nuclear plant, the programmatic controls and interfaces must also be described for different modular configurations, as required by § 53.440(i), including any restrictions that will be necessary during the construction and startup of any given unit to ensure the safe operation of the overall commercial nuclear plant to be licensed under this part. (8) 
                                    *Programmatic Controls for Normal Operations.*
                                     A description of how programmatic controls, including monitoring programs, would provide assurance that design features and procedures will enable the holder of an OL or COL to comply with § 53.260. (9) 
                                    *Design Features Supporting the Protection of Plant Workers.*
                                     A description of the design features required by § 53.430 to support the holder of an OL or COL complying with § 53.270. (10) 
                                    *Programmatic Controls for Protection of Plant Workers.*
                                     A description of how programmatic controls, including monitoring programs, would provide assurance that design features and procedures will enable the holder of an OL or COL to comply with § 53.270. (11) 
                                    *Codes and Standards.*
                                     A description of generally accepted consensus codes and standards used to design the design features, as required by § 53.440(b). (12) 
                                    *Materials.*
                                     A description of the materials used for SR and NSRSS SSCs and a description of the qualification of these materials for their service conditions over the plant lifetime, as required by § 53.440(c). (13) 
                                    *Integrity Assessment Program.*
                                     A description of a design integrity assessment program that addresses the elements described in § 53.440(d). (14) 
                                    *Safety and Security.*
                                     Confirmation that safety and security were considered together in the design process, as required by § 53.440(f). (15) 
                                    *Criticality.*
                                     Information demonstrating how the applicant will comply with requirements for criticality accidents in § 53.440(m). (16) 
                                    *Multi-unit Plants.*
                                     For an application for standard design certification of a multi-unit commercial nuclear plant, the possible operating configurations of the reactor units, including common systems, interface requirements, and system interactions, as required by § 53.440(i). (17) 
                                    *SSC Classification.*
                                     (i) The classification of SSCs according to their safety significance under § 53.460(a). (ii) For SR and NSRSS SSCs, the conditions under which they must perform the safety functions required by § 53.230, including environmental conditions. (18) 
                                    *Probabilistic Risk Assessment.*
                                     A description of the probabilistic risk assessment (PRA) required by § 53.450(a) and its results. (19) 
                                    *Analyses.*
                                     A description of the analyses performed under § 53.450(b) through (g) that includes the following information: (i) A description of the analysis of LBEs and its results, as described in § 53.240. This analysis description must— (A) Address the elements in § 53.450(e) and (f); and (B) Under § 53.460(c)— (
                                    *1*) Describe any human actions that are necessary to prevent or mitigate LBEs; (
                                    *2*) Describe how those human actions are capable of being reliably performed under the postulated environmental conditions present; and (
                                    *3*) Describe how those human actions would be addressed by programs established under subpart F of this part. (ii)(A) A description of how SSCs relied on to meet the safety criteria defined in § 53.210 are protected against or designed to withstand the effects of external hazards under § 53.510. (B) The information necessary to demonstrate that the commercial nuclear plant complies with the earthquake engineering criteria in § 53.480. (iii) A description of the defense-in-depth measures required by § 53.250. (iv) A description of all plant operating states where there is the potential for the uncontrolled release of radioactive material to the environment, as required by § 53.450(b)(4). (v) A description of the events that challenge plant control and safety systems whose failure could lead to an undesirable end state and/or radioactive material release, as required by § 53.450(b)(5). (vi) A description of the analytical codes used in modeling plant behavior in analyses of LBEs and how these codes are qualified for the range of conditions for which they were used, as required by § 53.450(d). (vii) If not described in addressing paragraph (5) of this section, the results of other analyses required by § 53.450(g). (20) 
                                    *Special Treatments.*
                                     A description of special treatments established as required by § 53.460. (21) 
                                    *Analytical Margins.*
                                     A description of any alternative criteria adopted to demonstrate analytical margins supporting operational flexibilities, if applicable, as required by § 53.470. (22) 
                                    *Quality Assurance.*
                                     A description of the QAP applied to the design of the SSCs of the commercial nuclear plant, as required by § 53.460(b). The description of the QAP for a commercial nuclear plant must include a discussion of how the applicable requirements of appendix B to part 50 of this chapter were satisfied. (23) 
                                    *Design Features and Controls to Address the Minimization of Contamination.*
                                     The information required by § 20.1406 of this chapter. (24) 
                                    *Interface Requirements.*
                                     (i) A description analysis, and evaluation of the interfaces between the standard design and the balance of the commercial nuclear plant that may impact the ability of the plant to demonstrate compliance with the functional design criteria or the safety criteria of subparts B and C of this part. (ii) Confirmation that interface requirements are verifiable through inspections, testing, or analysis. These requirements must be sufficiently detailed to allow for completion of the final safety analysis by license applicants that reference the certified design under this subpart. The method to be used for verification of interface requirements must be included as part of the proposed ITAAC required by § 53.1241(a)(3).
                                    ( print page 87087) (iii) A representative conceptual design for those portions of the plant for which the application does not seek certification to aid the NRC in its review of the FSAR and to permit assessment of the adequacy of the interface requirements under paragraph (a)(24)(i) of this section. (25) 
                                    *Technical Qualifications.*
                                     A description of the technical qualifications of the applicant to engage in the proposed activities in accordance with the regulations in this chapter. (26) 
                                    *Technical Specifications.*
                                     Proposed technical specifications prepared under § 53.710(a) for those areas addressed by the design certification. (27) 
                                    *Role of personnel.*
                                     Information to address the following for the role of personnel in ensuring safe operations: (i) A description of how the human factors engineering design requirements of § 53.440(n)(1) are addressed; (ii) A description of how the human system interface design requirements of § 53.440(n)(2) are addressed; (iii) A concept of operations that is of sufficient scope and detail to address the requirements of § 53.440(n)(3); (iv) A functional requirements analysis and function allocation that is of sufficient scope and detail to address the requirements of § 53.440(n)(4). (b) [Reserved]
 § 53.1241
  (a) In addition to the FSAR, the application must also include the following: (1) 
                                    *Environmental report.*
                                     An environmental report as required by § 51.55 of this chapter. (2) 
                                    *Availability Controls*
                                     (if not included in the FSAR). A description of the controls on plant operations, including availability controls, to provide reasonable confidence that the configurations and special treatments for SR and NSRSS SSCs provide the capabilities and reliabilities required to demonstrate compliance with the safety criteria of § 53.220, or more restrictive alternative criteria adopted under § 53.470. (3) 
                                    *Inspections, tests, analyses, and acceptance criteria.*
                                     The proposed ITAAC that are necessary and sufficient to provide reasonable assurance that, if the inspections, tests, and analyses are performed and the acceptance criteria met, a facility that incorporates the design certification has been constructed and will be operated in conformity with the design certification, the provisions of the Act, and the Commission's rules and regulations. (4) 
                                    *Safeguards information.*
                                     A description of the program to protect Safeguards Information against unauthorized disclosure in accordance with the requirements in §§ 73.21 and 73.22 of this chapter, as applicable. (b) If there are SSCs of the plant which required research and development to confirm the adequacy of their design, provide a report in the application which documents the resolution of any safety questions associated with such SSCs. (c) A description of how the performance of each design feature has been demonstrated capable of fulfilling functional design criteria considering interdependent effects through either analysis, appropriate test programs, prototype testing, operating experience, or a combination thereof, in accordance with § 53.440(a).
 § 53.1242
  (a) 
                                    *Standards for review of applications.*
                                     Applications filed under this part will be reviewed for compliance with the standards set out in this part and 10 CFR parts 20, 51, and 73. (b) 
                                    *Administrative review of applications; hearings.*
                                     (1) A standard design certification is a rule that will be issued under the provisions of subpart H of 10 CFR part 2, as supplemented by the provisions of this section. The Commission must initiate the rulemaking after an application has been filed under § 53.1233 and must specify the procedures to be used for the rulemaking. The notice of proposed rulemaking published in the 
                                    **Federal Register**
                                     must provide an opportunity for the submission of comments on the proposed design certification rule. If, at the time a proposed design certification rule is published in the 
                                    **Federal Register**
                                     under this paragraph, the Commission decides that a legislative hearing should be held, the information required by § 2.1502(c) of this chapter must be included in the 
                                    **Federal Register**
                                     document for the proposed design certification. (2) Following the submission of comments on the proposed design certification rule, the Commission may, at its discretion, hold a legislative hearing under the procedures in subpart O of part 2 of this chapter. The Commission must publish a document in the 
                                    **Federal Register**
                                     of its decision to hold a legislative hearing. The document must contain the information specified in § 2.1502(c) of this chapter and specify whether the Commission or a presiding officer will conduct the legislative hearing. (3) Notwithstanding anything in § 2.390 of this chapter to the contrary, proprietary information will be protected in the same manner and to the same extent as proprietary information submitted in connection with applications for licenses, provided that the design certification will be published in chapter I of this title. (c) 
                                    *Reference to an issued operating license or combined license.*
                                     In those cases where a design certification application is preceded by the issuance of an OL or custom COL for a commercial nuclear plant that is essentially the same as the standard design for which certification is being requested, the NRC review will follow the processes for referencing a standard design approval in § 53.1221, to the extent practicable.
 § 53.1245
  The Commission must refer a copy of the application to the ACRS. The ACRS must report on those portions of the application which concern safety.
 § 53.1248
  (a) After conducting a rulemaking proceeding under § 53.1242 on an application for a standard design certification and receiving the report to be submitted by the ACRS under § 53.1245, the Commission may issue a standard design certification in the form of a rule for the design that is the subject of the application, if the Commission determines that— (1) The application demonstrates compliance with the applicable standards and requirements of the Act and the Commission's regulations; (2) Notifications, if any, to other agencies or bodies have been duly made; (3) There is reasonable assurance that the standard design conforms with the provisions of the Act and the Commission's regulations; (4) The applicant is technically qualified; (5) The proposed ITAAC are necessary and sufficient, within the scope of the standard design, to provide reasonable assurance that, if the inspections, tests, and analyses are performed and the acceptance criteria met, the facility has been constructed and will be operated in accordance with the design certification, the provisions of the Act, and the Commission's regulations; (6) Issuance of the standard design certification will not be inimical to the common defense and security or to the health and safety of the public;
                                    ( print page 87088) (7) The findings required by subpart A of part 51 of this chapter have been made; and (8) The applicant has implemented the QAP described or referenced in the Safety Analysis Report. (b) The design certification rule must specify the site parameters, design characteristics, and any additional requirements and restrictions of the design certification rule. (c) After the Commission has adopted a final design certification rule, the applicant must not permit any individual to have access to or any facility to possess restricted data or classified National Security Information until the individual and/or facility has been approved for access under the provisions of 10 CFR parts 25 and/or 95, as applicable.
 § 53.1251
  (a) Except as provided in paragraph (b) of this section, a standard design certification issued under this subpart is valid for 15 years from the effective date of the rule. (b) A standard design certification continues to be valid beyond the date of expiration in any proceeding on an application for a COL or an OL under this part that references the standard design certification and is docketed either before the date of expiration of the certification, or, if a timely application for renewal of the certification has been filed, before the Commission has determined whether to renew the certification. A design certification also continues to be valid beyond the date of expiration in any hearing held under § 53.1452 before operation begins under a COL that references the design certification. (c) An applicant for a CP, OL, COL, or ML under this part may, at its own risk, reference in its application a design for which a design certification application has been docketed but not granted.
 § 53.1254
  (a) Not less than 12 nor more than 36 months before the expiration of the initial 15-year period, or any later renewal period, any person may apply for renewal of the certification. An application for renewal must contain all information necessary to bring up to date the information and data contained in the previous application. The Commission will require, before renewal of certification, that engineering documents, such as analyses, drawings, procurement specifications, or construction and installation specifications, be completed and available for audit if the more detailed information is necessary for the Commission to verify the information in the application and make its safety determination. Notice and comment procedures must be used for a rulemaking proceeding on the application for renewal. The Commission, in its discretion, may require the use of additional procedures in individual renewal proceedings. (b) A design certification, either original or renewed, for which a timely application for renewal has been filed remains in effect until the Commission has determined whether to renew the certification. If the certification is not renewed, it continues to be valid in certain proceedings under § 53.1251. (c) The Commission must refer a copy of the application for renewal to the ACRS. The ACRS must report on those portions of the application which concern safety and must apply the criteria set forth in § 53.1257.
 § 53.1257
  (a) The Commission must issue a rule granting the renewal if the design, either as originally certified or as modified during the rulemaking on the renewal, complies with the Act and the Commission's regulations applicable and in effect at the time the certification was issued. (b) The Commission may impose other requirements if it determines that— (1) They are necessary for adequate protection to public health and safety or common defense and security; (2) They are necessary for compliance with the Commission's regulations and orders applicable and in effect at the time the design certification was issued; or (3) There is a substantial increase in overall protection of the public health and safety or the common defense and security to be derived from the new requirements, and the direct and indirect costs of implementing those requirements are justified in view of this increased protection. (c) In addition, the applicant for renewal may request an amendment to the design certification. The Commission must grant the amendment request if it determines that the amendment will comply with the Act and the Commission's regulations in effect at the time of renewal. If the amendment request entails such an extensive change to the design certification that an essentially new standard design is being proposed, an application for a design certification must be filed in accordance with this subpart. (d) Denial of renewal does not bar the applicant, or another applicant, from filing a new application for certification of the design, which proposes design changes that correct the deficiencies cited in the denial of the renewal.
 § 53.1260
  Each renewal of certification for a standard design will be for not less than 10, nor more than 15 years.
 § 53.1263
  (a)(1) While a standard design certification rule is in effect under §§ 53.1251 or 53.1260, the Commission may not modify, rescind, or impose new requirements on the certification information, whether on its own motion, or in response to a petition from any person, unless the Commission determines in a rulemaking that the change— (i) Is necessary either to bring the certification information or the referencing plants into compliance with the Commission's regulations applicable and in effect at the time the certification was issued; (ii) Is necessary to provide adequate protection of the public health and safety or the common defense and security; (iii) Reduces unnecessary regulatory burden and maintains protection to public health and safety and the common defense and security; (iv) Provides the detailed design information to be verified under those ITAAC that are directed at certification information (
                                    *i.e.,*
                                     design acceptance criteria); (v) Is necessary to correct material errors in the certification information; (vi) Substantially increases overall safety, reliability, or security of facility design, construction, or operation, and the direct and indirect costs of implementation of the rule change are justified in view of this increased safety, reliability, or security; or (vii) Contributes to increased standardization of the certification information. (2)(i) In a rulemaking under § 53.1263(a)(1), except for § 53.1263(a)(1)(ii), the Commission will give consideration to whether the benefits justify the costs for plants that are already licensed or for which an application for a permit or license is under consideration. (ii) The rulemaking procedures for changes under § 53.1263(a)(1) must provide for notice and opportunity for public comment. (3) Any modification the NRC imposes on a design certification rule 
                                    ( print page 87089)
                                    under paragraph (a)(1) of this section will be applied to all plants referencing the certified design, except those to which the modification has been rendered technically irrelevant by action taken under paragraphs (a)(4) or (b) of this section. (4) The Commission may not impose new requirements by plant-specific order on any part of the design of a specific plant referencing the design certification rule if that part was approved in the design certification while a design certification rule is in effect under § 53.1248, unless— (i) A modification is necessary to secure compliance with the Commission's regulations applicable and in effect at the time the certification was issued, or to assure adequate protection of the public health and safety or the common defense and security; and (ii) Special circumstances as defined in § 53.080 are present. In addition to the factors listed in § 53.080, the Commission must consider whether the special circumstances which § 53.080 requires to be present outweigh any decrease in safety that may result from the reduction in standardization caused by the plant-specific order. (5) Except as provided in § 2.335 of this chapter, in making the findings required for issuance of a COL, CP, OL, or ML, or for any hearing under § 53.1452, the Commission must treat as resolved those matters resolved in connection with the issuance or renewal of a design certification rule. (b) An applicant who references a design certification rule may request an exemption from one or more elements of the certification information. The Commission may grant such a request only if it determines that the exemption will comply with the requirements of § 53.080. In addition to the factors listed in § 53.080, the Commission must consider whether the special circumstances that § 53.080 requires to be present outweigh any decrease in safety that may result from the reduction in standardization caused by the exemption. The granting of an exemption on request of an applicant is subject to litigation in the same manner as other issues in the OL or COL hearing. (c) The Commission will require, before granting a CP, COL, OL, or ML that references a design certification rule, that information normally contained in engineering documents, such as analyses, drawings, procurement specifications, or construction and installation specifications, be completed and available for audit if the more detailed information is necessary for the Commission to verify the information in the application and make its safety determination, including the determination that the application is consistent with the certification information. This information may be acquired by appropriate arrangements with the design certification applicant.
 § 53.1270
  Sections 53.1270 through 53.1295 set out the requirements and procedures applicable to Commission issuance of a license under this part authorizing manufacture of manufactured reactors to be installed at sites not identified in the ML application.
 § 53.1273
  Any person, except one excluded by § 53.1118, may file an application for an ML under this part with the Director, Office of Nuclear Reactor Regulation.
 § 53.1276
  Each application for an ML must include the information contained in § 53.1109(a) through (e), and (j).
 § 53.1279
  (a) 
                                    *Final Safety Analysis Report-siting and design.*
                                     The application must include an FSAR containing the information set forth below, with a level of design information sufficient to enable the Commission to judge the applicant's proposed means of ensuring that the manufacturing conforms to the design and to reach a final conclusion on all safety questions associated with the design, permit the preparation of construction and installation specifications by an applicant who seeks to use the manufactured reactor, and permit the preparation of acceptance and inspection requirements by the NRC. The application must include the following information: (1) 
                                    *Site parameters.*
                                     The site parameters postulated for the design under this part, including the design-basis external hazard levels for the relevant external hazards, and an analysis and evaluation of the design in terms of those site parameters. (2) 
                                    *Design information.*
                                     Except as specified in this paragraph, the design information equivalent to that required for a standard design certification as defined in § 53.1239(a)(2) through (27) for those portions of a commercial nuclear plant included in the manufactured reactor. (3) 
                                    *Quality assurance program.*
                                     A description of the QAP applied to the design, and to be applied to the fabrication and testing of the SSCs of the manufactured reactor under § 53.620(a)(6), including a discussion of how the applicable requirements of appendix B to part 50 of this chapter will be satisfied; (4) 
                                    *Conceptual designs.*
                                     Representative conceptual designs for one or more commercial nuclear plants using the manufactured reactor; (5) 
                                    *Operating configurations.*
                                     If multiple manufactured reactors may be installed at a commercial nuclear plant, a description of the possible operating configurations, including common systems, interface requirements, and system interactions. The final safety analysis must also account for differences among the possible configurations, including any restrictions that will be necessary during the construction and startup of a given manufactured reactor to ensure the safe operation of any commercial nuclear reactor already operating; (6) 
                                    *Interface requirements.*
                                     (i) The interface requirements between the manufactured reactor and the remaining portions of the commercial nuclear plant or connections to other facilities outside of the commercial nuclear plant. (ii) Confirmation that interface requirements are verifiable through inspections, testing, or analysis. These requirements must be sufficiently detailed to allow for completion of the final safety analysis by license applicants that reference the manufactured reactor manufactured under this subpart. Applicants for a COL under this part will need to verify the interface requirements at the installation site. The method to be used for verification of interface requirements must be included as part of the proposed ITAAC required by § 53.1282(a). (iii) Information to support development of radiation monitoring programs required under subpart F of this part by an applicant for a COL, including potential pathways for radionuclides produced within the manufactured reactor to enter interfacing systems. (b) 
                                    *Final Safety Analysis Report—manufacturing information.*
                                     The FSAR must include the following information related to the manufacturing processes, organization, controls, and inspections: (1) A description, including references to generally accepted consensus codes and standards, of the processes that will be used to procure, fabricate, and assemble components that make up the manufactured reactor. The description should clearly define which activities are proposed to be within the 
                                    ( print page 87090)
                                    scope of the ML and those, such as the making of a component to be procured from a separate company for installation in the manufactured reactor, that are not considered to be within the scope of the ML; (2) A description of the organizational and management structure singularly responsible for direction of design and manufacture of the manufactured reactor. The information should include a description of the management plans, technical qualifications, and controls in place to demonstrate compliance with the requirements of § 53.620; (3) A description of the inspections and tests to be performed as part of the manufacturing process, including the inspection of procured components, inspection and testing of fabrication processes such as the molding, welding, or coating of components, and inspections and testing of the assembled manufactured reactor or portions of the manufactured reactor; (4) A description of the fitness-for-duty program required by part 26 of this chapter and its implementation. (c) 
                                    *Deployment of the completed manufactured reactor.*
                                     The application must include the following information related to the deployment of a manufactured reactor: (1) Procedures governing the preparation of the manufactured reactor or portions of the manufactured reactor for shipping to the site where it is to be operated; the conduct of shipping; and verifying the condition of the shipped items upon receipt at the site; (2) Details of the interaction of the design, manufacture, and installation of a manufactured reactor within the applicant's organization and the manner by which the applicant will ensure close integration between the designer, contractors, and any facility in which the manufactured reactor is to be installed; (3) A description of the measures to be used for the control of interfaces, including the consideration of key site parameters, between the holder of the ML and the holder of the COL for the commercial nuclear plant at which the manufactured reactor is to be installed. (d) 
                                    *Special considerations for factory fueling.*
                                     In addition to the above paragraphs, an application for an ML for a manufactured reactor that will be fueled at the factory under a 10 CFR part 70 license must include the following information related to loading fuel and the required independent physical mechanisms to prevent criticality and to otherwise provide assurance that the fueled manufactured reactor can be successfully transported, installed, and operated at a site for which the Commission has issued a COL that authorizes construction and operation of a commercial nuclear plant using the manufactured reactor: (1) A description of the procedures used during the fueling of the manufactured reactor that ensure that the configuration of fuel within the fueled manufactured reactor is consistent with the design and analyses supporting operation of the manufactured reactor under the COL at the place of operation. The description may reference the applicable 10 CFR part 70 application and other sections of the Safety Analysis Report supporting the ML license application. (i) The application must describe the measures taken for in-factory inspections and non-nuclear testing performed to ensure that the configuration of fuel within the fueled manufactured reactor is consistent with the design and analyses supporting operation of the manufactured reactor under the COL at the place of operation. (ii) The application must describe the design features included in the manufactured reactor to prevent criticality, including at least two independent mechanisms each of which is sufficient to prevent criticality, the associated functional design criteria applied to those design features, and the physical and programmatic controls implemented during manufacturing, storage, and transport that are credited to assure the features function as designed when subject to potential hazards and human errors. The descriptions must include how those measures will be controlled during installation under the ML and removal under the COL at the place of operation. (2) A description of the procedures governing the transfer of responsibilities for the fueled manufactured reactor from the holder of the ML to the holder of the COL for the installation site. (3) If available at the time of filing the ML application or, if not available at the time of filing the ML application, submitted as an amendment to the ML or ML application at the time of filing the Part 70 application, a description of the programs needed to demonstrate compliance with the requirements of § 53.620(d) and 10 CFR parts 70, 71, and 73 for the receipt, storage, and loading of SNM into a manufactured reactor and the transport of the fueled manufactured reactor to a site for which the Commission has issued a COL that authorizes construction and operation of a commercial nuclear plant using the manufactured reactor, including the following. (i) A physical security program in accordance with § 53.620(d)(2)(i). (ii) A cybersecurity program in accordance with § 53.620(d)(2)(i).
 § 53.1282
  (a) 
                                    *Inspections, tests, analyses, and acceptance criteria.*
                                     (1) The application must contain proposed inspections, tests, and analyses that the COL holder must perform, and the acceptance criteria that are necessary and sufficient to provide reasonable assurance that, if the inspections, tests, and analyses are performed and the acceptance criteria met: (i) The reactor has been manufactured in conformity with the ML, the provisions of the Act, and the Commission's rules and regulations; and (ii) The manufactured reactor will be operated in conformity with the approved design and any license authorizing operation of the manufactured reactor. (2) If the application references a standard design certification, the ITAAC contained in the certified design must apply to those portions of the facility design that are covered by the design certification. (3) If the application references a standard design certification, the application may include a notification that a required inspection, test, or analysis in the design certification ITAAC has been successfully completed and that the corresponding acceptance criterion has been met. The 
                                    **Federal Register**
                                     notification required by § 53.1285 must indicate that the application includes this notification. (b) 
                                    *Environmental report.*
                                     (1) The application must contain an environmental report as required by § 51.54 of this chapter. (2) If the ML application references a standard design certification, the environmental report need not contain a discussion of severe accident mitigation design alternatives for the manufactured reactor as used in a commercial nuclear plant. (c) 
                                    *Safeguards information.*
                                     The application must contain a description of the program to protect safeguards information against unauthorized disclosure in accordance with the requirements in §§ 73.21 and 73.22 of this chapter, as applicable. (d) 
                                    *Performance demonstration.*
                                     A description of how the performance of each design feature has been demonstrated capable of fulfilling functional design criteria considering interdependent effects through either analysis, appropriate test programs, prototype testing, operating experience, 
                                    ( print page 87091)
                                    or a combination thereof, in accordance with § 53.440(a).
 § 53.1285
  (a) 
                                    *Standards for review of applications.*
                                     Applications for MLs under this part will be reviewed according to the applicable standards set out in this subpart as well as applicable standards in this part and 10 CFR parts 20, 25, 26, 51, 70, 71, 73, and 75. (b) 
                                    *Administrative review of applications, hearings.*
                                     A proceeding on an ML is subject to all applicable procedural requirements contained in 10 CFR part 2, including the requirements for docketing in § 2.101(a)(1) through (4) of this chapter, and the requirements for issuance of a notice of proposed action in § 2.105 of this chapter, 
                                    *provided, however,*
                                     that the designated sections may not be construed to require that the environmental report or draft or final environmental impact statement include an assessment of the benefits of constructing and/or operating the manufactured reactor or an evaluation of alternative energy sources. All hearings on MLs are governed by the hearing procedures contained in 10 CFR part 2, subparts C, E, G, L, and N.
 § 53.1286
  The Commission must refer a copy of the application to the ACRS. The ACRS must report on those portions of the application which concern safety.
 § 53.1287
  (a) After completing any hearing under § 53.1285(b), and receiving the report submitted by the ACRS, the Commission may issue an ML if the Commission finds that— (1) Applicable standards and requirements of the Act and the Commission's regulations have been met; (2) There is reasonable assurance that the manufactured reactor will be manufactured, and can be transported, incorporated into a commercial nuclear plant, and operated in conformity with the ML, the provision of the Act, and the Commission's regulations; (3) The proposed manufactured reactor can be incorporated into a commercial nuclear plant and operated at sites having characteristics that fall within the site parameters postulated for the design of the manufactured reactors without undue risk to the health and safety of the public; (4) The applicant is technically qualified to design and manufacture the proposed manufactured reactor; (5) The proposed ITAAC are necessary and sufficient, within the scope of the ML, to provide reasonable assurance that the manufactured reactor has been manufactured and will be operated in conformity with the license, the provisions of the Act, and the Commission's regulations; (6) The issuance of a license to the applicant will not be inimical to the common defense and security or to the health and safety of the public; and (7) The findings required by subpart A of 10 CFR part 51 have been made. (b) Each ML issued under this subpart must specify— (1) Terms and conditions as the Commission deems necessary and appropriate; (2) Technical specifications for operation of the manufactured reactor, as the Commission deems necessary and appropriate; (3) Site parameters and design characteristics for the manufactured reactor; (4) The interface requirements to be met by the site-specific elements of the facility, such as the energy conversions systems and ultimate heat sink, not within the scope of the manufactured reactor; and (5) The entity with design authority for the manufactured reactor covered by the license.
 § 53.1288
  (a)(1) Notwithstanding any provision in § 53.1590, during the term of an ML issued under this part the Commission may not modify, rescind, or impose new requirements on the design of the manufactured reactor, or the requirements for the manufacture of the manufactured reactor, unless the Commission determines that a modification is necessary to bring the design of the reactor or its manufacture into compliance with the Commission's requirements applicable and in effect at the time the ML was issued, or to provide reasonable assurance of adequate protection to public health and safety or common defense and security. (2) Any modification to the design of a manufactured reactor that is imposed by the Commission under paragraph (a)(1) of this section will be applied to all manufactured reactors manufactured under the license, including those that have already been transported and sited, except those manufactured reactors to which the modification has been rendered technically irrelevant by action taken under § 53.1530 or paragraph (b) of this section. (3) In making the findings required under this part for issuance of a COL, in any hearing under § 53.1452, or in any enforcement hearing other than one initiated by the Commission under paragraph (a)(1) of this section, for which a manufactured reactor manufactured under this subpart is referenced or used, the Commission must treat as resolved those matters resolved in the proceeding on the application for issuance or renewal of the ML, including the adequacy of design of the manufactured reactor, the costs and benefits of severe accident mitigation design alternatives, and the bases for not incorporating severe accident mitigation design alternatives into the design of the manufactured reactor to be manufactured. (b) An applicant who references or uses a manufactured reactor manufactured under an ML under this part may include in the application a request for a departure from the design characteristics, site parameters, terms and conditions, or approved design of the manufactured reactor. The Commission may grant a request only if it determines that the departure will comply with the requirements of § 53.080, and that the special circumstances outweigh any decrease in safety that may result from the reduction in standardization caused by the departure. The granting of a departure on request of an applicant is subject to litigation in the same manner as other issues in the COL hearing.
 § 53.1291
  An ML issued under this part is valid for not less than 5, nor more than 15 years from the date of issuance. Upon expiration of the ML, the manufacture of any uncompleted manufactured reactors must cease unless a timely application for renewal has been docketed with the NRC.
 § 53.1293
  An ML may be transferred under § 53.1570.
 § 53.1295
  (a)(1) Not less than 12 months, nor more than 5 years before the expiration of the ML, or any later renewal period, the holder of the ML issued under this part may apply for a renewal of the license. An application for renewal must contain all information necessary to bring up to date the information and data contained in the previous application. (2) The filing of an application for a renewed license must be in accordance with subpart A of 10 CFR part 2 and § 53.1100.
                                    ( print page 87092) (3) An ML issued under this part, either original or renewed, for which a timely application for renewal has been filed, remains in effect until the Commission has made a final determination on the renewal application, 
                                    *provided, however,*
                                     that the holder of an ML may not begin manufacture of a manufactured reactor less than 6 months before the expiration of the license. (4) Any person whose interest may be affected by renewal of the license may request a hearing on the application for renewal. The request for a hearing must comply with § 2.309 of this chapter. If a hearing is granted, notice of the hearing will be published in accordance with § 2.104 of this chapter. (5) The Commission must refer a copy of the application for renewal to the ACRS. The ACRS must report on those portions of the application which concern safety. (b) The Commission may grant the renewal if the Commission determines— (1) The ML complies with the Act and the Commission's regulations and orders applicable and in effect at the time the ML was originally issued; and (2) Any new requirements the Commission may wish to impose are— (i) Necessary for adequate protection to public health and safety or common defense and security; (ii) Necessary for compliance with the Commission's regulations and orders applicable and in effect at the time the ML was originally issued; or (iii) A substantial increase in overall protection of the public health and safety or the common defense and security to be derived from the new requirements, and the direct and indirect costs of implementation of those requirements are justified in view of this increased protection. (c) A renewed ML may be issued for a term of not less than 5, nor more than 15 years, plus any remaining years on the ML then in effect before renewal. The renewed license must be subject to the requirements of § 53.1288.
 § 53.1300
  Sections 53.1300 through 53.1348 set out the requirements and procedures applicable to Commission issuance of a CP for commercial nuclear plants. A CP for the construction of a commercial nuclear plant under this part will be issued before the issuance of an OL if the application is otherwise acceptable and will be converted upon completion of the facility and Commission action, into an OL as provided under §§ 53.1360 through 53.1405.
 § 53.1306
  An application for a CP must include the information required by § 53.1109 and the following information: (a) Information sufficient to demonstrate to the Commission the financial qualification of the applicant to carry out, under the regulations in this chapter, the activities for which the permit is sought. As applicable, the following should be provided: (1) The information that demonstrates that the applicant possesses or has reasonable assurance of obtaining the funds necessary to cover estimated construction costs and related fuel cycle costs, including estimates of the total construction costs and related fuel cycle costs of the facility and must indicate the source(s) of funds to cover these costs. (2) Each application for a CP submitted by a newly formed entity organized for the primary purpose of constructing and operating a facility must also include information showing: (i) The legal and financial relationships the entity has or proposes to have with its stockholders or owners; (ii) The stockholders' or owners' financial ability to meet any contractual obligation to the entity that they have incurred or proposed to incur; and (iii) Any other information considered necessary by the Commission to enable it to determine the applicant's financial qualification; and (3) The Commission may request an established entity or newly-formed entity to submit additional or more detailed information respecting its financial arrangements and status of funds if the Commission considers this information appropriate. This may include information regarding an applicant's ability to continue the conduct of the activities authorized by the CP and to decommission the facility. (b) If the applicant proposes to construct or alter a facility, the application must state the earliest and latest dates for completion of the construction or alteration.
 § 53.1309
  The application must contain a Preliminary Safety Analysis Report (PSAR) that describes the facility and the limits on its operation and presents a preliminary safety analysis of the SSCs of the facility as a whole. The PSAR must include the following information, at a level of detail sufficient to enable the Commission to reach a conclusion on safety matters that must be resolved by the Commission before issuance of a CP: (a)(1) 
                                    *Site information.*
                                     An application for a CP for a commercial nuclear reactor must include the site information equivalent to that required for an early site permit in § 53.1146(a)(1)(iv) through (x). (2) 
                                    *Design information.*
                                     Except as specified in this paragraph, an application for a CP for a commercial nuclear plant must include the design information equivalent to that required for a standard design certification as defined in § 53.1239(a)(2) through (27). (i) 
                                    *Quality assurance program.*
                                     A description of the QAP to be applied to the design, fabrication, construction, and testing of the SSCs of the facility under § 53.610(a)(6), including a discussion of how the requirements of appendix B to part 50 of this chapter will be satisfied. (ii) 
                                    *Preliminary design information.*
                                     The information provided in the application may include some aspects of the design that are not fully developed, and the information is therefore preliminary. The completed design, including any changes during construction, must be described in the FSAR required in § 53.1369 that supports an application for an OL. (iii) 
                                    *Planned research or testing.*
                                     Descriptions of how design features and related functional design criteria will fulfill the safety criteria in subpart B, or more restrictive alternative criteria adopted under § 53.470, and how that has been or will be demonstrated through either analysis, appropriate test programs, experience, or a combination thereof. Where any design feature has not been fully developed or demonstrated to fulfill the functional design criteria at the time of an application for a CP, the applicant must provide a plan for future analysis, research and development, test programs, gathering of experience, or a combination thereof to provide reasonable confidence that the required demonstration will be available for an application for an OL. (iv) 
                                    *Programmatic controls.*
                                     Descriptions of the programmatic controls may include those to be provided in the FSAR or other licensing basis documents because they are necessary to achieve and maintain the reliability and capability of SSCs relied upon to demonstrate compliance with the established safety criteria and functional design criteria required in subpart B, and to maintain consistency with analyses required by § 53.450. (3) 
                                    *Technical qualifications.*
                                     A description of the technical qualifications of the applicant to engage in the proposed activities under the regulations in this chapter.
                                    ( print page 87093) (4) 
                                    *Emergency preparedness.*
                                     A description of the applicant's preliminary plans for coping with emergencies based on: (i) Except as provided in paragraph (a)(4)(ii) of this section, the requirements in appendix E to part 50. (ii) For a commercial nuclear plant consisting of either small modular reactors or non-light-water reactors, the requirements in either § 50.160 or appendix E to part 50. (5) 
                                    *Physical security.*
                                     A report that provides a preliminary description of how the site characteristics support the development of adequate security plans and measures consistent with the requirements in § 53.540. (6) 
                                    *Fitness-for-duty program.*
                                     A description of the fitness-for-duty (FFD) program required by 10 CFR part 26 and its implementation. (b) A description of the program to protect Safeguards Information against unauthorized disclosure in accordance with the requirements in §§ 73.21 and 73.22 of this chapter, as applicable.
 § 53.1312
  (a) In addition to the PSAR, the application must include the following: (1) An environmental report either under § 51.50(a) of this chapter if an LWA under § 53.1130 is not requested in conjunction with the CP application, or under §§ 51.49 and 51.50(a) of this chapter if an LWA is requested in conjunction with the CP application; or (2) If the applicant wishes to request that an LWA under § 53.1130 be issued before issuance of the CP, the information otherwise required by § 53.1130, in accordance with either § 2.101(a)(1) through (a)(5), or § 2.101(a)(9) of this chapter. (b) If the CP application references an early site permit, standard design approval, or standard design certification issued under this part, then the following requirements apply: (1) The PSAR need not contain information or analyses submitted to the Commission in connection with the referenced NRC approval, permit, or certification, provided, however, that the PSAR incorporates the material by reference and confirms that the site and design of the facility falls within parameter values postulated in the referenced NRC approval, permit, or certification. (2) The PSAR must provide a means to demonstrate that all terms and conditions that have been included in the referenced NRC approval, permit, or certification will be satisfied by the date of issuance of the OL, as appropriate. If the PSAR does not demonstrate that each site characteristic falls within the corresponding postulated site parameter and each design characteristic of the facility falls within the corresponding postulated design parameter, the application must justify a departure, variance, or exemption from the referenced NRC approval, license, or certification in regard to that particular site or design characteristic in compliance with the requirements of this part. (3) If a referenced early site permit approves complete and integrated emergency plans, or major features of emergency plans, then the PSAR must include any new or additional information that updates and corrects the information that was provided under § 53.1146(b)(2) and discuss whether the new or additional information materially changes the bases for compliance with the applicable requirements.
 § 53.1315
  (a) 
                                    *Standards for review of applications.*
                                     Applications filed under this part will be reviewed according to the standards set out in this part and 10 CFR parts 20, 51, 73, and 140. (b) 
                                    *Administrative review of applications; hearings.*
                                     A proceeding on a CP application is subject to all applicable procedural requirements contained in 10 CFR part 2, including the requirements for docketing (§ 2.101 of this chapter) and issuance of a notice of hearing (§ 2.104 of this chapter). All hearings on CP applications are governed by the procedures contained in 10 CFR part 2.
 § 53.1318
  If the application for a CP under this part references an early site permit, standard design approval, or standard design certification, the scope and nature of matters resolved for the application are governed by the relevant provisions addressing finality, including §§ 53.1188, 53.1221, and 53.1263.
 § 53.1324
  The Commission must refer a copy of the application to the ACRS. The ACRS must report on those portions of the application that concern safety and must apply the standards referenced in § 53.1315, in accordance with the finality provisions in § 53.1318.
 § 53.1327
  (a) If the application does not reference an early site permit which authorizes the holder to perform the activities under § 53.1130, the applicant may not perform those activities without obtaining the separate authorization required by § 53.1130. Authorization may be granted only after the presiding officer in the proceeding on the application has made the findings and determination required by § 53.1130(b)(1)(ii) and (iv), and the Director, Office of Nuclear Reactor Regulation makes the determination required by § 53.1130(b)(1)(iii). (b) If, after an applicant has performed the activities permitted by paragraph (a) of this section, the application for the CP is withdrawn or denied, then the applicant must implement an approved site redress plan.
 § 53.1330
  (a) Applicants for a CP under this part, or any amendment to a CP, may include in the application a request for an exemption from one or more of the Commission's regulations. The Commission may grant a request if it determines that the exemption complies with § 53.080. (b) An applicant for a CP who has filed an application referencing an NRC approval, permit, or certification issued under this part may include in the application a request for exemptions, departures, or variances related to the subject referenced NRC approval, permit, or certification. In determining whether to grant the departure, variance, or exemption, the Commission must apply the same technically relevant criteria as were applicable to the application for the original or renewed approval, license, or certification.
 § 53.1333
  (a) After conducting a hearing in accordance with § 53.1315 and receiving the report submitted by the ACRS, the Commission may issue a CP only if the Commission finds that— (1) The applicant has described the proposed design of the facility and has identified the major features or components incorporated therein for the protection of the health and safety of the public; (2) Such further technical or design information as may be required to complete the safety analysis, and which can reasonably be left for later consideration, will be supplied in the FSAR; (3) Safety features or components, if any, that require research and development have been described by the applicant and the applicant has identified, and there will be conducted, 
                                    ( print page 87094)
                                    a research and development program reasonably designed to resolve any safety questions associated with such features or components; and (4) On the basis of the foregoing, there is reasonable assurance of the following— (i) Such safety questions will be satisfactorily resolved at or before the latest date stated in the application for completion of construction of the proposed facility; and (ii) Taking into consideration the site criteria contained subpart D to this part, the proposed facility can be constructed and operated at the proposed location without undue risk to the health and safety of the public. (b) A CP must contain the terms and conditions for the permit, as the Commission deems necessary and appropriate. The Commission may, in its discretion, incorporate in any CP provisions requiring the applicant to furnish periodic reports of the progress and results of research and development programs designed to resolve safety questions.
 § 53.1336
  Notwithstanding any provision in § 53.1590, a CP constitutes an authorization to proceed with construction but does not constitute Commission approval of the safety of any design feature or specification unless the applicant specifically requests such approval and such approval is incorporated in the permit. The applicant, at its option, may request such approvals in the CP or by amendment to the CP. If approved by the NRC and included in the permit, the NRC will consider modifications to the approved design features or specifications in accordance with § 53.1590.
 § 53.1342
  (a) A CP will state the earliest and latest dates for completion of construction or alteration of the facility, not to exceed 40 years from date of issuance. (b) If the proposed construction or alteration of the facility is not completed by the latest completion date, the CP shall expire, and all rights are forfeited. However, upon good cause shown, the Commission will extend the completion date for a reasonable period of time. The Commission will recognize, among other things, developmental problems attributed to the experimental nature of the facility or fire, flood explosion, strike, sabotage, domestic violence, enemy action an act of the elements, and other acts beyond the control of the permit holder, as a basis for extending the completion date.
 § 53.1345
  A CP may be transferred under § 53.1570.
 § 53.1348
  When a permit holder has determined to permanently cease construction, the holder must, within 30 days, submit a written certification to the NRC.
 § 53.1360
  Sections 53.1360 through 53.1405 set out the requirements and procedures applicable to Commission issuance of an OL for a nuclear power facility.
 § 53.1366
  An application for an OL must include the information required by § 53.1109 and the following information: (a) Except for an electric utility applicant, information sufficient to demonstrate to the Commission the financial qualification of the applicant to carry out, in accordance with the regulations in this chapter, the activities for which the license is sought. As applicable, the following should be provided: (1) The applicant must submit information that demonstrates the applicant possesses or has reasonable assurance of obtaining the funds necessary to cover estimated operation costs for the period of the license. The applicant must submit estimates for total annual operating costs for each of the first 5 years of operation of the facility. The applicant must also indicate the source(s) of funds to cover these costs. (2) Each application for an OL submitted by a newly-formed entity organized for the primary purpose of operating the facility must also include information showing— (i) The legal and financial relationships the entity has or proposes to have with its stockholders or owners; (ii) The stockholders' or owners' financial ability to meet any contractual obligation to the entity which they have incurred or proposed to incur; and (iii) Any other information considered necessary by the Commission to enable it to determine the applicant's financial qualification. (3) The Commission may request an established entity or newly formed entity to submit additional or more detailed information respecting its financial arrangements and status of funds if the Commission considers this information appropriate. This may include information regarding a licensee's ability to continue the conduct of the activities authorized by the license and to decommission the facility. (b) The application must include information in the form of a report, as described in subpart G, indicating how reasonable assurance will be provided that funds will be available to decommission the facility, including a copy of the financial instrument obtained to satisfy the requirements of § 53.1040.
 § 53.1369
  *Final Safety Analysis Report.*
                                     The application must contain an FSAR that describes the facility and the limits on its operation and presents a safety analysis of the SSCs of the facility as a whole. The FSAR must include the following information, at a level of detail sufficient to enable the Commission to reach a final conclusion on all safety matters that must be resolved by the Commission before issuance of an OL. The FSAR must include the following information: (a) 
                                    *Site information.*
                                     An application for an OL for a commercial nuclear reactor must include the site information equivalent to that required for an early site permit in § 53.1146(a)(1)(iv) through (x), including all current information, such as the results of environmental and meteorological monitoring programs, which has been developed since issuance of the CP, relating to site evaluation factors identified in this part. (b) 
                                    *Design information.*
                                     Except as specified in this paragraph, an FSAR for an OL for a commercial nuclear plant must include the final design information equivalent to that required for a standard design certification as defined in § 53.1239(a)(2) through (7), (a)(9), and (a)(11) through (a)(27). (1) The completed design, including any changes during construction, must be described. (2) Where any design feature had not been fully developed or demonstrated at the time of application for the CP, the applicant must provide the analysis, research and development, test programs, gathering of experience, or a combination thereof to provide the required demonstration to fulfill the functional design criteria. (c) 
                                    *Technical qualifications.*
                                     A description of the technical qualifications of the applicant to engage in the proposed activities in accordance with the regulations in this chapter.
                                    ( print page 87095) (d) 
                                    *Integrity assessment program.*
                                     A description of an Integrity Assessment Program that addresses the elements described in § 53.870. (e) 
                                    *Safeguards information.*
                                     A description of the program to protect Safeguards Information against unauthorized disclosure in accordance with the requirements in §§ 73.21 and 73.22 of this chapter, as applicable. (f) 
                                    *Emergency response facility or facilities.*
                                     Description of location and capabilities to be established for command and control, support, and coordination of onsite and offsite, as applicable, functions during reactor accident conditions. (g) 
                                    *Role of personnel.*
                                     (1) A description of the completed assessments related to the role of personnel in ensuring safe operations considering the analyses required by § 53.730. These assessments must include the following: (i) Human factors engineering design requirements of § 53.730(a); (ii) Human system interface design requirements of § 53.730(b); (iii) Concept of operations of § 53.730(c); (iv) Functional requirements analysis and function allocation of § 53.730(d); (2) A description of the program to be used for evaluating and applying operating experience as required by § 53.730(e); (3) A staffing plan and supporting analyses as required by § 53.730(f). (h) 
                                    *Training, examination, and proficiency programs.*
                                     (1) A description of the training, examination, and proficiency programs required by § 53.730(g); (2) A description of the training programs required by § 53.830. (i) 
                                    *Emergency plan.*
                                     Emergency plans complying with the requirements of § 53.855. (1) Include all emergency plan certifications, as applicable, that have been obtained from the State, local, and participating Tribal governmental agencies with emergency planning responsibilities that are wholly or partially within the EPZ plume exposure pathway. These certifications must state that— (i) The proposed emergency plans are practicable; (ii) These agencies are committed to participating in any further development of the plans, including any required field demonstrations; and (iii) These agencies are committed to executing their responsibilities under the plans in the event of an emergency. (2) If certifications cannot be obtained after sustained, good faith efforts by the applicant, then the application must contain information, including a utility plan, sufficient to show that the proposed plans provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency at the site. (3) If complete and integrated emergency plans were approved as part of an early site permit, or submitted, reviewed, and approved as part of the CP application, new certifications that demonstrate compliance with the requirements of paragraph (i)(1) of this section are not required. (j) 
                                    *Organization.*
                                     A description of the applicant's organizational structure, allocations of responsibilities and authorities, and personnel qualifications requirements for operation. (k) 
                                    *Maintenance program.*
                                     A description of a maintenance program under § 53.715. (l) 
                                    *Quality assurance.*
                                     A description of the QAP that demonstrates compliance with the requirements under § 53.865. (m) 
                                    *Radiation protection program.*
                                     A radiation protection program description under § 53.850. (n) 
                                    *Security program.*
                                     A physical security plan that describes how the applicant will comply with § 53.860 (and 10 CFR part 11, if applicable, including the identification and description of jobs as required by § 11.11(a) of this chapter, at the proposed facility). The plan must list tests, inspections, audits, and other means to be used to demonstrate compliance with the requirements of 10 CFR parts 11 and 73, if applicable. (o) 
                                    *Safeguards contingency plan.*
                                     A safeguards contingency plan in accordance with the criteria set forth in appendix C to 10 CFR part 73. The safeguards contingency plan must include plans for dealing with threats, thefts, and radiological sabotage, as defined in 10 CFR part 73, relating to the SNM and nuclear facilities licensed under this chapter and in the applicant's possession and control. Each application for this type of license must include the information contained in the applicant's safeguards contingency plan. (Implementing procedures required for this plan need not be submitted for approval.) 1 (p) 
                                    *Security training and qualification.*
                                     A training and qualification plan that describes how the applicant will demonstrate compliance with the criteria set forth in § 73.100 of this chapter or appendix B to 10 CFR part 73. (q) 
                                    *Cybersecurity plan.*
                                     A cybersecurity plan in accordance with the criteria set forth in § 73.54 or § 73.110 of this chapter. (r) 
                                    *Security, safeguards and cybersecurity plan implementation.*
                                     A description of the implementation of the physical security plan, safeguards contingency plan, training and qualification plan, and cybersecurity plan. Each applicant who prepares a physical security plan, a safeguards contingency plan, a training and qualification plan, or a cybersecurity plan must protect the plans and other related Safeguards Information against unauthorized disclosure in accordance with the requirements of §§ 73.21 and 73.22 of this chapter. (s) 
                                    *Fire protection program.*
                                     A description of the fire protection program under § 53.875. (t) 
                                    *Inservice inspection/inservice testing program.*
                                     A description of the inservice inspection and inservice testing programs under § 53.880. (u) [Reserved] (v) [Reserved] (w) 
                                    *General employee training.*
                                     A description of the training program required to demonstrate compliance with § 53.830 and its implementation. (x) 
                                    *Fitness-for-duty program.*
                                     A description of the FFD program required by 10 CFR part 26 and its implementation. (y) 
                                    *Other programs.*
                                     A description and evaluation of the results of the applicant's programs, including research and development, if any, to demonstrate that any safety questions identified at the CP stage have been resolved. (z) 
                                    *Safety design feature performance.*
                                     A description of how the performance of each safety design feature has been demonstrated capable of fulfilling functional design criteria considering interdependent effects through either analysis, appropriate test programs, prototype testing, operating experience, or a combination thereof, in accordance with § 53.440(a). (aa) 
                                    *Technical specifications.*
                                     Proposed technical specifications prepared in accordance with the requirements of § 53.710(a). 1
                                         A physical security plan that contains all the information required in both § 73.55 or § 73.100 of this chapter and appendix C to 10 CFR part 73 satisfies the requirement for a contingency plan.
 § 53.1372
  In addition to the FSAR, the application must also include the following: (a) 
                                    *Environmental report.*
                                     An environmental report in accordance with § 51.53(b) of this chapter. (b) 
                                    *Availability controls (if not included in the FSAR).*
                                     A description of 
                                    ( print page 87096)
                                    the controls on plant operations, including availability controls, to provide reasonable confidence of safe operation and that the configurations and special treatments for SR and NSRSS SSCs provide the capabilities and reliabilities required to satisfy the safety criteria of § 53.220, or more restrictive alternative criteria adopted under § 53.470, if not addressed by Technical Specifications under § 53.1369(aa).
 § 53.1375
  (a) 
                                    *Standards for review of applications.*
                                     Applications filed under this part will be reviewed according to the standards set out in 10 CFR parts 20, 26, 51, 53, 73, and 140. (b) 
                                    *Administrative review of applications; hearings.*
                                     A proceeding on an OL is subject to all applicable procedural requirements contained in 10 CFR part 2, including the requirements for docketing (§ 2.101 of this chapter) and issuance of a notice of hearing (§ 2.104 of this chapter). All hearings on OLs are governed by the procedures contained in 10 CFR part 2.
 § 53.1381
  The Commission must refer a copy of the application to the ACRS. The ACRS must report on those portions of the application that concern safety and must apply the standards referenced in § 53.1375.
 § 53.1384
  (a) Applicants for an OL under this part, or any amendment to an OL, may include in the application a request for an exemption from one or more of the Commission's regulations. The Commission may grant an exemption request if it determines that the exemption complies with § 53.080. (b) An applicant for an OL who has filed an application referencing an NRC approval, permit, license, or certification issued under this part may include in the application a request for departures, variances, or exemptions related to the subject referenced NRC approval, permit, license, or certification. In determining whether to grant the departure, variance, or exemption, the Commission must apply the same technically relevant criteria as were applicable to the application for the original or renewed approval, license, or certification.
 § 53.1387
  Upon completion of the construction or alteration of a facility, in compliance with the terms and conditions of the construction permit and subject to any necessary testing of the facility for health or safety purposes, the Commission will, in the absence of good cause shown to the contrary, issue an OL or an appropriate amendment of the license, as the case may be. (a)(1) After receiving the report submitted by the ACRS, the Commission may issue an OL if the Commission finds that— (i) Construction of the facility has been substantially completed in conformity with the CP and the application as amended, the provisions of the Act, and the rules and regulations of the Commission; (ii) Any required notifications to other agencies or bodies have been duly made; (iii) The facility will operate in conformity with the application as amended, the provisions of the Act, and the rules and regulations of the Commission; (iv) There is reasonable assurance that— (A) the activities authorized by the OL can be conducted without endangering the health and safety of the public; and (B) such activities will be conducted in compliance with the regulations in this chapter. (v) The applicant is technically and financially qualified to engage in the activities authorized, however, no finding of financial qualification is necessary for an electric utility applicant for an OL; (vi) Issuance of the license will not be inimical to the common defense and security or to the health and safety of the public; (vii) The applicable provisions of 10 CFR part 140 have been satisfied; and (viii) The findings required by subpart A of 10 CFR part 51 have been made. (2) [Reserved] (b) [Reserved] (c) The OL will include appropriate provisions with respect to any uncompleted items of construction and such limitations or conditions as are required to assure that operation during the period of the completion of such items will not endanger public health and safety. (d) The Commission will issue an OL in such form and containing such conditions and limitations, including technical specifications, as it deems necessary and appropriate.
 § 53.1390
  After issuance of an OL, the Commission may not modify, add, or delete any term or condition of the OL, except in accordance with the provisions of § 53.1590.
 § 53.1396
  The Commission will issue an OL under this part for the term requested by the applicant, not to exceed 40 years from the date of issuance, or for the estimated useful life of the facility if the Commission determines that the estimated useful life is less than the term requested.
 § 53.1399
  An OL may be transferred under § 53.1570.
 § 53.1402
  The filing of an application for a renewed license must be in accordance with § 53.1595.
 § 53.1405
  Each OL for a facility that has permanently ceased operations continues in effect beyond the expiration date to authorize ownership and possession of the facility until the Commission notifies the licensee in writing that the license is terminated. During this period of continued effectiveness, the licensee must— (a) Take actions necessary to decommission and decontaminate the facility and continue to maintain the facility, including, where applicable, the storage, control, and maintenance of the spent fuel in a safe condition; and (b) Conduct activities in accordance with all other restrictions applicable to the facility in accordance with the NRC's regulations and the provisions of the OL for the facility.
 § 53.1410
  Sections 53.1410 through 53.1461 set out the requirements and procedures applicable to Commission issuance of COLs for commercial nuclear plants under this part.
 § 53.1413
  An application for a COL must include the information required by § 53.1109 and the following information: (a) Except for an electric utility applicant in regard to financial assurance required after a Commission finding under § 53.1452, the application must include information sufficient to demonstrate to the Commission the financial qualification of the applicant to carry out, in accordance with the regulations in this chapter, the activities for which the permit or license is sought. As applicable, the following should be provided: (1) The applicant must submit information that demonstrates that the applicant possesses or has reasonable 
                                    ( print page 87097)
                                    assurance of obtaining the funds necessary to cover estimated construction costs and related fuel cycle costs. The applicant must submit estimates of the total construction costs of the facility and related fuel cycle costs and must indicate the source(s) of funds to cover these costs. (2) The applicant must submit information that demonstrates the applicant possesses or has reasonable assurance of obtaining the funds necessary to cover estimated operation costs for the period of the license. The applicant must submit estimates for total annual operating costs for each of the first 5 years of operation of the facility. The applicant must also indicate the source(s) of funds to cover these costs. (3) Each application for a COL submitted by a newly-formed entity organized for the primary purpose of constructing and operating a facility must also include information showing— (i) The legal and financial relationships the entity has or proposes to have with its stockholders or owners; and (ii) The stockholders' or owners' financial ability to meet any contractual obligation to the entity which they have incurred or proposed to incur; and (iii) Any other information considered necessary by the Commission to enable it to determine the applicant's financial qualification. (4) The Commission may request an established entity or newly formed entity to submit additional or more detailed information respecting its financial arrangements and status of funds if the Commission considers this information appropriate. This may include information regarding a licensee's ability to continue the conduct of the activities authorized by the license and to decommission the facility. (b) The application must include information in the form of a report, as described in subpart G of this part, indicating how reasonable assurance will be provided that funds will be available to decommission the facility.
 § 53.1416
  (a) 
                                    *Final Safety Analysis Report.*
                                     The application must contain an FSAR that describes the facility and the limits on its operation and presents a safety analysis of the SSCs of the facility as a whole. The Commission will require, before issuance of a COL, that engineering documents, such as analyses, drawings, procurement specifications, or construction and installation specifications, be completed and available for audit if the more detailed information is necessary for the Commission to verify the information in the application and make its safety determination. The FSAR must include the following information, at a level of detail sufficient to enable the Commission to reach a final conclusion on all safety matters that must be resolved by the Commission before issuance of a COL: (1) 
                                    *Site information.*
                                     An application for a COL for a commercial nuclear reactor must include the site information required for an early site permit in § 53.1146(a)(1)(iv) through (x). (2) 
                                    *Design information.*
                                     An application for a COL for a commercial nuclear plant must include the design information equivalent to that required for a standard design certification as defined in § 53.1239(a)(2) through (7), (a)(9), and (a)(11) through (27). (3) 
                                    *Technical qualifications.*
                                     A description of the technical qualifications of the applicant to engage in the proposed activities in accordance with the regulations in this chapter. (4) 
                                    *Integrity assessment program.*
                                     A description of an Integrity Assessment Program that addresses the elements described in § 53.870. (5) 
                                    *Safeguards information.*
                                     A description of the program to protect Safeguards Information against unauthorized disclosure in accordance with the requirements in §§ 73.21 and 73.22 of this chapter, as applicable. (6) 
                                    *Emergency response facility or facilities.*
                                     Description of the locations and capabilities to be established for command and control, support, and coordination of onsite and offsite, as applicable, functions during reactor accident conditions. (7) 
                                    *Role of personnel.*
                                     (i) A description of the completed assessments related to the role of personnel in ensuring safe operations considering the analyses required by § 53.730. These assessments must include the following: (A) Human factors engineering design requirements of § 53.730(a); (B) Human system interface design requirements of § 53.730(b); (C) Concept of operations of § 53.730(c); and (D) Functional requirements analysis and function allocation of § 53.730(d); (ii) A description of the program to be used for evaluating and applying operating experience as required by § 53.730(e); (iii) A staffing plan and supporting analyses as required by § 53.730(f). (8) 
                                    *Training, examination, and proficiency programs.*
                                     (i) A description of the training, examination, and proficiency programs required by § 53.730(g); and (ii) A description of the training programs required by § 53.830. (9) 
                                    *Emergency plan.*
                                     Emergency plans complying with the requirements of § 53.855. (i) The emergency plan must include, as applicable, all emergency plan certifications that have been obtained from the State, local, and participating Tribal governmental agencies with emergency planning responsibilities. The certifications must state that— (A) The proposed emergency plans are practicable; (B) These agencies are committed to participating in any further development of the plans, including any required field demonstrations; and (C) These agencies are committed to executing their responsibilities under the plans in the event of an emergency. (ii) If certifications cannot be obtained after sustained, good faith efforts by the applicant, then the application must contain information, including a utility plan, sufficient to show that the proposed plans provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency at the site. (10) 
                                    *Organization.*
                                     A description of the applicant's organizational structure, allocations of responsibilities and authorities, and personnel qualifications requirements for operation. (11) 
                                    *Maintenance program.*
                                     A description of a maintenance program under § 53.715. (12) 
                                    *Quality assurance.*
                                     A description of the QAP under § 53.865. (13) 
                                    *Radiation protection program.*
                                     A radiation protection program description under § 53.850. (14) 
                                    *Security program.*
                                     A physical security plan that describes how the applicant will comply with § 53.860 (and 10 CFR part 11, if applicable, including the identification and description of jobs as required by § 11.11(a) of this chapter, at the proposed facility). The plan must list tests, inspections, audits, and other means to be used to demonstrate compliance with the requirements of 10 CFR parts 11 and 73, if applicable. (15) 
                                    *Safeguards contingency plan.*
                                     A safeguards contingency plan in accordance with the criteria set forth in appendix C to 10 CFR part 73. The safeguards contingency plan must include plans for dealing with threats, thefts, and radiological sabotage, as defined in 10 CFR part 73, relating to the SNM and nuclear facilities licensed under this chapter and in the applicant's possession and control. Each 
                                    ( print page 87098)
                                    application for this type of license must include the information contained in the applicant's safeguards contingency plan.1
                                     (Implementing procedures required for this plan need not be submitted for approval.) (16) 
                                    *Security training and qualification.*
                                     A training and qualification plan that describes how the applicant will demonstrate compliance with the criteria set forth in § 73.100 of this chapter or appendix B to 10 CFR part 73. (17) 
                                    *Cybersecurity plan.*
                                     A cybersecurity plan in accordance with the criteria set forth in § 73.54 or § 73.110 of this chapter. (18) 
                                    *Security, safeguards and cybersecurity plan implementation.*
                                     A description of the implementation of the physical security plan, safeguards contingency plan, training and qualification plan, and cybersecurity plan. Each applicant who prepares a physical security plan, a safeguards contingency plan, a training and qualification plan, or a cybersecurity plan must protect the plans and other related Safeguards Information against unauthorized disclosure in accordance with the requirements of §§ 73.21 and 73.22 of this chapter. (19) 
                                    *Fire protection program.*
                                     A description of the fire protection program under § 53.875. (20) 
                                    *Inservice inspection/inservice testing program.*
                                     Descriptions of inservice inspection and inservice testing programs under § 53.880. (21) [Reserved] (22) [Reserved] (23) 
                                    *General employee training.*
                                     A description of the training program required to demonstrate compliance with § 53.830 and its implementation. (24) 
                                    *Fitness-for-duty program.*
                                     A description of the FFD program under part 26 of this chapter and its implementation. (25) 
                                    *Technical specifications.*
                                     Proposed technical specifications prepared in accordance with the requirements of § 53.710(a). (b) If there are SSCs of the plant for which research and development is necessary to confirm the adequacy of their design, a report which documents the resolution of any safety questions associated with such SSCs. (c) A description of how the performance of each safety design feature has been demonstrated capable of fulfilling functional design criteria considering interdependent effects through either analysis, appropriate test programs, prototype testing, operating experience, or a combination thereof, in accordance with § 53.440(a). (d) If the COL application references an early site permit, then the following requirements apply: (1) The FSAR need not contain information or analyses submitted to the Commission in connection with the early site permit provided that the FSAR must either include or incorporate by reference the early site permit Site Safety Analysis Report and contain, in addition to the information and analyses otherwise required, information sufficient to demonstrate that the design of the facility falls within the site characteristics and design parameters specified in the early site permit. (2) If the FSAR does not demonstrate that design of the facility falls within the site characteristics and design parameters, the application must include a request for a variance that complies with the requirements of §§ 53.1188(d) and 53.1437. (3) The FSAR must demonstrate that all terms and conditions that have been included in the early site permit will be satisfied by the date of issuance of the COL. Any terms or conditions of the early site permit that could not be met by the time of issuance of the COL must be set forth as terms or conditions of the COL. (4) If the early site permit approves complete and integrated emergency plans, or major features of emergency plans, then the FSAR must include any new or additional information that updates and corrects the information that was provided under § 53.1146(b)(2) and discuss whether the new or additional information materially changes the bases for compliance with the applicable requirements. The application must identify changes to the emergency plans or major features of emergency plans that have been incorporated into the proposed facility emergency plans and that constitute or would constitute a change in an emergency plan that results in reducing the licensee's capability to perform an emergency planning function in the event of a radiological emergency. (5) If complete and integrated emergency plans are approved as part of the early site permit, new certifications meeting the requirements of paragraph (a)(9)(i) of this section are not required. (e) If the COL application references a standard design approval, then the following requirements apply: (1) The FSAR need not contain information or analyses submitted to the Commission in connection with the design approval, provided, however, that the FSAR must either include or incorporate by reference the standard design approval FSAR and must contain, in addition to the information and analyses otherwise required, information sufficient to demonstrate that the characteristics of the site fall within the site parameters specified in the design approval. In addition, the plant-specific PRA information must use the PRA information for the design approval and must be updated to account for site specific design information and any design changes or departures. (2) The FSAR must demonstrate that all terms and conditions that have been included in the design approval will be satisfied by the date of issuance of the COL. (f) If the COL application references a standard design certification, then the following requirements apply: (1) The FSAR need not contain information or analyses submitted to the Commission in connection with the standard design certification, provided, however, that the FSAR must either include or incorporate by reference the standard design certification FSAR and must contain, in addition to the information and analyses otherwise required, information sufficient to demonstrate that the site characteristics fall within the site parameters specified in the standard design certification. In addition, the plant-specific PRA information must use the PRA information for the standard design certification and must be updated to account for site-specific design information and any design changes or departures. (2) The FSAR must demonstrate that the interface requirements established for the design under § 53.1239(a)(24) have been met. (3) The FSAR must demonstrate that all requirements and restrictions set forth in the referenced standard design certification rule must be satisfied by the date of issuance of the COL. Any requirements and restrictions set forth in the referenced standard design certification rule that could not be satisfied by the time of issuance of the COL, must be set forth as terms or conditions of the COL. (g) If the COL application references the use of one or more manufactured reactors licensed under § 53.1270, then the following requirements apply: (1) The FSAR need not contain information or analyses submitted to the Commission in connection with the ML, provided, however, that the FSAR must either include or incorporate by reference the ML FSAR and must contain, in addition to the information and analyses otherwise required, information sufficient to demonstrate that the site characteristics fall within the site parameters specified in the ML. 
                                    ( print page 87099)
                                    In addition, the plant-specific PRA information must use the PRA information for the manufactured reactor and must be updated to account for site-specific design information and any design changes or departures. (2) The FSAR must demonstrate that the interface requirements established for the design have been met. (3) The FSAR must demonstrate that all terms and conditions that have been included in the ML will be satisfied by the date of issuance of the COL. Any terms or conditions of the ML that could not be met by the time of issuance of the COL, must be set forth as terms or conditions of the COL. (h) Each applicant for a COL under this part must protect Safeguards Information against unauthorized disclosure in accordance with the requirements in §§ 73.21 and 73.22 of this chapter, as applicable. 1
                                         A physical security plan that contains all the information required in both § 73.55 or § 73.100 of this chapter and appendix C to 10 CFR part 73 demonstrates compliance with the requirement for a contingency plan.
 § 53.1419
  (a) In addition to the FSAR, the application must also include the following: (1) 
                                    *Environmental report.*
                                     (i) An environmental report either in accordance with § 51.50(c) of this chapter if an LWA under § 53.1130 is not requested in conjunction with the COL application, or in accordance with §§ 51.49 and 51.50(c) of this chapter if an LWA is requested in conjunction with the COL application; or (ii) If the applicant wishes to request that an LWA under § 53.1130 be issued before issuance of the COL, the information otherwise required by § 53.1130, in accordance with either § 2.101(a)(1) through (a)(4), or § 2.101(a)(9) of this chapter; (2) 
                                    *Availability controls (if not included in the FSAR).*
                                     A description of the controls on plant operations, including availability controls, to provide reasonable confidence of safe operation and that the configurations and special treatments for SR SSCs and NSRSS SSCs provide the capabilities and reliabilities required to satisfy the safety criteria of § 53.220, or more restrictive alternative criteria adopted under § 53.470, if not addressed by Technical Specifications under § 53.1416(a)(25); and (3) 
                                    *Inspections, tests, analyses, and acceptance criteria.*
                                     The proposed inspections, tests, and analyses, including those applicable to emergency planning, that the licensee must perform, and the acceptance criteria that are necessary and sufficient to provide reasonable assurance that, if the inspections, tests, and analyses are performed and the acceptance criteria met, the facility has been constructed and will be operated in conformity with the COL, the provisions of the Act, and the Commission's rules and regulations. (i) If the application references an early site permit with ITAAC, the early site permit ITAAC must apply to those aspects of the COL which are approved in the early site permit. (ii) If the application references a standard design certification, the ITAAC contained in the certified design must apply to those portions of the facility design which are approved in the standard design certification. (iii) If the application references an ML, the ITAAC contained in the ML must apply to those portions of the facility design which are approved in the ML. (iv) If the application references an early site permit with ITAAC, a standard design certification, an ML, or combination thereof, the application may include a notification that a required inspection, test, or analysis in the ITAAC has been successfully completed and that the corresponding acceptance criterion has been met. The 
                                    **Federal Register**
                                     notification required by § 53.1422 of this chapter must indicate that the application includes this notification. (b) [Reserved]
 § 53.1422
  (a) 
                                    *Standards for review of applications.*
                                     Applications filed under this part will be reviewed according to the standards set out in this part and 10 CFR parts 20, 51, 73, and 140. (b) 
                                    *Administrative review of applications; hearings.*
                                     A proceeding on a COL is subject to all applicable procedural requirements contained in 10 CFR part 2, including the requirements for docketing (§ 2.101 of this chapter) and issuance of a notice of hearing (§ 2.104 of this chapter). If an applicant requests a Commission finding on certain ITAAC with the issuance of the COL, then those ITAAC will be identified in the notice of hearing. All hearings on COLs are governed by the procedures contained in 10 CFR part 2.
 § 53.1425
  If the application for a COL under this part references an early site permit, standard design certification rule, standard design approval, or ML, issued under this part, the scope and nature of matters resolved for the application and any COL issued are governed by the relevant provisions addressing finality, including §§ 53.1188, 53.1221, 53.1263, and 53.1288.
 § 53.1431
  The Commission must refer a copy of the application to the ACRS. The ACRS must report on those portions of the application that concern safety and must apply the standards referenced in § 53.1422, in accordance with the finality provisions in § 53.1425.
 § 53.1434
  (a) If the application for a COL under this part does not reference an early site permit which authorizes the holder to perform the activities under § 53.1130(b), the applicant may not perform those activities without obtaining the separate authorization required by § 53.1130(a). Authorization may be granted only after the presiding officer in the proceeding on the application has made the findings and determination required by § 53.1130(b)(1)(ii) and (b)(1)(iv), and the Director, Office of Nuclear Reactor Regulation makes the determination required by § 53.1130(b)(1)(iii). (b) If, after an applicant has performed the activities permitted by a LWA issued under § 53.1130, the application for the COL is withdrawn or denied, then the applicant must implement the approved site redress plan.
 § 53.1437
  (a) An applicant for a COL, or any amendment to a COL, may include in the application a request for an exemption from one or more of the Commission's regulations. (1) If the request is for an exemption from any part of a referenced standard design certification rule, the Commission may grant the request if it determines that the exemption complies with any exemption provisions of the referenced standard design certification rule, or with § 53.1263 if there are no applicable exemption provisions in the referenced standard design certification rule. (2) For all other requests for exemptions, the Commission may grant a request if it determines that the exemption complies with § 53.080. (b) An applicant for a COL who has filed an application referencing an early site permit issued under § 53.1158 may include in the application a request for a variance from one or more site characteristics, design parameters, or 
                                    ( print page 87100)
                                    terms and conditions of the permit, or from the Site Safety Analysis Report. In determining whether to grant the variance, the Commission must apply the same technically relevant criteria as were applicable to the application for the original or renewed site permit. Once a COL referencing an early site permit is issued, variances from the early site permit will not be granted for that CP or COL. (c) An applicant for a COL who has filed an application referencing use of a manufactured reactor may include in the application a request for a departure from one or more design characteristics, site parameters, terms and conditions, or approved design of the manufactured reactor under the ML issued under § 53.1287. The Commission may grant such a request only if it determines that the departure will comply with the requirements of § 53.080, and that the special circumstances outweigh any decrease in safety that may result from the reduction in standardization caused by the departure. (d) Issuance of a variance under paragraph (b) of this section or a departure under paragraph (c) of this section is subject to litigation during the COL proceeding in the same manner as other issues material to that proceeding.
 § 53.1440
  (a)(1) After conducting a hearing under § 53.1422(b) and receiving the report submitted by the ACRS, the Commission may issue a COL if the Commission finds that— (i) The applicable standards and requirements of the Act and the Commission's regulations have been met; (ii) Any required notifications to other agencies or bodies have been duly made; (iii) There is reasonable assurance that the facility will be constructed and will operate in conformity with the license, the provisions of the Act, and the Commission's regulations; (iv) The applicant is technically and financially qualified to engage in the activities authorized; however, no finding of financial qualification is necessary for an electric utility applicant for a COL; (v) Issuance of the license will not be inimical to the common defense and security or to the health and safety of the public; and (vi) The findings required by subpart A of 10 CFR part 51 have been made. (2) The Commission may also find, at the time it issues the COL, that certain acceptance criteria in one or more of the ITAAC in a referenced early site permit, standard design certification, or ML have been met. This finding will finally resolve that those acceptance criteria have been met, those acceptance criteria will be deemed to be excluded from the COL, and findings under § 53.1452(g) with respect to those acceptance criteria are unnecessary. (b) The Commission must identify within the COL the inspections, tests, and analyses, including those applicable to emergency planning, that the licensee must perform, and the acceptance criteria that, if met, are necessary and sufficient to provide reasonable assurance that the facility has been constructed and will be operated in conformity with the license, the provisions of the Act, and the Commission's rules and regulations. (c) A COL must contain the terms and conditions, including technical specifications, as the Commission deems necessary and appropriate.
 § 53.1443
  (a) After issuance of a COL, the Commission may not modify, add, or delete any term or condition of the COL, the design of the facility, the ITAAC contained in the license that are not derived from a referenced standard design certification or ML, except under the provisions of § 53.1452 or § 53.1590. (b) If the COL does not reference a standard design certification or use of a manufactured reactor under an ML issued under § 53.1287, then a licensee may make changes in the facility as described in the FSAR (as updated) and make changes in the procedures as described in the FSAR (as updated) under the applicable change processes in § 53.1550. (c) If the COL references a certified design, then— (1) Changes to or departures from information within the scope of the referenced standard design certification rule are subject to the applicable change processes in that rule; and (2) Changes that are not within the scope of the referenced standard design certification rule are subject to the applicable change processes in subpart I of this part, unless they also involve changes to or noncompliance with information within the scope of the referenced standard design certification rule. In these cases, the applicable provisions of this section and the standard design certification rule apply. (d) If the COL references use of a manufactured reactor under an ML issued under this part, then— (1) Changes to or departures from information within the scope of the manufactured reactor's design are subject to the change processes in § 53.1288; and (2) Changes that are not within the scope of the manufactured reactor's design are subject to the applicable change processes in subpart I. (e) The Commission may issue and make immediately effective any amendment to a COL upon a determination by the Commission that the amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person. The amendment may be issued and made immediately effective in advance of the holding and completion of any required hearing. The amendment will be processed under the procedures specified in § 53.1515. (f) Any modification to, addition to, or deletion from the terms and conditions of a COL, including any modification to, addition to, or deletion from the inspections, tests, and analyses, or related acceptance criteria contained in the license is a proposed amendment to the license. There must be an opportunity for a hearing on the amendment.
 § 53.1449
  (a) 
                                    *Licensee schedule for inspections, tests, or analyses.*
                                     The licensee must submit to the NRC, no later than 1 year after issuance of the COL or at the start of construction as defined at § 53.020, whichever is later, its schedule for completing the inspections, tests, or analyses in the ITAAC. The licensee must submit updates to the ITAAC schedules every 6 months thereafter and, within 1 year of its scheduled date for initial loading of fuel (or, for a fueled manufactured reactor, within 1 year of its scheduled date for initiating the physical removal of any one of the independent physical mechanisms to prevent criticality required under § 53.620(d)(1)), the licensee must submit updates to the ITAAC schedule every 30 days until the final notification is provided to the NRC under paragraph (c)(1) of this section. (b) 
                                    *Licensee and applicant conduct of activities subject to ITAAC.*
                                     With respect to activities subject to an ITAAC, an applicant for a COL may proceed at its own risk with design and procurement activities, and a licensee may proceed at its own risk with design, procurement, construction, and preoperational activities, even though the NRC may not have found that any one of the prescribed acceptance criteria are met. (c) 
                                    *Licensee notifications.*
                                     (1) 
                                    *ITAAC closure notification.*
                                     The licensee must notify the NRC that prescribed inspections, tests, and analyses have been performed and that the prescribed acceptance criteria are met. The 
                                    ( print page 87101)
                                    notification must contain sufficient information to demonstrate that the prescribed inspections, test, and analyses have been performed and that the prescribed acceptance criteria are met. (2) 
                                    *ITAAC post-closure notifications.*
                                     Following the licensee's ITAAC closure notifications under paragraph (c)(1) of this section until the Commission makes the finding under § 53.1452(g), the licensee must notify the NRC, in a timely manner, of new information that materially alters the basis for determining that either inspections, tests, or analyses were performed as required, or that acceptance criteria are met. The notification must contain sufficient information to demonstrate that, notwithstanding the new information, the prescribed inspections, tests, and analyses have been performed as required, and the prescribed acceptance criteria are met. (3) 
                                    *Uncompleted ITAAC notification.*
                                     If the licensee has not provided, by the date 225 days before the scheduled date for initial loading of fuel (or, for a fueled manufactured reactor, by the date 225 days before the scheduled date for initiating the physical removal of any one of the independent physical mechanisms to prevent criticality required under § 53.620(d)(1)), the notification required by paragraph (c)(1) of this section for all ITAAC, then the licensee must notify the NRC that the prescribed inspections, tests, or analyses for all uncompleted ITAAC will be performed and that the prescribed acceptance criteria will be met prior to operation. The notification must be provided no later than the date 225 days before the scheduled date for initial loading of fuel (or, for a fueled manufactured reactor, no later than the date 225 days before the scheduled date for initiating the physical removal of any one of the independent physical mechanisms to prevent criticality required under § 53.620(d)(1)), and must provide sufficient information to demonstrate that the prescribed inspections, tests, or analyses will be performed and the prescribed acceptance criteria for the uncompleted ITAAC will be met, including, but not limited to, a description of the specific procedures and analytical methods to be used for performing the prescribed inspections, tests, and analyses and determining that the prescribed acceptance criteria are met. (4) 
                                    *All ITAAC complete notification.*
                                     The licensee must notify the NRC that all ITAAC are complete. (d) 
                                    *Licensee determination of noncompliance with ITAAC.*
                                     (1) In the event that an activity is subject to an ITAAC derived from a referenced standard design certification and the licensee has not demonstrated that the prescribed acceptance criteria are met, the licensee may take corrective actions to successfully complete that ITAAC or request an exemption from the standard design certification ITAAC, as applicable. A request for an exemption must also be accompanied by a request for a license amendment under subpart I. (2) In the event that an activity is subject to an ITAAC not derived from a referenced standard design certification and the licensee has not demonstrated that the prescribed acceptance criteria are met, the licensee may take corrective actions to successfully complete that ITAAC or request a license amendment under subpart I. (e) 
                                    *NRC inspection, publication of notices, and availability of licensee notifications.*
                                     The NRC must ensure that the prescribed inspections, tests, and analyses in the ITAAC are performed. (1) At appropriate intervals until the last date for submission of requests for hearing under § 53.1452, the NRC must publish notices in the 
                                    **Federal Register**
                                     of the NRC staff's determination of the successful completion of inspections, tests, and analyses. (2) The NRC must make publicly available the licensee notifications under paragraph (c) of this section. The NRC must, no later than the date of publication of the notice of intended operation required by § 53.1452(a), make publicly available those licensee notifications under paragraph (c) of this section that have been submitted to the NRC at least 7 days before that notice.
 § 53.1452
  (a) The licensee must notify the NRC of its scheduled date for initial loading of fuel no later than 270 days before the scheduled date and must notify the NRC of updates to its schedule every 30 days thereafter.1
                                     Not less than 180 days before the date scheduled for initial loading of fuel into a plant by a licensee that has been issued a COL under this part, the Commission must publish notice of intended operation in the 
                                    **Federal Register**
                                    .2
                                     The notice must provide that any person whose interest may be affected by operation of the plant may, within 60 days, request that the Commission hold a hearing on whether the facility as constructed complies, or on completion will comply, with the acceptance criteria in the COL, except that a hearing must not be granted for those ITAAC that the Commission found were met under § 53.1440(a)(2). (b) A request for hearing under paragraph (a) of this section must show, 
                                    *prima facie*
                                    — (1) That one or more of the acceptance criteria of the ITAAC in the COL have not been, or will not be, met; and (2) The specific operational consequences of nonconformance that would be contrary to providing reasonable assurance of adequate protection of the public health and safety. (c) The Commission, acting as the presiding officer, must determine whether to grant or deny the request for hearing under the applicable requirements of § 2.309 of this chapter. If the Commission grants the request, the Commission, acting as the presiding officer, must determine whether during a period of interim operation there will be reasonable assurance of adequate protection to the public health and safety. The Commission's determination must consider the petitioner's 
                                    *prima facie*
                                     showing and any answers thereto. If the Commission determines there is such reasonable assurance, it must allow operation during an interim period under the COL. (d) The Commission, in its discretion, must determine appropriate hearing procedures, whether informal or formal adjudicatory, for any hearing under paragraph (a) of this section, and must state its reasons therefore. (e) The Commission must, to the maximum possible extent, render a decision on issues raised by the hearing request within 180 days of the publication of the notice provided by paragraph (a) of this section or by the anticipated date for initial loading of fuel into the reactor, whichever is later. (f) A petition to modify the terms and conditions of the COL will be processed as a request for action under § 2.206 of this chapter. The petitioner must file the petition with the Secretary of the Commission. Before the licensed activity allegedly affected by the petition (fuel loading, low power testing, etc.) commences, the Commission must determine whether any immediate action is required. If the petition is granted, then an appropriate order will be issued. Fuel loading and operation under the COL will not be affected by the granting of the petition unless the order is made immediately effective. (g) The licensee must not operate the facility until the Commission makes a finding that the acceptance criteria in the COL are met, except for those acceptance criteria that the Commission found were met under § 53.1440(a)(2). If the COL is for a modular design, each 
                                    ( print page 87102)
                                    reactor unit may require a separate finding as construction proceeds. (h) After the Commission has made the finding in paragraph (g) of this section, the ITAAC do not, by virtue of their inclusion in the COL, constitute regulatory requirements either for licensees or for renewal of the license; except for the specific ITAAC for which the Commission has granted a hearing under paragraph (a) of this section, all ITAAC expire upon final Commission action in the proceeding. However, subsequent changes to the facility or procedures described in the FSAR (as updated) must comply with the requirements in § 53.1443(e) or (f), as applicable. 1
                                         For licensees installing fueled manufactured reactors under a COL, the COL holder must instead notify the NRC of its scheduled date for initiating the physical removal of any one of the independent physical mechanisms to prevent criticality required under § 53.620(d)(1) no later than 270 days before the scheduled date and must notify the NRC of updates to its schedule every 30 days thereafter. 2
                                         For licensees installing fueled manufactured reactors under a COL, the Commission must instead publish notice of intended operation in the 
                                        **Federal Register**
                                         not less than 180 days before the date scheduled for initiating the physical removal of any one of the independent physical mechanisms to prevent criticality required under § 53.620(d)(1).
 § 53.1455
  A COL is issued for a specified period not to exceed 40 years from the date on which the Commission makes a finding that acceptance criteria are met under § 53.1452(g) or allowing operation during an interim period under the COL under § 53.1452(c).
 § 53.1456
  A COL may be transferred under § 53.1570.
 § 53.1458
  The filing of an application for a renewed license must be in accordance with § 53.1595.
 § 53.1461
  Each COL for a facility that has permanently ceased operations continues in effect beyond the expiration date to authorize ownership and possession of the facility until the Commission notifies the licensee in writing that the license is terminated. During this period of continued effectiveness, the licensee must— (a) Take actions necessary to decommission and decontaminate the facility and continue to maintain the facility, including, where applicable, the storage, control and maintenance of the spent fuel, in a safe condition; and (b) Conduct activities in accordance with all other restrictions applicable to the facility in accordance with the NRC's regulations and the provisions of the COL for the facility.
 § 53.1470
  (a) Except as otherwise specified in this section, the provisions of this section apply to CP, OL, and COL applications for commercial nuclear plants of identical design (the “common design”) under this part. (b) Each application for a CP, OL, or COL submitted pursuant to this section must be submitted as specified in §§ 53.1300, 53.1360, or 53.1410, respectively, and § 2.101 of this chapter. Each application must state that the applicant wishes to construct a facility identical to a facility proposed for one or more sites other than the applicant's (the “common design”), and the applicant wishes to have the application considered under this section. Each application must list each of the other applications to be treated together under this section. (c) Each application must include the information required by the applicable sections of this subpart, 
                                    *provided however,*
                                     that the application must identify the common design, and, if applicable, reference a standard design certification or standard design approval under this part, or the use of a reactor manufactured under this part. The FSAR for each application must either incorporate by reference or include the final safety analysis of the common design, including, if applicable, the FSAR for the referenced standard design certification, standard design approval, or the manufactured reactor. (d) Each application submitted pursuant to this section must contain an environmental report under §§ 53.1312(a)(1), 53.1372(a), or 53.1419(a)(1), as applicable, that complies with the applicable provisions of 10 CFR part 51, 
                                    *provided, however,*
                                     that the application may incorporate by reference a single environmental report on the environmental impacts of the common design that are applicable to each site. (e) Upon a determination that each application is acceptable for docketing under § 2.101 of this chapter, each application will be docketed and a notice of docketing for each application will be published in the 
                                    **Federal Register**
                                    , under § 2.104 of this chapter, 
                                    *provided, however,*
                                     that the notice must state that the application will be processed under the provisions of this section and subpart D of 10 CFR part 2. At the discretion of the Commission, a single notice of docketing for multiple applications may be published in the 
                                    **Federal Register**
                                    . (f) The NRC must prepare an environmental assessment or draft and final environmental impact statements for each of the applications under 10 CFR part 51. Scoping under §§ 51.28 and 51.29 of this chapter for each of the license applications may be conducted simultaneously and joint scoping may be conducted with respect to the environmental issues relevant to the common design. If the applications reference a standard design certification, then the environmental assessment or environmental impact statement for each of the applications must incorporate by reference the standard design certification environmental assessment. If the applications do not reference a standard design certification, then the NRC must prepare environmental assessments or draft and final supplemental environmental impact statements which address severe accident mitigation design alternatives for the common design, which must be incorporated by reference into the environmental assessment or environmental impact statement prepared for each application. Scoping under §§ 51.28 and 51.29 of this chapter for the supplemental environmental impact statement may be conducted simultaneously and may be part of the scoping for each of the applications. (g) The ACRS must report on each of the applications as required by the applicable sections of this subpart. Each report must be limited to those safety matters for each application that are not relevant to the common design. In addition, the ACRS must separately report on the safety of the common design, 
                                    *provided, however,*
                                     that the report need not address the safety of a referenced standard design certification or reactor manufactured under this part. (h) The Commission must designate a presiding officer to conduct the proceeding with respect to the health and safety, common defense and security, and environmental matters relating to the common design and affecting at least two applications. The hearing will be governed by the applicable provisions of subparts A, C, G, L, N, and O of 10 CFR part 2 relating to applications for CPs, OLs, and COLs. The presiding officer must issue a partial initial decision on the common design. (i) If the design for the power reactor(s) proposed in a particular 
                                    ( print page 87103)
                                    application is not identical to the others, that application may not be processed under this section and subpart D of 10 CFR part 2. (j) As used in this section, the design of a nuclear power reactor included in a single referenced Safety Analysis Report means the design of those SSCs important to radiological health and safety and the common defense and security.
 Subpart I—Maintaining and Revising Licensing-Basis Information § 53.1500
  This subpart provides the requirements for each holder of a license for a commercial nuclear plant licensed under this part to maintain licensing-basis information as defined in § 53.020; evaluate changes to site characteristics, plant design features, and programmatic controls to determine needed approvals and revisions; and submit appropriate updates to the U.S. Nuclear Regulatory Commission (NRC).
 § 53.1502
  (a) Each license issued under this part is subject to the provisions of the Atomic Energy Act of 1954, as amended, (the Act) and to all rules, regulations, and orders of the Commission. The terms and conditions of the license will be subject to amendment, revision, or modification, by reason of amendments of the Act or by reason of rules, regulations, and orders issued in accordance with the terms of the Act. (b) Each license issued under this part must be subject to all conditions imposed as a matter of law by sections 401(a)(2) and 401(d) of the Federal Water Pollution Control Act, as amended (33 U.S.C.A. 1341(a)(2) and (d)). (c) A holder of an operating license (OL) or combined license (COL) under this part may take reasonable action that departs from a license condition or a technical specification included in a license issued under this part in a national security emergency established by a law enacted by the Congress or by an order or directive issued by the President pursuant to statutes or the Constitution of the United States. The authority under this paragraph must be exercised in accordance with law, including section 57e of the Act, and is in addition to the authority granted under § 53.740(h), which remains in effect unless otherwise directed by the Commission during a national security emergency. The authority under this paragraph may be exercised— (1) When this action is immediately needed to implement national security objectives as designated by the national command authority through the Commission; and (2) No action consistent with license conditions and technical specifications that can satisfy national security objectives is immediately apparent. (d)(1) If the NRC finds that the state of emergency preparedness does not provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency (including findings based on requirements of 10 CFR part 50, appendix E, section IV.D.3) and if the deficiencies (including deficiencies based on requirements of 10 CFR part 50, appendix E, section IV.D.3) are not corrected within 4 months of that finding, the Commission will determine whether the facility must be shut down or cease operations until such deficiencies are remedied or whether other enforcement action is appropriate. In determining whether a shutdown or other enforcement action is appropriate, the Commission will take into account, among other factors, whether the licensee can demonstrate to the Commission's satisfaction that the deficiencies in the plan are not significant for the plant in question, or that adequate interim compensating actions have been or will be taken promptly, or that that there are other compelling reasons for continued operation. (2) If the planning standards for radiological emergency preparedness apply to offsite emergency response plans, or if the planning activities in § 50.160(b)(1)(iv)(B) apply, then the NRC will base its finding on a review of the Federal Emergency Management Agency findings and determinations as to whether State, participating Tribal and local emergency plans are adequate and capable of being implemented, and on the NRC assessment as to whether the licensee's emergency plans are adequate and capable of being implemented. Nothing in this paragraph must be construed as limiting the authority of the Commission to take action under any other regulation or authority of the Commission or at any time other than that specified in this paragraph.
 § 53.1505
  (a) Sections 53.1510 through 53.1520 provide the process for a licensee to request and the NRC to issue amendments to licenses, including any conditions contained therein, technical specifications or other attachments to a license, and any orders issued by the NRC modifying a license. Sections 53.1525 and 53.1530 govern proposed changes to a commercial nuclear plant referencing a certified design or manufacturing license (ML). (b) A licensee may propose changing licensing-basis information established by NRC regulations by requesting an exemption in accordance with § 53.080.
 § 53.1510
  Whenever a holder of a license under this part desires to amend the license, an application for an amendment must be filed with the Commission, as specified in § 53.040, that fully describes the changes desired and, following as far as applicable, the form prescribed for original applications. Applications for amendments involving changes to plant structures, systems, and components (SSCs), programmatic controls, or the role of plant personnel must include an assessment of the changes in relation to the safety requirements in subpart B of this part and the analyses requirements of § 53.450 as applicable, an analysis of whether the amendment involves no significant hazards consideration using the standards in § 53.1520, and a consideration of environmental factors.
 § 53.1515
  The Commission will use the following procedures for an application requesting an amendment to an OL or COL issued under this part. (a) 
                                    *Public notices.*
                                     (1)(i) The Commission may publish in the 
                                    **Federal Register**
                                     under § 2.105 of this chapter an individual notice of proposed action for an amendment for which it makes a proposed determination that no significant hazards consideration is involved, or, at least once every 30 days, publish a periodic 
                                    **Federal Register**
                                     notice of proposed actions, which identifies each amendment issued and each amendment proposed to be issued since the last such periodic notice, or it may publish both such notices. (ii) For each amendment proposed to be issued, the notice will (A) Contain the staff's proposed determination under the standards in § 53.1520; (B) Provide a brief description of the amendment and of the facility involved; (C) Solicit public comments on the proposed determination; and (D) Provide for a 30-day comment period. (iii) The comment period will begin on the day after the date of the publication of the first notice, and, normally, the amendment will not be granted until after this comment period expires.
                                    ( print page 87104) (2) The Commission may inform the public about the final disposition of an amendment request for which it has made a proposed determination of no significant hazards consideration either by issuing an individual notice of issuance under § 2.106 of this chapter or by publishing such a notice in its periodic system of 
                                    **Federal Register**
                                     notices. In either event, it will not make and will not publish a final determination of no significant hazards consideration unless it receives a request for a hearing on that amendment request. (3) Where the Commission makes a final determination that no significant hazards consideration is involved and that the amendment should be issued, the amendment will be effective on issuance, even if adverse public comments have been received and even if an interested person meeting the provisions for intervention called for in § 2.309 of this chapter has filed a request for a hearing. The Commission need hold any required hearing only after it issues an amendment, unless it determines that a significant hazards consideration is involved, in which case the Commission will provide an opportunity for a prior hearing. (4) Where the Commission finds that an emergency situation exists, in that failure to act in a timely way would result in derating or shutdown of a commercial nuclear reactor, or in prevention of either resumption of operation or of increase in power output up to the plant's licensed power level, it may issue a license amendment involving no significant hazards consideration without prior notice and opportunity for a hearing or for public comment. In such a situation, the Commission will not publish a notice of proposed determination on no significant hazards consideration but will publish a notice of issuance under § 2.106 of this chapter providing for opportunity for a hearing and for public comment after issuance. The Commission expects its licensees to apply for license amendments in a timely fashion. It will decline to dispense with notice and comment on the determination of no significant hazards consideration if it determines that the licensee has abused the emergency provision by failing to make timely application for the amendment and thus itself creating the emergency. Whenever an emergency situation exists, a licensee requesting an amendment must explain why this emergency situation occurred and why it could not avoid this situation, and the Commission will assess the licensee's reasons for failing to file an application sufficiently in advance of that event. (5) Where the Commission finds that exigent circumstances exist, in that a licensee and the Commission must act quickly and that time does not permit the Commission to publish a 
                                    **Federal Register**
                                     notice allowing 30 days for prior public comment, and it also determines that the amendment involves no significant hazards considerations, it— (i)(A) Will either issue a 
                                    **Federal Register**
                                     notice providing notice of an opportunity for hearing and allowing at least 2 weeks from the date of the notice for prior public comment; or (B) Will use local media to provide reasonable notice to the public in the area surrounding a licensee's facility of the licensee's amendment and of its proposed determination as described in paragraph (a)(1) of this section, consulting with the licensee on the proposed media release and on the geographical area of its coverage; (ii) Will provide for a reasonable opportunity for the public to comment, using its best efforts to make available to the public whatever means of communication it can for the public to respond quickly, and, in the case of telephone comments, have these comments recorded or transcribed, as necessary and appropriate; (iii) When it has issued a local media release, may inform the licensee of the public's comments, as necessary and appropriate; (iv) Will publish a notice of issuance under § 2.106 of this chapter; (v) Will provide a hearing after issuance, if one has been requested by a person who satisfies the provisions for intervention specified in § 2.309 of this chapter; and (vi) Will require the licensee to explain the exigency and why the licensee cannot avoid it and use its normal public notice and comment procedures in paragraph (a)(1) of this section if it determines that the licensee has failed to use its best efforts to make a timely application for the amendment in order to create the exigency and to take advantage of this procedure. (6) Where the Commission finds that significant hazards considerations are involved, it will issue a 
                                    **Federal Register**
                                     notice providing an opportunity for a prior hearing even in an emergency situation, unless it finds an imminent danger to the health or safety of the public, in which case it will issue an appropriate order or rule under 10 CFR part 2. (b) 
                                    *State consultation.*
                                     (1) At the time a licensee requests an amendment, it must notify the State in which its facility is located of its request by providing that State with a copy of its application and its reasoned analysis about no significant hazards considerations and indicate on the application that it has done so. (2) The Commission will advise the State of its proposed determination about no significant hazards consideration normally by sending it a copy of the 
                                    **Federal Register**
                                     notice. (3) The Commission will make the names of the Project Manager or other NRC personnel it designated to consult with the State available to the State official designated to consult about its proposed determination. The Commission will consider any comments of that State official. If it does not hear from the State in a timely manner, it will consider that the State has no interest in its determination; nonetheless, to ensure that the State is aware of the application, before it issues the amendment, it will make a good faith effort to communicate directly with that official. (Inability to consult with a responsible State official following good faith attempts will not prevent the Commission from making effective a license amendment involving no significant hazards consideration.) (4) The Commission will make a good faith attempt to consult with the State before it issues a license amendment involving no significant hazards consideration. If, however, it does not have time to use its normal consultation procedures because of an emergency situation, it will attempt to communicate directly with the appropriate State official. (Inability to consult with a responsible State official following good faith attempts will not prevent the Commission from making effective a license amendment involving no significant hazards consideration, if the Commission deems it necessary in an emergency situation.) (5) After the Commission issues the requested amendment, it will send a copy of its determination to the State. (c) 
                                    *Caveats about State consultation.*
                                     (1) The State consultation procedures in paragraph (b) of this section do not give the State a right— (i) To veto the Commission's proposed or final determination; (ii) To a hearing on the determination before the amendment becomes effective; or (iii) To insist upon a postponement of the determination or upon issuance of the amendment. (2) These procedures do not alter present provisions of law that reserve to the Commission exclusive responsibility for setting and enforcing radiological 
                                    ( print page 87105)
                                    health and safety requirements for commercial nuclear plants.
 § 53.1520
  (a) In determining whether an amendment to a license will be issued to the applicant, the Commission will be guided by the considerations which govern the issuance of initial licenses to the extent applicable and appropriate. If the application is for amendment of an OL or COL and involves the material alteration of a commercial nuclear plant, a construction permit (CP) will be issued before the issuance of the amendment to the license, provided however, that if the application involves a material alteration to a manufactured reactor under this part before its installation at a site, or a COL before the date that the Commission makes the finding under § 53.1452(g), no application for or issuance of a CP is required. If the amendment involves a significant hazards consideration, the Commission will give notice of its proposed action— (1) Under § 2.105 of this chapter before acting thereon; and (2) As soon as practicable after the application has been docketed. (b) The Commission will be particularly sensitive to a license amendment request that involves irreversible consequences (such as one that permits a significant increase in the amount of effluents or radiation emitted by a commercial nuclear plant). (c) The Commission may make a final determination, under the procedures in § 53.1515, that a proposed amendment to an OL or a COL for a commercial nuclear plant under this part involves no significant hazards consideration, if operation of the plant in accordance with the proposed amendment would not— (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of an accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
 § 53.1525
  (a) A holder of an OL or COL who references a design certification rule issued under this part must request an exemption if proposing to change one or more elements of the certification information. The Commission may grant such a request only if it determines that the exemption will comply with the requirements of § 53.080 and that the special circumstances outweigh any decrease in safety that may result from the reduction in standardization caused by the departure. (b) The request for an exemption must be included with any associated license amendment request, which must be requested and processed in accordance with §§ 53.1510, 53.1515, and 53.1520. (c) Licensees must evaluate changes to the design as described in the Final Safety Analysis Report (FSAR) not involving changes to the certification information using the criteria in § 53.1550.
 § 53.1530
  (a) The holder of an ML may not make changes to the design of the manufactured reactor authorized to be manufactured without obtaining an amendment pursuant to § 53.1510 and, as applicable, § 53.1520. (b) The holder of a COL under this part who references or uses a manufactured reactor under this part must request approval for any proposed departure from the design characteristics, site parameters, terms and conditions, or approved design of the manufactured reactor. The application for such departures must be submitted and processed in accordance with §§ 53.1510, 53.1515, and 53.1520. In those cases where an ML references a design certification rule, the amendment application from the holder of the COL must also request an exemption from the design certification rule under § 53.1525 if one or more elements of the certification information are adversely affected by the proposed change. The holder of the COL must evaluate changes to the commercial nuclear plant as described in the FSAR but outside of the scope of the referenced ML using the criteria in § 53.1550.
 § 53.1535
  (a) The holder of a CP or limited work authorization (LWA) under this part may request an amendment to the CP or LWA in order to gain Commission approval of the safety of selected design features or specifications, including proposed departures from a design certification rule or ML. Amendments to CPs or LWAs under this part must be requested and processed under §§ 53.1510 and 53.1520. (b) The holder of a COL under this part for which the NRC has not yet made a finding in accordance with § 53.1452(g) must request amendments required by § 53.1525 or § 53.1550 no later than 45 days from the date the licensee begins the construction of the SSCs to implement the change or departure requiring NRC approval. The licensee proceeds with such changes at its own risk recognizing that there is a possibility that the amendment will not be granted.
 § 53.1540
  (a) Sections 53.1545 through 53.1565 provide the process for a holder of an OL or COL to modify licensing-basis information and to evaluate potential changes to its facilities, procedures, programs, and organizations to determine if NRC approval is required. (b) Definitions for the purposes of §§ 53.1545 through 53.1565— *Change*
                                     means a modification or addition to, or removal from, the commercial nuclear plant or procedures that affects a design feature or related functional design criteria, method of performing or controlling the functions of design features, or an evaluation that demonstrates that intended functions will be accomplished. *Departure from a method of evaluation described in the Final Safety Analysis Report (FSAR) (as updated) used in establishing the functional design criteria for safety-related structures, systems, or components or in the safety analyses*
                                     means— (1) Changing any of the elements of the method described in the FSAR (as updated) unless the results of the analysis are conservative or essentially the same; or (2) Changing from a method described in the FSAR to another method unless that method has been approved by NRC for the intended application. *Facility as described in the FSAR (as updated)*
                                     means— (1) The SSCs that are described in the FSAR (as updated), (2) The design and performance requirements for such SSCs described in the FSAR (as updated), and (3) The evaluations or methods of evaluation included in the FSAR (as updated) for such SSCs which demonstrate that their intended function(s) will be accomplished. *Final Safety Analysis Report (as updated)*
                                     means the FSAR submitted under § 53.1369 or § 53.1416, as amended and supplemented, and as updated under § 53.1545, as applicable. *Procedures as described in the Final Safety Analysis Report (as updated)*
                                     means those procedures that contain information described in the FSAR (as updated) such as how SSCs are operated and controlled (including assumed operator actions and response times).
 § 53.1545
  (a) Each holder of an OL or COL under this part for which the Commission has made the finding under § 53.1452(g) must update the FSAR originally submitted as part of the application for the license every 24 months or more frequently to assure that the information included in the report contains the latest information developed. The submittal must include the effects on the content of the FSAR of— (1) Changes made to the facility or procedures as described in the FSAR; (2) Safety analyses and evaluations performed by the licensee either in support of approved license amendments or in support of conclusions that changes did not require a license amendment under § 53.1550; (3) Updates to the probabilistic risk assessments required under § 53.450; (4) The cumulative effects of the changes to the facility or procedures on the margins to the safety criteria in §§ 53.210, 53.220, 53.450(e), and 53.470 since the last FSAR update; and (5) Analyses of new safety issues performed by or on behalf of the licensee at Commission request. (b)(1) The licensee must submit revisions containing updated information to the Commission, under § 53.040, identifying the location of revised or new information. (2) The submittal must include— (i) A certification by a duly authorized officer of the licensee that either the information accurately presents changes made since the previous submittal, necessary to reflect information and analyses submitted to the Commission or prepared pursuant to Commission requirement, or that no such changes were made; and (ii) An identification of changes made under the provisions of § 53.1550 but not previously submitted to the Commission. (c) Each applicant for or holder of a COL under this part for which the Commission has not made the finding under § 53.1452(g) must submit an update to the FSAR annually by providing the information required in (a)(1) through (a)(5) of this section and meeting the requirements of paragraph (b) of this section. Combined license applicants who have requested the NRC to suspend its review of the COL application and COL holders who have informed the NRC that they do not plan to pursue construction need not submit an annual update of the FSAR. If a COL applicant requests that the NRC resume its review, or a COL holder notifies the NRC that the COL holder plans to commence or resume construction, then the COL applicant or holder must submit to NRC an update to its FSAR within 90 days of the request or notification, as applicable, and annually thereafter. (d) The FSAR (as updated) must be retained by the licensee until the Commission terminates its license. (e) Each holder of an ML under this part must submit an update of the FSAR reflecting any modification to the design that is directed or approved by the Commission under § 53.1288 or § 53.1530, and any new analyses of the design requested by the Commission under § 53.1580.
 § 53.1550
  (a) The holder of an OL or COL may make changes in the facility as described in the FSAR (as updated) and make changes in the procedures as described in the FSAR (as updated) without obtaining a license amendment pursuant to § 53.1510 only if— (1) A change to the technical specifications incorporated in the license is not required; and (2) The change meets all of the following criteria: (i) Does not result in an increase to the frequency or consequences of an event sequence such that an event sequence not previously identified as risk significant becomes risk significant by the analyses performed in accordance with § 53.450(e). (ii) Does not result in an increase to the frequency or consequences of an event sequence such that an event sequence identified as risk significant in accordance with § 53.450(e) exceeds the licensing-basis event evaluation criteria required to be established in accordance with § 53.450(e). (iii) Does not involve either of the following: (A) a change to the 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 there is more than a minimal reduction in the margin between the calculated comprehensive risks posed by the commercial nuclear plant and the safety criteria of § 53.220. (iv) Does not involve a departure from a method of evaluation described in the FSAR (as updated) used in assessing licensing-basis events in accordance with § 53.450 unless the results of the analysis under § 53.450 are conservative or essentially the same, the revised method of evaluation has been previously approved by the NRC for the intended application, or the revised method of evaluation can be used under an NRC-endorsed consensus code or standard. (v) Does not result in the escalation in the safety classification of an SSC from non-safety-related to non-safety-related but safety-significant or from non-safety-related but safety-significant to safety-related. (vi) Does not result in more than a minimal decrease in defense in depth. (vii) For commercial nuclear plants licensed under this part for which alternative evaluation criteria are adopted in accordance with § 53.470, does not result in a change to the frequency or consequences of event sequences such that the calculated margins between the results for event sequences evaluated in accordance with § 53.450(e) and the alternative evaluation criteria decreases by 25 percent or more. (viii) Does not result in the identification of a new design-basis accident in accordance with § 53.450(f). (ix) Does not result in a decrease by 10 percent or more in the margin between the consequence of any design-basis accident and the safety criteria in § 53.210. (x) Does not prevent meeting the design requirements in § 53.440(j) to limit the release of radionuclides from reactor systems, waste stores, or other significant inventories of radioactive materials assuming the impact of a large, commercial aircraft. (3) In implementing this paragraph, the FSAR (as updated) is considered to include FSAR changes since submittal of the last update of the FSAR under § 53.1545. (4) The provisions in this section do not apply to changes to the facility or procedures when the applicable regulations establish more specific criteria for accomplishing such changes. (b)(1) A licensee who references a design certification rule may make departures from the standard design, without prior Commission approval, unless the proposed departure involves a change to the design as described in the rule certifying the design, in which case the requirements of § 53.1525 are applicable. (2) The licensee must maintain records of all departures from the certified design of the facility and these records must be maintained and available for audit until the termination of the license. The licensee must identify the location and nature of departures from licensing-basis information within supporting documents for a certified design within the updates to the Safety Analysis Report required by § 53.1545.
                                    ( print page 87107) (3) Licensees for which the NRC has docketed the certifications required under § 53.1070 need not retain records of departures from the design of the facility associated with SSCs that have been permanently removed from service using an NRC-approved change process. (c)(1) The licensee must maintain records of changes in the facility and procedures made under paragraph (a) of this section. These records must include a written evaluation which provides the bases for the determination that the change does not require a license amendment under paragraph (a)(2) of this section. (2) The licensee must submit, as specified in § 53.040, a report containing a brief description of any departures and changes, including a summary of the evaluation of each. A report must be submitted at intervals not to exceed 24 months. For COLs, the report must be submitted at intervals not to exceed 6 months during the period from the date of application for a COL to the date the Commission makes its findings under § 53.1452(g). (3) The records of changes in the facility must be maintained until the termination of an OL or COL issued under this part, or the termination of a renewed license issued under § 53.1595—whichever is later. Records of changes in procedures must be maintained for a period of 5 years.
 § 53.1560
  (a) Each holder under this part of an OL or COL for which the Commission has made the finding under § 53.1452(g) must biennially or more frequently update the program documents submitted as part of an application to obtain or maintain the license to assure that the information included in the documents contains the latest information developed. The submittals must include the effects on the content of the program documents of— (1) Changes made in the facility, procedures, licensee's organization, or site environs; (2) Safety analyses and evaluations performed by the applicant or licensee either in support of approved license amendments or in support of conclusions that changes did not require a license amendment in accordance with § 53.1550; (3) Analyses of new safety issues performed by or on behalf of the licensee at Commission request; and (4) Changes to the programs as a result of operating experience, corrective actions, or other reasons deemed appropriate to ensure the programs serve their underlying purpose to support the requirements in subpart B of this part or other NRC regulations. (b)(1) The licensee must submit revisions containing updated information to the Commission, as specified in § 53.040, identifying the location of revised or new information. (2) The submittal must include— (i) A certification by a duly authorized officer of the licensee that either the information accurately presents changes made since the previous submittals, necessary to reflect information and analyses submitted to the Commission or prepared pursuant to Commission requirement, or that no such changes were made; and (ii) An identification of changes made under the provisions of § 53.1550 but not previously submitted to the Commission. (c) The updated program documents must be retained by the licensee until the Commission terminates their license.
 § 53.1565
  (a) A licensee may make changes to the facility, procedures, or organizations or address changes to site environs as described in the program documents included in licensing-basis information without obtaining prior NRC approval only if— (1) A change to the technical specifications incorporated in the license is not required; (2) An exemption from an NRC regulation is not required; and (3) The change conforms to program-specific requirements included in regulations in this part, technical specifications, or the NRC-approved program document included and reviewed as part of a license application under subpart H or an amendment under this subpart. (b) In implementing this section, the program documents (as updated) include changes since submittal of the last updates of the program documents pursuant to § 53.1560. (c) The provisions in this section do not apply to changes to the program documents when the applicable regulations establish more specific criteria for accomplishing such changes. (d) To make changes to the facility, procedures, or organizations or to address changes to site environs as described in the program documents included in licensing-basis information for individual programs, the following requirements must be satisfied: (1) 
                                    *Quality assurance program—operation.*
                                     (i) Each holder under this part of an OL or COL, after the Commission makes the finding under § 53.1452(g), may make a change to a previously accepted quality assurance program (QAP) description included or referenced in the Safety Analysis Report without prior NRC approval, provided the change does not reduce the commitments in the program description as accepted by the NRC. Changes to the QAP description that do not reduce the commitments must be submitted to the NRC in accordance with the requirements of § 53.1545. In addition to QAP changes involving administrative improvements and clarifications, spelling corrections, punctuation, or editorial items, the following changes are not considered to be reductions in commitment: (A) The use of a quality assurance (QA) standard approved by the NRC which is more recent than the QA standard in the licensee's QAP at the time of the change; (B) The use of a QA alternative or exception approved by an NRC safety evaluation, provided that the bases of the NRC approval are applicable to the licensee's facility; (C) The use of generic organizational position titles that clearly denote the position function, supplemented as necessary by descriptive text, rather than specific titles; (D) The use of generic organizational charts to indicate functional relationships, authorities, and responsibilities, or, alternately, the use of descriptive text; (E) The elimination of QAP information that duplicates language in QA regulatory guides and QA standards to which the licensee is committed; and (F) Organizational revisions that ensure that persons and organizations performing QA functions continue to have the requisite authority and organizational freedom, including sufficient independence from cost and schedule when opposed to safety considerations. (ii) Changes to the QAP description that do reduce the commitments must be submitted to the NRC and receive NRC approval prior to implementation, as follows: (A) Changes made to the QAP description as presented in the Safety Analysis Report or in a topical report must be submitted as specified in § 53.040. (B) The submittal of a change to the Safety Analysis Report QAP description must include all pages affected by that change and must be accompanied by a forwarding letter identifying the change, the reason for the change, and the basis for concluding that the revised program incorporating the change continues to satisfy the criteria of appendix B to part 
                                    ( print page 87108)
                                    50 of this chapter and the Safety Analysis Report QAP description commitments previously accepted by the NRC (the letter need not provide the basis for changes that correct spelling, punctuation, or editorial items). (C) A copy of the forwarding letter identifying the change must be maintained as a facility record for 3 years. (D) Changes to the QAP description included or referenced in the Safety Analysis Report shall be regarded as accepted by the Commission upon receipt of a letter to this effect from the appropriate reviewing office of the Commission or 60 days after submittal to the Commission, whichever occurs first. (2) 
                                    *Quality assurance program—siting, construction, and manufacturing.*
                                     Each holder of an LWA, early site permit, CP, ML, or COL, before the Commission makes the finding under § 53.1452(g) of this chapter, under this part may make a change to a previously accepted QAP description included or referenced in the Safety Analysis Report without prior NRC approval, provided the change does not reduce the commitments in the program description previously accepted by the NRC. Changes to the QAP description that do not reduce the commitments must be submitted to NRC within 90 days. Changes to the QAP description that reduce the commitments must be submitted to NRC and receive NRC approval before implementation, as follows: (i) Changes to the Safety Analysis Report must be submitted for review as specified in § 53.040. Changes made to NRC-accepted QA topical report descriptions must be submitted as specified in § 53.040. (ii) The submittal of a change to the Safety Analysis Report QAP description must include all pages affected by that change and must be accompanied by a forwarding letter identifying the change, the reason for the change, and the basis for concluding that the revised program incorporating the change continues to satisfy the criteria of appendix B of part 50 of this chapter and the Safety Analysis Report QAP description commitments previously accepted by the NRC (the letter need not provide the basis for changes that correct spelling, punctuation, or editorial items). (iii) A copy of the forwarding letter identifying the changes must be maintained as a facility record for 3 years. (iv) Changes to the QAP description included or referenced in the Safety Analysis Report shall be regarded as accepted by the Commission upon receipt of a letter to this effect from the appropriate reviewing office of the Commission or 60 days after submittal to the Commission, whichever occurs first. (3) 
                                    *Emergency preparedness program.*
                                     (i) Definitions for the purpose of paragraph (d)(3) of this section: (A) 
                                    *Change*
                                     means an action that results in modification or addition to, or removal from, the licensee's emergency plan. All such changes are subject to the provisions of this section except where the applicable regulations establish specific criteria for accomplishing a particular change. (B) 
                                    *Emergency plan*
                                     means the document(s), prepared and maintained by the licensee, that identify and describe the licensee's methods for maintaining emergency preparedness and responding to emergencies. An emergency plan includes the plan as originally approved by the NRC and all subsequent changes made by the licensee with, and without, prior NRC review and approval under paragraph (d)(3) of this section. (C) 
                                    *Emergency planning function*
                                     means a capability or resource necessary to prepare for and respond to a radiological emergency. (D) 
                                    *Reduction in effectiveness*
                                     means a change in an emergency plan that results in reducing the licensee's capability to perform an emergency planning function in the event of a radiological emergency. (ii)(A) Except as provided in paragraph (d)(3)(ii)(B) of this section, a holder of an OL under this part, or a COL under this part after the Commission makes the finding under § 53.1452(g), must follow and maintain the effectiveness of an emergency plan that meets the requirements in appendix E to part 50 of this chapter and the planning standards of § 50.47(b). (B) A holder of an OL under this part for a commercial nuclear plant consisting of small modular reactors (SMRs) or non-light-water reactors, or a holder of a COL under this part after the Commission makes the finding under § 53.1452(g) for a commercial nuclear plant consisting of either SMRs or non-light-water reactors, must follow and maintain the effectiveness of either an emergency plan that meets the requirements in § 50.160 or an emergency plan that meets the requirements in appendix E to part 50 of this chapter and the planning standards of § 50.47(b). (iii)(A) Except as provided in paragraph (d)(3)(iii)(B) of this section, the licensee may make changes to its emergency plan without NRC approval only if the licensee performs and retains an analysis demonstrating that the changes do not reduce the effectiveness of the plan and the plan, as changed, continues to meet the requirements in appendix E to part 50 of this chapter and the planning standards of § 50.47(b). (B) A license under this part for a commercial nuclear plant consisting of either SMRs or non-light-water reactors may make changes to its emergency plan without NRC approval only if the licensee performs and retains an analysis demonstrating that the changes do not reduce the effectiveness of the plan and the plan, as changed, continues to meet either the requirements in § 50.160 or the requirements in appendix E to part 50 and the planning standards of § 50.47(b). (iv) The changes to a licensee's emergency plan that reduce the effectiveness of the plan as defined in paragraph (d)(3)(i)(D) of this section may not be implemented without prior approval by the NRC. A licensee desiring to make such a change must submit an application for an amendment to its license. In addition to the filing requirements of §§ 53.1510 and 53.1515, the request must include all emergency plan pages affected by that change and must be accompanied by a forwarding letter identifying the change, the reason for the change, and the basis for concluding that the licensee's emergency plan, as revised, will continue to meet either the requirements in § 50.160 to this chapter or the requirements in appendix E to part 50 of this chapter and the planning standards of § 50.47(b) of this chapter. (v) The licensee must retain a record of each change to the emergency plan made without prior NRC approval for a period of three years from the date of the change and shall submit, as specified in § 53.040, a report of each such change, including a summary of its analysis, within 30 days after the change is put in effect. (vi) The licensee must retain the emergency plan and each change for which prior NRC approval was obtained pursuant to paragraph (d)(3)(iv) of this section as a record until the Commission terminates the license for the nuclear power reactor. (vii)(A) The licensee must provide for the development, revision, implementation, and maintenance of its emergency preparedness program. The licensee must ensure that all program elements are reviewed by persons who have no direct responsibility for the implementation of the emergency preparedness program either—
                                    ( print page 87109) (
                                    *1*) At intervals not to exceed 12 months; or (
                                    *2*) As necessary, based on an assessment by the licensee against performance indicators, and as soon as reasonably practicable after a change occurs in personnel, procedures, equipment, or facilities that potentially could adversely affect emergency preparedness, but no longer than 12 months after the change. In any case, all elements of the emergency preparedness program must be reviewed at least once every 24 months. (B) The review must include an evaluation for adequacy of interfaces with State participating Tribal and local governments and of licensee drills, exercises, capabilities, and procedures. The results of the review, along with recommendations for improvements, must be documented, reported to the licensee's corporate and plant management, and retained for a period of 5 years. The part of the review involving the evaluation for adequacy of interface with State, participating Tribal and local governments must be available to the appropriate State, participating Tribal and local governments. (4) 
                                    *Security programs.*
                                     (i) The licensee must prepare and maintain safeguards contingency plan procedures in accordance with appendix C of part 73 of this chapter for affecting the actions and decisions contained in the Responsibility Matrix of the safeguards contingency plan. The licensee may not make a change that would decrease the safeguard effectiveness of a physical security plan, or guard training and qualification plan, or cybersecurity plan submitted under subpart H or part 73 of this chapter, or of the first four categories of information (Background, Generic Planning Base, Licensee Planning Base, Responsibility Matrix) contained in a licensee safeguards contingency plan submitted under subpart H or part 73 of this chapter, as applicable, without prior approval of the Commission. A licensee desiring to make such a change must submit an application for amendment to the licensee's license under §§ 53.1510, 53.1515, and 53.1520. (ii) The licensee may make changes to the plans referenced in paragraph (4)(i) of this section without prior Commission approval if the changes do not decrease the safeguards effectiveness of the plan. The licensee must maintain records of changes to the plans made without prior Commission approval for a period of 3 years from the date of the change, and must submit, as specified in § 53.040, a report containing a description of each change within 2 months after the change is made. Prior to the safeguards contingency plan being put into effect, the licensee must have— (A) All safeguards capabilities specified in the safeguards contingency plan available and functional; (B) Detailed procedures developed according to appendix C to part 73 of this chapter available at the licensee's site; and (C) All appropriate personnel trained to respond to safeguards incidents as outlined in the plan and specified in the detailed procedures. (iii) The licensee must provide for the development, revision, implementation, and maintenance of its safeguards contingency plan. The licensee must ensure that all program elements are reviewed by individuals independent of both security program management and personnel who have direct responsibility for implementation of the security program either— (A) At intervals not to exceed 12 months; or (B) As necessary, based on an assessment by the licensee against performance indicators, and as soon as reasonably practicable after a change occurs in personnel, procedures, equipment, or facilities that potentially could adversely affect security, but no longer than 12 months after the change. In any case, all elements of the safeguards contingency plan must be reviewed at least once every 24 months. (iv) The review must include a review and audit of safeguards contingency procedures and practices, an audit of the security system testing and maintenance program, and a test of the safeguards systems along with commitments established for response by local law enforcement authorities. The results of the review and audit, along with recommendations for improvements, must be documented, reported to the licensee's corporate and plant management, and kept available at the plant for inspection for a period of 3 years.
 § 53.1570
  (a) No commercial nuclear plant license issued under this part, or any right thereunder, shall be transferred, assigned, or in any manner disposed of, either voluntarily or involuntarily, directly or indirectly, through transfer of control of the license to any person, unless the Commission gives its consent in writing. (b)(1) An application for transfer of a license must include— (i) As much of the information described in §§ 53.1109, 53.1306, 53.1366, and 53.1413 with respect to the identity and technical and financial qualifications of the proposed transferee as would be required by those sections if the application were for an initial license. The Commission may require additional information such as data respecting proposed safeguards against hazards from radioactive materials and the applicant's qualifications to protect against such hazards. (ii) A statement of the purposes for which the transfer of the license is requested, the nature of the transaction necessitating or making desirable the transfer of the license, and an agreement to limit access to Restricted Data or Classified National Security Information pursuant to § 53.1115. The Commission may require any person who submits an application for license pursuant to the provisions of this section to file a written consent from the existing licensee or a certified copy of an order or judgment of a court of competent jurisdiction attesting to the person's right (subject to the licensing requirements of the Act and these regulations) to possession of the facility or site involved. (2) [Reserved] (c) After appropriate notice to interested persons, including the existing licensee, and observance of such procedures as may be required by the Act or regulations or orders of the Commission, the Commission will approve an application for the transfer of a license, if the Commission determines— (1) That the proposed transferee is qualified to be the holder of the license; and (2) That transfer of the license is otherwise consistent with applicable provisions of law, regulations, and orders issued by the Commission pursuant thereto.
 § 53.1575
  (a) When the holder of an OL or COL under this part has determined to permanently cease operations the licensee must, within 30 days, submit a written certification to the NRC, consistent with the requirements of § 53.1070. (b) Once fuel has been permanently removed from the reactor system, the licensee must submit a written certification to the NRC that meets the requirements of § 53.1070. (c)(1) Upon docketing of the certifications for permanent cessation of operations and permanent removal of fuel from the reactor system, or when a final legally effective order to permanently cease operations has come into effect, the license no longer authorizes operation of the reactor or 
                                    ( print page 87110)
                                    emplacement or retention of fuel into the reactor system. (2) Activities associated with decommissioning will be carried out in accordance with the requirements and procedures in subpart G of this part. (3) The Commission shall terminate the license if it determines that— (i) The remaining dismantlement has been performed in accordance with the approved license termination plan required in subpart G of this part; and (ii) The final radiation survey and associated documentation, including an assessment of dose contributions associated with parts released for use before approval of the license termination plan, demonstrate that the facility and site have met the criteria for decommissioning in subpart E of 10 CFR part 20. (d) A holder of a CP or COL under this part may request the termination of the license as well as licenses issued by the NRC under parts 30, 40, or 70 of this chapter prior to plant operations. Such requests may support an immediate NRC approval of the site for unrestricted use.
 § 53.1580
  Each licensee under this part must at any time before termination of the license, upon request of the Commission, submit, as specified in § 53.040 written statements, signed under oath or affirmation, to enable the Commission to determine whether or not the license should be modified, suspended, or revoked. Except for information sought to verify licensee compliance with the current licensing basis for that facility, the NRC must prepare the reason or reasons for each information request prior to issuance to ensure that the burden to be imposed on respondents is justified in view of the potential safety significance of the issue to be addressed in the requested information. Each such justification provided for an evaluation performed by the NRC staff must be approved by the Executive Director for Operations or his or her designee prior to issuance of the request.
 § 53.1585
  A license or standard design approval issued under this part may be revoked, suspended, or modified, in whole or in part, for any material false statement in the application or in the supplemental or other statement of fact required of the applicant; or because of conditions revealed by the application or statement of fact of any report, record, inspection, or other means which would warrant the Commission to refuse to grant a license or approval on an original application; or for failure to manufacture a reactor, or construct or operate a facility in accordance with the terms of the license, provided, however, that failure to make timely completion of the proposed construction or alteration of a facility under a CP under this part shall be governed by the provisions of § 53.1342(b); or for violation of, or failure to observe, any of the terms and provisions of the Act, regulations, license, approval, or order of the Commission.
 § 53.1590
  (a)(1) Backfitting means the modification of or addition to systems, structures, components, or design of a facility; or the design approval or ML for a facility; or the procedures or organization required to design, construct or operate a facility; any of which may result from a new or amended provision in the Commission's regulations or the imposition of a regulatory staff position interpreting the Commission's regulations that is either new or different from a previously applicable staff position after the date of the commercial nuclear plant license issued under this part. (2) Except as provided in paragraph (a)(4) of this section, the Commission shall require a systematic and documented analysis pursuant to paragraph (b) of this section for backfits which it seeks to impose. (3) Except as provided in paragraph (a)(4) of this section, the Commission shall require the backfitting of a facility only when it determines, based on the analysis described in paragraph (b) of this section, that there is a substantial increase in the overall protection of the public health and safety or the common defense and security to be derived from the backfit and that the direct and indirect costs of implementation for that facility are justified in view of this increased protection. (4) The provisions of paragraphs (a)(2) and (a)(3) of this section are inapplicable and, therefore, backfit analysis is not required and the standards in paragraph (a)(3) of this section do not apply where the Commission or staff, as appropriate, finds and declares, with appropriate documented evaluation for its finding, either— (i) That a modification is necessary to bring a facility into compliance with a license or the rules or orders of the Commission, or into conformance with written commitments by the licensee; or (ii) That regulatory action is necessary to ensure that the facility provides adequate protection to the health and safety of the public and is in accord with the common defense and security; or (iii) That the regulatory action involves defining or redefining what level of protection to the public health and safety or common defense and security should be regarded as adequate. (5) The Commission must always require the backfitting of a facility if it determines that such regulatory action is necessary to ensure that the facility provides adequate protection to the health and safety of the public and is in accord with the common defense and security. (6) The documented evaluation required by paragraph (a)(4) of this section must include a statement of the objectives of and reasons for the modification and the basis for invoking the exception. If immediately effective regulatory action is required, then the documented evaluation may follow rather than precede the regulatory action. (7) If there are two or more ways to achieve compliance with a license or the rules or orders of the Commission, or with written licensee commitments, or there are two or more ways to reach a level of protection which is adequate, then ordinarily the applicant or licensee is free to choose the way which best suits its purposes. However, should it be necessary or appropriate for the Commission to prescribe a specific way to comply with its requirements or to achieve adequate protection, then cost may be a factor in selecting the way, provided that the objective of compliance or adequate protection is met. (b) In reaching the determination required by paragraph (a)(3) of this section, the Commission will consider how the backfit should be scheduled in light of other ongoing regulatory activities at the facility and, in addition, will consider information available concerning any of the following factors as may be appropriate and any other information relevant and material to the proposed backfit: (1) The statement of the specific objectives that the proposed backfit is designed to achieve; (2) The general description of the activity that would be required by the licensee or applicant in order to complete the backfit; (3) The potential change in the risk to the public from the accidental off-site release of radioactive material; (4) The potential impact on radiological exposure of facility employees;
                                    ( print page 87111) (5) The installation and continuing costs associated with the backfit, including the cost of facility downtime or the cost of construction delay; (6) The potential safety impact of changes in plant or operational complexity, including the relationship to proposed and existing regulatory requirements; (7) The estimated resource burden on the NRC associated with the proposed backfit and the availability of such resources; (8) The potential impact of differences in facility type, design or age on the relevancy and practicality of the proposed backfit; (9) Whether the proposed backfit is interim or final and, if interim, the justification for imposing the proposed backfit on an interim basis. (c) No licensing action will be withheld during the pendency of backfit analyses required by the Commission's rules. (d) The Executive Director for Operations shall be responsible for implementation of this section, and all analyses required by this section shall be approved by the Executive Director for Operations or his or her designee.
 § 53.1595
  Licenses may be renewed by the Commission upon expiration of the period of the license.
 Subpart J—Reporting and Other Administrative Requirements § 53.1600
  Each applicant and licensee under this part must ensure that U.S. Nuclear Regulatory Commission (NRC) inspectors have unfettered access to sites and facilities licensed or proposed to be licensed in § 53.1610, must maintain records and make reports to the NRC in accordance with requirements in §§ 53.1620 through 53.1650, must satisfy financial qualification and reporting requirements in §§ 53.1660 through 53.1700, and must obtain and maintain required financial protections in case of an accident in §§ 53.1720 and 53.1730.
 § 53.1610
  (a) Each applicant for or holder of a manufacturing license (ML), operating license (OL), combined license (COL), construction permit (CP), or early site permit must permit inspection, by duly authorized representatives of the Commission, of its records, premises, activities, and of licensed materials in possession or use, related to the license or CP or early site permit as may be necessary to effectuate the purposes of the Atomic Energy Act of 1956, as amended, (the Act) and the Energy Reorganization Act of 1974, as amended. (b)(1) Each holder of an ML, OL, COL, or CP must, upon request by the Director, Office of Nuclear Reactor Regulation, provide rent-free office space for the exclusive use of the Commission inspection personnel. Heat, air conditioning, light, electrical outlets, and janitorial services must be furnished by each licensee and each holder of a CP. The office must be convenient to and have full access to the facility and must provide the inspectors both visual and acoustic privacy. (2) For a site or facility with an assigned resident inspector, the space provided must be adequate to accommodate a full-time inspector, a part-time secretary, and transient NRC personnel and must be generally commensurate with other office facilities at the site. For sites or facilities assigned multiple resident inspectors, additional space may be requested. The office space that is provided must be subject to the approval of the Director, Office of Nuclear Reactor Regulation. All furniture, supplies, and communication equipment will be furnished by the Commission. (3) For a site or facility without an assigned resident inspector, temporary space to accommodate periodic or special inspections must be provided. The office space must be generally commensurate with other office accommodations at the site. (4) The licensee or permit holder must afford any NRC resident inspector assigned to that site, or other NRC inspectors identified by the Regional Administrator as likely to inspect the facility, immediate unfettered access, equivalent to access provided regular plant employees, following proper identification and compliance with applicable access control measures for security, radiological protection, and personal safety. (5) The licensee or permit holder must ensure that the arrival and presence of an NRC inspector, who has been properly authorized facility access as described in paragraph (b)(4) of this section, is not announced or otherwise communicated by its employees or contractors to other persons at the facility unless specifically requested by the NRC inspector.
 § 53.1620
  (a) Each holder of an ML, OL, COL, CP, or early site permit must maintain all records and make all reports, in connection with the activity, as may be required by the conditions of the license or permit or by the regulations and orders of the Commission in effectuating the purposes of the Act and the Energy Reorganization Act of 1974, as amended. Reports must be submitted in accordance with § 53.040. (b) [Reserved] (c) Records that are required by the regulations in this part, by license condition, or by technical specifications must be retained for the period specified by the appropriate regulation, license condition, or technical specification. If a retention period is not otherwise specified, these records must be retained until the Commission terminates the facility license or, in the case of an early site permit, until the permit expires. (d)(1) Records which must be retained under this part may be the original or a reproduced copy or a microform if the reproduced copy or microform is duly authenticated by authorized personnel and the microform is capable of producing a clear and legible copy after storage for the period specified by Commission regulations. The record may also be stored in electronic media with the capability of producing legible, accurate, and complete records during the required retention period. Records such as letters, drawings, and specifications, must include all pertinent information such as stamps, initials, and signatures. The licensee must maintain adequate safeguards against tampering with, and loss of records. (2) If there is a conflict between the Commission's regulations in this part, license condition, or technical specification, or other written Commission approval or authorization pertaining to the retention period for the same type of record, the retention period specified in the regulations in this part for such records shall apply unless the Commission, under § 53.080 of this part, has granted a specific exemption from the record retention requirements in the regulations in this part. (e) Each licensee must notify the Commission as specified in § 53.040 of this part, of successfully completing power ascension testing or startup testing as applicable within 30 calendar days of completing the testing.
 § 53.1630
  (a) 
                                    *General requirements.*1
                                     (1) Each holder of an OL under this part or a COL under this part after the Commission makes the finding under § 53.1452(g), must notify the NRC Operations Center 
                                    ( print page 87112)
                                    via the Emergency Notification System (ENS) of— (i) The declaration of any of the Emergency Classes specified in the licensee's approved Emergency Plan; or (ii) Those non-emergency events specified in paragraph (b) of this section that occurred within 3 years of the date of discovery. (2) If the ENS is inoperative, the licensee must make the required notifications via commercial telephone service, other dedicated telephone system, or any other method which will ensure that a report is made as soon as practical to the NRC Headquarters Operations Center at the numbers specified in appendix A to part 73 of this chapter. (3) The licensee must notify the NRC immediately after notification of the appropriate State or local agencies and not later than 1 hour after the time the licensee declares one of the Emergency Classes. (4) The licensee must activate the data links with the NRC as specified in their emergency plans after declaring an Emergency Class for events of actual or potential substantial degradation of plant safety or security, probable risk to site personnel life, or site equipment damage caused by hostile action. The data links may also be activated by the licensee during emergency drills or exercises if the licensee's computer system has the capability to transmit the exercise data. (5) When making a report under paragraph (a)(1) of this section, the licensee must identify— (i) The Emergency Class declared; or (ii) Paragraph (b)(1), “One-hour reports,” paragraph (b)(2), “Four-hour reports,” or paragraph (b)(3), “Eight-hour reports,” as the paragraph of this section requiring notification of the non-emergency event. (b) 
                                    *Non-emergency events.*
                                     (1) 
                                    *One-hour reports.*
                                     If not reported as a declaration of an Emergency Class under paragraph (a) of this section, the licensee must notify the NRC as soon as practical and in all cases within one hour of the occurrence of any deviation from the plant's Technical Specifications authorized under § 53.740(h) of this part. (2) 
                                    *Four-hour reports.*
                                     If not reported under paragraphs (a) or (b)(1) of this section, the licensee must notify the NRC as soon as practical, and in all cases, within 4 hours of the occurrence of any of the following: (i) The initiation of any commercial nuclear plant shutdown required by the plant's Technical Specifications. (ii) Any event or condition that results in actuation of the reactor protection system when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. (iii) Any event or condition that results in an unplanned actuation of a safety-related (SR) standby cooling system or the unplanned sole reliance on an SR standby cooling system for those systems that are in constant operation. (iv) Any event or condition that results in an unplanned movement of, change of state in, or chemical interaction involving a significant amount of radioactive material within the commercial nuclear plant. (v) Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactively contaminated materials. (3) 
                                    *Eight-hour reports.*
                                     If not reported under paragraphs (a), (b)(1), or (b)(2) of this section, the licensee must notify the NRC as soon as practical and in all cases within 8 hours of the occurrence of any of the following: (i) Any event or condition that results in— (A) The condition of the commercial nuclear plant, including its principal safety barriers, being seriously degraded; or (B) The commercial nuclear plant being in a condition not analyzed under § 53.450 that significantly degrades plant safety. (ii) Any event or condition that results in valid actuation of an SR system, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. (iii) Any event or condition that at the time of discovery could have prevented the fulfilment of the safety functions identified under § 53.230. Events covered may include one or more procedural errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, and/or procedural inadequacies. However, individual component failures need not be reported pursuant to this paragraph if other equipment was operable and available to perform the required safety function. (iv) Any event requiring the transport of a radioactively contaminated person to an offsite medical facility for treatment. (v) Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (
                                    *e.g.,*
                                     significant portion of control room indication, ENS, or offsite notification system). (c) 
                                    *Follow-up notification:*
                                     With respect to the notifications made under paragraphs (a) and (b) of this section, in addition to making the required initial notification, each licensee, must during the course of the event— (1) Immediately Report: (i) any further degradation in the level of safety of the plant or other worsening plant conditions, including those that require the declaration of any of the Emergency Classes, if such a declaration has not been previously made, or (ii) any change from one Emergency Class to another, or (iii) a termination of the Emergency Class. (2) Immediately Report: (i) the results of ensuing evaluations or assessments of plant conditions, (ii) the effectiveness of response or protective measures taken, and (iii) important information related to plant behavior that is not understood. (3) Maintain an open, continuous communication channel with the NRC Operation Center upon request by the NRC. 1
                                         Other requirements for immediate notification of the NRC by licensed operating commercial nuclear plants are contained elsewhere in this chapter, in particular §§ 20.1906, 20.2202, 72.216, 73.77, and 73.1200 of this chapter.
 § 53.1640
  (a) 
                                    *Reportable events.*
                                     (1) Each commercial nuclear plant licensee holding an OL under this part or a COL under this part after the Commission makes the finding under § 53.1452(g), must submit a Licensee Event Report (LER) for any event of the type described in this paragraph within 60 days after discovery of the event. In the case of an invalid actuation reported under § 53.1640(a)(2), other than automatic reactor shutdown when the reactor is critical, the licensee may, at its option, provide a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER. Unless otherwise specified in this section, the licensee must report an event if it occurred within 3 years of the date of discovery regardless of the plant mode or power level, and regardless of the significance of the structure, system, or component that initiated the event. (2) The licensee must report— (i)(A) The completion of any commercial nuclear plant shutdown required by the plant's Technical Specifications.
                                    ( print page 87113) (B) Any operation or condition which was prohibited by the plant's Technical Specifications except when— *(1)*
                                     The Technical Specification is administrative in nature; *(2)*
                                     The event consisted solely of a case of a late surveillance test where the oversight was corrected, the test was performed, and the equipment was found to be capable of performing its specified safety functions; or *(3)*
                                     The Technical Specification was revised prior to discovery of the event such that the operation or condition was no longer prohibited at the time of the event. (C) Any deviation from the plant's Technical Specifications authorized under § 53.740(h). (ii) Any event or condition that resulted in— (A) The condition of the commercial nuclear plant, including its principal safety barriers, being seriously degraded; or (B) The commercial nuclear plant being in a condition not analyzed under § 53.450 that significantly degrades plant safety. (iii) Any natural phenomena or other external condition that posed an actual threat to the safety of the commercial nuclear plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the commercial nuclear plant. (iv) Any event or condition that resulted in inadvertent operation of any structures, systems, and component classified as SR for an identified safety function under § 53.460 or the unplanned sole reliance on an SR system for those systems that are in constant operation, except when— (A) The actuation resulted from and was part of a pre-planned sequence during testing; or (B) The actuation was invalid and— (
                                    *1*) Occurred while the system was properly removed from service; or (
                                    *2*) Occurred after the safety function had been already completed. (v) Any event or condition that could have prevented the fulfillment of the safety functions identified under § 53.230. (vi) Events covered in paragraph (a)(2)(v) of this section may include one or more procedural errors, equipment failures, and/or discovery of design, fabrication, construction, and/or procedural inadequacies. However, individual component failures need not be reported pursuant to paragraph (a)(2)(v) of this section if any other equipment was operable and available to perform the required safety function. (vii)(A) Any event or condition that as a result of a single cause could have prevented the fulfillment of any of the safety functions identified under § 53.230. (B) Events covered in paragraph (a)(2)(vii)(A) of this section may include cases of procedural error, equipment failure, and/or discovery of a design, analysis, fabrication, construction, and/or procedural inadequacy. However, licensees are not required to report an event pursuant to paragraph (a)(2)(vii)(A) of this section if the event results from— (
                                    *1*) A shared dependency among trains or channels that is a natural or expected consequence of the approved plant design; or (
                                    *2*) Normal and expected wear or degradation. (viii)(A) Any airborne radioactive release that, when averaged over a time period of 1-hour, resulted in airborne radionuclide concentrations in an unrestricted area that exceeds 20 times the applicable concentration limits specified in appendix B to 10 CFR part 20, table 2, column 1. (B) Any liquid effluent release that, when averaged over a time period of 1-hour, exceeds 20 times the applicable concentrations specified in appendix B to 10 CFR part 20, table 2, column 2, at the point of entry into the receiving waters (
                                    *i.e.,*
                                     unrestricted area) for all radionuclides except tritium and dissolved noble gases. (ix) Any event that posed an actual threat to the safety of the commercial nuclear plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the plant, including fires, toxic gas releases, or radioactive releases. (b) 
                                    *Contents.*
                                     The LER must contain— (1) A brief abstract describing the major occurrences during the event, including all component or system failures that contributed to the event and significant corrective action taken or planned to prevent recurrence. (2)(i) A clear, specific narrative description of what occurred so that knowledgeable readers conversant with the design of commercial nuclear plants, but not familiar with the details of a particular plant, can understand the complete event. (ii) The narrative description must include the following specific information as appropriate for the particular event: (A) Plant operating conditions before the event. (B) Status of systems, structures, or components that were inoperable at the start of the event and that contributed to the event. (C) Dates and approximate time of the occurrences. (D) The cause of each component or system failure or personnel error, if known. (E) The failure mode, mechanism, and effect of each failed component, if known. (F) [Reserved] (G) For failures of components with multiple functions, include a list of systems or secondary functions that were also affected. (H) For failure that rendered a component or system classified as SR or non-safety-related but safety-significant inoperable, an estimate of the elapsed time from the discovery of the failure until the component or system was returned to service. (I) The method of discovery of each component or system failure or procedural error. (J) For each human performance related root cause, the licensee must discuss the cause(s) and circumstances. (K) Automatically and manually initiated safety system responses. (L) The manufacturer and model number (or other identification) of each component that failed during the event. (3) An assessment of the safety consequences and implications of the event. This assessment must include— (i) The availability of systems or components that could have performed the same function as the components and systems that failed during the event, and (ii) For events that occurred when the reactor was shut down, the availability of systems or components that are needed to shut down the reactor and maintain safe shutdown conditions, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident. (4) A description of any corrective actions planned as a result of the event, including those to reduce the probability of similar events occurring in the future. (5) Reference to any previous similar events at the same plant that are known to the licensee. (6) The name and contact information of a person within the licensee's organization who is knowledgeable about the event and can provide additional information concerning the event and the plant's characteristics. (c) 
                                    *Supplemental Information.*
                                     The Commission may require the licensee to submit specific additional information beyond that required by paragraph (b) of this section if the Commission finds that supplemental material is necessary for complete understanding of an unusually complex or significant event. These requests for supplemental information will be made in writing and the licensee 
                                    ( print page 87114)
                                    must submit, as specified in § 53.040, the requested information as a supplement to the initial LER. (d) 
                                    *Submission of Reports.*
                                     Licensee Event Reports must be prepared on Form NRC 366 and submitted to the NRC, as specified in § 53.040. (e) 
                                    *Report Legibility.*
                                     The reports and copies that licensees are required to submit to the Commission under the provisions of this section must be of sufficient quality to permit legible reproduction and micrographic processing.
 53.1645
  (a) Each holder of an OL, and each holder of a COL after the Commission has made the finding under § 53.1452(g), must submit radiological reports as required by 10 CFR part 20, as well as an Annual Radioactive Effluent Release Report and an Annual Radiological Environmental Operating Report. The Annual Radioactive Effluent Release Report must specify the quantity of each of the principal radionuclides released to unrestricted areas in liquid and in gaseous effluents and an estimate of the dose received by the maximally exposed member of the public in an unrestricted area from effluents and direct radiation from contained sources during the previous calendar year. The Annual Radiological Environmental Operating Report must provide data on measurable levels of radiation and radioactive materials in the environment, must include an evaluation of the relationship between quantities of radioactive material released in effluents and resultant radiation doses to individuals from principal pathways of exposure, and must include the results of environmental monitoring during the previous calendar year. These reports must also include any other information as may be required by the Commission to estimate maximum potential annual radiation doses to the public. The reports must be submitted as specified in § 53.040 by May 15 of each successive year. If the total effective dose equivalent to members of the public in unrestricted areas during the reporting period is greater than the as low as is reasonably achievable (ALARA) design objectives established under § 53.425, the report must specify the causes for exceeding the ALARA design objective and describe any corrective actions. On the basis of these reports and any additional information the Commission may obtain from the licensee or others, the Commission may require the licensee to take action as the Commission deems appropriate. (b) If during any calendar quarter the radiation exposure to a member of the public in the unrestricted areas, calculated on the same basis as the respective ALARA design objective exposure, exceeds one-half of the annual ALARA design objective exposure, the licensee must submit a report as specified in § 53.040. The report shall specify the causes for exceeding one-half the annual ALARA design objective exposure in a quarter and describe corrective actions that the licensee will take to maintain radiation exposure to levels within the ALARA design objectives for the remainder of the year. The report shall be submitted within 30 days from the end of the quarter when one-half of the annual ALARA design objective exposure was exceeded.
 § 53.1650
  (a) In response to a written request by the Commission, each applicant for a CP or license and each recipient of a CP or a license must submit facility information, as described in § 75.10 of this chapter, on International Atomic Energy Agency (IAEA) Design Information Questionnaire forms and site information on DOC/NRC Form AP-A and associated forms; (b) As required by the Additional Protocol, must submit location information described in § 75.11 of this chapter on DOC/NRC Form AP-1 and associated forms; and (c) Must permit verification thereof by the IAEA and take other action as necessary to implement the US/IAEA Safeguards Agreement, as described in part 75 of this chapter.
 § 53.1660
  Sections 53.1670 through 53.1700 set out the requirements and procedures related to financial qualifications and related reporting requirements.
 § 53.1670
  Except for an electric utility applicant for a license to operate a commercial nuclear plant, an applicant for a CP, OL, or COL under this part must possess or have reasonable assurance of obtaining the funds necessary for the activities for which the permit or license is sought.
 § 53.1680
  With respect to any commercial nuclear plant of a type described in § 53.020, each licensee and each holder of a CP must submit its annual financial report, including the certified financial statements, to the Commission, as specified in § 53.040, upon issuance of the report. However, licensees and holders of a CP who submit a Form 10-Q with the Securities and Exchange Commission or a Form 1 with the Federal Energy Regulatory Commission need not submit the annual financial report or the certified financial statement under this section.
 § 53.1690
  (a) An electric utility licensee holding an OL or COL (including a renewed license) for a commercial nuclear plant, no later than seventy-five (75) days prior to ceasing to be an electric utility in any manner not involving a license transfer under § 53.1399 or § 53.1456 must provide the NRC with the financial qualifications information that would be required for obtaining an initial OL or COL under this part. The financial qualifications information must address the first full 5 years of operation after the date the licensee ceases to be an electric utility. (b)(1) Any holder of a license issued under this part must notify the appropriate NRC Regional Administrator, in writing, immediately following the filing of a voluntary or involuntary petition for bankruptcy under any chapter of title 11 (Bankruptcy) of the United States Code by or against— (i) The licensee; (ii) An entity (as 11 U.S.C. 101(14) defines that term) controlling the licensee or listing the license or licensee as property of the estate; or (iii) An affiliate (as 11 U.S.C. 101(2) defines that term) of the licensee. (2) This notification must indicate— (i) The bankruptcy court in which the petition for bankruptcy was filed; and (ii) The date of the filing of the petition.
 § 53.1700
  (a) Pursuant to section 184 of the Act, the Commission consents, without individual application, to the creation of any mortgage, pledge, or other lien upon any facility not owned by the United States which is the subject of a license or upon any leasehold or other interest in such facility; provided— (1) That the rights of any creditor so secured may be exercised only in compliance with and subject to the same requirements and restrictions as would apply to the licensee pursuant to the provisions of the license, the Act, and regulations issued by the Commission under the Act; and (2) That no creditor so secured may take possession of the facility pursuant to the provisions of this section prior to either the issuance of a license from the Commission authorizing such possession or the transfer of the license.
                                    ( print page 87115) (b) Any creditor so secured may apply for transfer of the license covering such facility by filing an application for transfer of the license under § 53.1570. The Commission will act upon such application under subpart I of this part. (c) Nothing contained in this regulation shall be deemed to affect the means of acquiring, or the priority of, any tax lien or other lien provided by law. (d) As used in this section— *License*
                                     includes any license under this part, which may be issued by the Commission with regard to a facility. *Creditor*
                                     includes, without implied limitation, the trustee under any mortgage, pledge or lien on a facility made to secure any creditor, any trustee or receiver of the facility appointed by a court of competent jurisdiction in any action brought for the benefit of any creditor secured by such mortgage, pledge or lien, any purchaser of such facility at the sale thereof upon foreclosure of such mortgage, pledge, or lien or upon exercise of any power of sale contained therein, or any assignee of any such purchaser. *Facility*
                                     includes, but is not limited to, a site which is the subject of an early site permit under this part, and a reactor manufactured under an ML under this part.
 § 53.1710
  Sections 53.1720 and 53.1730 set out the requirements and procedures related to licensees obtaining and maintaining insurance to cover stabilization and decontamination activities in the event of an accident and financial protection in accordance with part 140, “Financial Protection Requirements and Indemnity Agreements,” of this chapter.
 § 53.1720
  Each commercial nuclear plant licensee under this part must take reasonable steps to obtain insurance available at reasonable costs and on reasonable terms from private sources or to demonstrate that it possesses an equivalent amount of protection covering the licensee's obligation, in the event of an accident at the licensee's commercial nuclear reactor, to stabilize and decontaminate the plant and the plant site at which such an accident may occur, provided that— (a) The insurance required by this section must have a minimum coverage limit for each commercial nuclear plant site of $1.06 billion, an amount based on plant-specific estimates of costs to stabilize and decontaminate a plant, or whatever amount of insurance is generally available from private sources, whichever is less. The required insurance must clearly state that, as and to the extent provided in paragraph (d)(1) of this section, any proceeds must be payable first for stabilization of the plant and next for decontamination of the plant and the plant site. If a licensee's coverage falls below the required minimum, the licensee must within 60 days take all reasonable steps to restore its coverage to the required minimum. The required insurance may, at the option of the licensee, be included within policies that also provide coverage for other risks, including, but not limited to, the risk of direct physical damage. (b)(1) With respect to policies issued or annually renewed, the proceeds of such required insurance must be dedicated, as and to the extent provided in this paragraph, to reimbursement or payment on behalf of the insured of reasonable expenses incurred or estimated to be incurred by the licensee in taking action to fulfill the licensee's obligation, in the event of an accident at the licensee's plant, to ensure that the plant is in, or is returned to, and maintained in, a safe and stable condition and that radioactive contamination is removed or controlled such that personnel exposures are consistent with the occupational exposure limits in 10 CFR part 20. These actions must be consistent with any other obligation the licensee may have under this chapter and must be subject to paragraph (d) of this section. As used in this section, an “accident” means an event that involves the release of radioactive material from its intended place of confinement within the commercial nuclear plant such that there is a present danger of release off site in amounts that would pose a threat to the public health and safety. (2) The stabilization and decontamination requirements set forth in paragraph (d) of this section must apply uniformly to all insurance policies required under this section. (c) The licensee shall report to the NRC on April 1 of each year the current levels of this insurance or financial security it maintains and the sources of this insurance or financial security. (d)(1) In the event of an accident at the licensee's plant, whenever the estimated costs of stabilizing the licensed plant and of decontaminating the plant and the plant site exceed one tenth of the minimum insurance under paragraph (a) of this section, the proceeds of the insurance required by this section must be dedicated to and used, first, to ensure that the licensed plant is in, or is returned to, and can be maintained in, a safe and stable condition so as to prevent any significant risk to the public health and safety and, second, to decontaminate the plant and the plant site in accordance with the licensee's cleanup plan as approved by order of the Director, Office of Nuclear Reactor Regulation. This priority on insurance proceeds must remain in effect for 60 days or, upon order of the Director, for such longer periods, in increments not to exceed 60 days except as provided for activities under the cleanup plan required in paragraphs (d)(3) and (d)(4) of this section, as the Director may find necessary to protect the public health and safety. Actions needed to bring the plant to and maintain the plant in a safe and stable condition may include one or more of the following, as appropriate: (i) Shutdown of the reactor(s) and other processes at the plant; (ii) Establishment and maintenance of long-term cooling with stable decay heat removal; (iii) Maintenance of sub-criticality; (iv) Control of radioactive releases; and (v) Securing of structures, systems, or components to minimize radiation exposure to onsite personnel or to the offsite public or to facilitate later decontamination or both. (2) The licensee must inform the Director, Office of Nuclear Reactor Regulation in writing when the plant is and can be maintained in a safe and stable condition so as to prevent any significant risk to the public health and safety. Within 30 days after the licensee informs the Director that the plant is in this condition, or at such earlier time as the licensee may elect or the Director may for good cause direct, the licensee must prepare and submit a cleanup plan for the Director's approval. The cleanup plan must identify and contain an estimate of the cost of each cleanup operation that will be required to decontaminate the reactor sufficiently to permit the licensee either to resume operation of the reactor or to apply to the Commission under subpart G of this part for authority to decommission the reactor and to surrender the license voluntarily. Cleanup operations may include one or more of the following, as appropriate: (i) Processing any contaminated materials generated by the accident and by decontamination operations to remove radioactive materials; (ii) Decontamination of surfaces inside the plant buildings to levels consistent with the Commission's occupational exposure limits in 10 CFR part 20, and decontamination or disposal of equipment;
                                    ( print page 87116) (iii) Decontamination or removal and disposal of internal parts, damaged fuel from the reactor coolant or fuel systems, or related process or waste systems; and (iv) Cleanup of the reactor coolant or fuel systems or related process or waste systems. (3) Following review of the licensee's cleanup plan, the Director will order the licensee to complete all operations that the Director finds are necessary to decontaminate the reactor sufficiently to permit the licensee either to resume operation of the reactor or to apply to the Commission under subpart G of this part for authority to decommission the reactor and to surrender the license voluntarily. The Director must approve or disapprove, in whole or in part for stated reasons, the licensee's estimate of cleanup costs for such operations. Such order may not be effective for more than one year, at which time it may be renewed. Each subsequent renewal order, if imposed, may be effective for not more than 6 months. (4) Of the balance of the proceeds of the required insurance not already expended to place the plant in a safe and stable condition under paragraph (b)(1) of this section, an amount sufficient to cover the expenses of completion of those decontamination operations that are the subject of the Director's order must be dedicated to such use, provided that, upon certification to the Director of the amounts expended previously and from time to time for stabilization and decontamination and upon further certification to the Director as to the sufficiency of the dedicated amount remaining, policies of insurance may provide for payment to the licensee or other loss payees of amounts not so dedicated, and the licensee may proceed to use in parallel (and not in preference thereto) any insurance proceeds not so dedicated for other purposes.
 § 53.1730
  Commercial nuclear plant licensees must satisfy the applicable provisions of part 140, “Financial Protection Requirements and Indemnity Agreements,” of this chapter.
 Subparts K and L [Reserved]
 Subpart M—Enforcement § 53.9000
  (a) The Commission may obtain an injunction or other court order to prevent a violation of the provisions of— (1) The Atomic Energy Act of 1954, as amended (the Act); (2) Title II of the Energy Reorganization Act of 1974, as amended; or (3) A regulation or order issued under those Acts. (b) The Commission may obtain a court order for the payment of a civil penalty imposed under Section 234 of the Act: (1) For violations of— (i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of the Act; (ii) Section 206 of the Energy Reorganization Act of 1974, as amended; (iii) Any rule, regulation, or order issued under the sections specified in paragraph (b)(1)(i) of this section; (iv) Any term, condition, or limitation of any license issued under the sections specified in paragraph (b)(1)(i) of this section. (2) For any violation for which a license may be revoked under section 186 of the Act.
 § 53.9010
  (a) Section 223 of the Act provides for criminal sanctions for willful violation of, attempted violation of, or conspiracy to violate, any regulation issued under sections 161b, 161i, or 161o of the Act. For purposes of section 223, all the regulations in part 53 are issued under one or more of sections 161b, 161i, or 161o, except for the sections listed in paragraph (b) of this section. (b) The regulations in 10 CFR part 53 that are not issued under sections 161b, 161i, or 161o for the purposes of section 223 are as follows: §§ 53.000, 53.015, 53.020, 53.040, 53.080, 53.090, 53.100, 53.110, 53.120, 53.600, 53.725, 53.726, 53.735, 53.760, 53.775, 53.790, 53.795, 53.820, 53.910, 53.1000, 53.1050, 53.1100, 53.1103, 53.1106, 53.1109, 53.1112, 53.1115, 53.1118, 53.1120, 53.1121, 53.1124, 53.1140, 53.1143, 53.1144, 53.1146, 53.1149, 53.1155, 53.1158, 53.1164, 53.1170, 53.1173, 53.1176, 53.1179, 53.1188, 53.1200, 53.1203, 53.1206, 53.1209, 53.1210, 53.1212, 53.1215, 53.1218, 53.1221, 53.1230, 53.1236, 53.1239, 53.1241, 53.1242, 53.1245, 53.1248, 53.1251, 53.1254, 52.1257, 52.1260, 53.1263, 53.1270, 53.1273, 53.1276, 53.1279, 53.1282, 53.1285, 53.1286, 53.1287, 53.1288, 53.1291, 53.1293, 53.125, 53.1300, 53.1306, 53.1309, 53.1312, 53.1315, 53.1318, 53.1324, 53.1330, 53.1333, 53.1336, 53.1348, 53.1360, 53.1366, 53.1369, 53.1372, 53.1375, 53.1381, 53.1384, 53.1387, 53.1390, 53.1396, 53.1401, 53.1405, 53.1410, 53.1416, 53.1419, 53.1422, 53.1425, 53.1431, 53.1437, 53.1440, 53.1443, 53.1452, 53.1455, 53.1456, 53.1458, 53.1461, 53.1470, 53.1500, 53.1510, 53.1515, 53.1520, 53.1525, 53.1530, 53.1535, 53.1540, 53.1560, 53.1585, 53.1590, 53.1595, 53.1600, 53.1660, 53.1670, 53.1700, 53.1710, 53.1730, 53.9000, 53.9010.
 PART 70—DOMESTIC LICENSING OF SPECIAL NUCLEAR MATERIAL
 129.The authority citation for 10 CFR part 70 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 51, 53, 57(d), 108, 122, 161, 182, 183, 184, 186, 187, 193, 223, 234, 274, 1701 (42 U.S.C. 2071, 2073, 2077(d), 2138, 2152, 2201, 2232, 2233, 2234, 2236, 2237, 2243, 2273, 2282, 2021, 2297f); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); Nuclear Waste Policy Act of 1982, secs. 135, 141 (42 U.S.C. 10155, 10161); 44 U.S.C. 3504 note. § 70.20a
 130.In § 70.20a, in paragraph (b) remove the phrase “parts 30 through 36, 39, 40, 50, 72, 110,” and add in its place the phrase “parts 30 through 36, 39, 40, 50, 53, 72, 110”. § 70.22
 131.In § 70.22, wherever it appears, remove the phrase “part 50” and add in its place the phrase “parst 50 or 53”. 132.In § 70.24, revise paragraph (d) to read as follows: § 70.24
  
 (d)(1) The requirements in paragraphs (a) through (c) of this section do not apply to a holder of a construction permit or operating license for a nuclear power reactor issued under part 50 or part 53 of this chapter or a combined license issued under part 52 or part 53 of this chapter, if the holder complies with the requirements of paragraph (b) of 10 CFR 50.68 or paragraph (m)(2) of 10 CFR 53.440,as applicable. (2) An exemption from § 70.24 held by a licensee who thereafter elects to comply with requirements of paragraph (b) of 10 CFR 50.68 or paragraph (m)(2) of 10 CFR 53.440 does not exempt that licensee from complying with any of the requirements in § 50.68 or § 53.440(m) of this chapter but shall be ineffective so long as the licensee elects to comply with § 50.68(b) or § 53.440(m)(2) of this chapter, as applicable.
 § 70.32
 133.In § 70.32, in paragraph (c)(1) introductory text, remove the phrase “part 50 of this chapter” and add in its place the phrase “parts 50 or 53 of this chapter”; and in paragraph (d) remove the phrase “or § 70.34 of this chapter, as appropriate.” and add in its place the phrase “, §§ 74.34 or 53.1510 of this chapter, as appropriate.”. ( print page 87117) 134.In § 70.50, revise paragraph (d) to read as follows: § 70.50
  
 (d) The provisions of § 70.50 do not apply to licensees subject to §§ 50.72 or 53.1630 of this chapter. They do apply to those 10 CFR parts 50 or 53 licensees possessing material licensed under 10 CFR part 70 that are not subject to the notification requirements in §§ 50.72 or 53.1630 of this chapter.
 PART 72—LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE
 135.The authority citation for 10 CFR part 72 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 183, 184, 186, 187, 189, 223, 234, 274 (42 U.S.C. 2071, 2073, 2077, 2092, 2093, 2095, 2099, 2111, 2201, 2210e, 2232, 2233, 2234, 2236, 2237, 2238, 2273, 2282, 2021); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); National Environmental Policy Act of 1969 (42 U.S.C. 4332); Nuclear Waste Policy Act of 1982, secs. 117(a), 132, 133, 134, 135, 137, 141, 145(g), 148, 218(a) (42 U.S.C. 10137(a), 10152, 10153, 10154, 10155, 10157, 10161, 10165(g), 10168, 10198(a)); 44 U.S.C. 3504 note. 136.In § 72.3, revise the definition for “ *Independent spent fuel storage installation or ISFSI*
                        ” to read as follows: § 72.3
  
 *Independent spent fuel storage installation or ISFSI*
                             means a complex designed and constructed for the interim storage of spent nuclear fuel, solid reactor-related GTCC waste, and other radioactive materials associated with spent fuel and reactor-related GTCC waste storage. An ISFSI which is located on the site of another facility licensed under this part or a facility licensed under part 50 or part 53 of this chapter and which shares common utilities and services with that facility or is physically connected with that other facility may still be considered independent.
 137.In § 72.30, revise paragraph (e)(5) to read as follows: § 72.30
  
 (e) * * * (5) In the case of licensees who are issued a power reactor license under parts 50 or 53 of this chapter or ISFSI licensees who are an electric utility, as defined in parts 50 or 53 of this chapter, with a specific license issued under this part, the methods of §§ 50.75(b), (e), and (h) or 53.1010, 53.1040, 53.1045(b), and 53.1060 of this chapter, as applicable. In the event that funds remaining to be placed into the licensee's ISFSI decommissioning external sinking fund are no longer approved for recovery in rates by a competent rate making authority, the licensee must make changes to provide financial assurance using one or more of the methods stated in paragraphs (a)(1) through (4) of this section.
 138.In § 72.32, revise paragraph (c)(2) to read as follows: § 72.32
  
 (c) * * * (2)(i) Located within the exclusion area as defined in 10 CFR part 100, of a nuclear power reactor licensed for operation by the Commission, the emergency plan that meets either the requirements in § 50.160 of this chapter or the requirements in appendix E to part 50 of this chapter and § 50.47(b) of this chapter shall be deemed to satisfy the requirements of this section. (ii) Located within the exclusion area, as defined in 10 CFR part 53, of a commercial nuclear plant licensed for operation by the Commission, the emergency plan that meets either the requirements in § 50.160 of this chapter or the requirements in appendix E to part 50 of this chapter and § 50.47(b) of this chapter shall be deemed to satisfy the requirements of this section.
 § 72.40
 139.In § 72.40, in paragraph (c) remove the phrase “under part 50 of this chapter,” and add in its place the phrase “under parts 50 or 53 of this chapter,”. 140.In § 72.75, revise paragraph (i)(1)(ii) to read as follows: § 72.75
  
 (i) * * * (1) * * * (ii) Licensees issued a general license under § 72.210, after the licensee has placed spent fuel on the ISFSI storage pad (if the ISFSI is located inside the collocated protected area, for a reactor licensed under parts 50 or 53 of this chapter) or after the licensee has transferred spent fuel waste outside the reactor licensee's protected area to the ISFSI storage pad (if the ISFSI is located outside the collocated protected area, for a reactor licensed under parts 50 or 53 of this chapter).
 § 72.184
 141.In § 72.184, in paragraph (a) remove the phrase “under part 50 of this chapter” and add in its place the phrase “under parts 50 or 53 of this chapter”. 142.Revise § 72.210 to read as follows: § 72.210
  A general license is hereby issued for the storage of spent fuel in an independent spent fuel storage installation at power reactor sites to persons authorized to possess or operate nuclear power reactors under 10 CFR parts 50, 52, or 53.
 143.In § 72.212, revise paragraph (b)(8) to read as follows: § 72.212
  
 (b) * * * (8) Before use of the general license, determine whether activities related to storage of spent fuel under this general license involve a change in the facility Technical Specifications or require a license amendment for the facility pursuant to §§ 50.59(c) or 53.1550 of this chapter. Results of this determination must be documented in the evaluations made in paragraph (b)(5) of this section.
 144.In § 72.218, revise paragraphs (a) and (b) to read as follows: § 72.218
  (a) The notification regarding the program for the management of spent fuel at the reactor required by §§ 50.54(bb) or 53.1060 of this chapter must include a plan for removal of the spent fuel stored under this general license from the reactor site. The plan must show how the spent fuel will be managed before starting to decommission systems and components needed for moving, unloading, and shipping this spent fuel. (b) An application for termination of a reactor operating license issued under 10 CFR part 50 and submitted under § 50.82 of this chapter, or a combined license issued under 10 CFR part 52 and submitted under § 52.110 of this chapter, or a reactor operating or combined license under 10 CFR part 53 and submitted under § 53.1070 of this chapter must contain a description of how the spent fuel stored under this 
                            ( print page 87118)
                            general license will be removed from the reactor site.
 PART 73—PHYSICAL PROTECTION OF PLANTS AND MATERIALS
 145.The authority citation for 10 CFR part 73 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 53, 147, 149, 161, 161A, 170D, 170E, 170H, 170I, 223, 229, 234, 1701 (42 U.S.C. 2073, 2167, 2169, 2201, 2201a, 2210d, 2210e, 2210h, 2210i, 2273, 2278a, 2282, 2297f); Energy Reorganization Act of 1974, secs. 201, 202 (42 U.S.C. 5841, 5842); Nuclear Waste Policy Act of 1982, secs. 135, 141 (42 U.S.C. 10155, 10161); 44 U.S.C. 3504 note. Section 73.37(b)(2) also issued under sec. 301, Pub. L. 96-295, 94 Stat. 789 (42 U.S.C. 5841 note). 146.In § 73.1, revise paragraph (b)(1)(i) to read as follows: § 73.1
  
 (b) * * * (1) * * * (i) The physical protection of production and utilization facilities licensed under parts 50, 52, or 53 of this chapter,
 147.In § 73.2, revise the introductory text and paragraph (a) to read as follows: § 73.2
  As used in this part: (a) Terms defined in parts 50, 52, 53, 70, and 95 of this chapter have the same meaning when used in this part.
 148.In § 73.8, revise paragraph (b) to read as follows: § 73.8
  
 (b) The approved information collection requirements contained in this part appear in §§ 73.5, 73.15, 73.17, 73.20, 73.21, 73.24, 73.25, 73.26, 73.27, 73.37, 73.40, 73.45, 73.46, 73.50, 73.54, 73.55, 73.56, 73.57, 73.58, 73.60, 73.67, 73.70, 73.72, 73.73, 73.74, 73.77, 73.100, 73.110, 73.120, 73.1200, 73.1205, 73.1210, 73.1215, and appendices B and C to this part.
 149.In § 73.50, revise the introductory text to read as follows: § 73.50
  Each licensee who is not subject to § 73.51, but who possesses, uses, or stores formula quantities of strategic special nuclear material that are not readily separable from other radioactive material and which have a total external radiation level in excess of 1 gray (100 rad) per hour at a distance of 1 meter (3.3 feet) from any accessible surfaces without intervening shielding other than at a nuclear reactor facility licensed under parts 50, 52, or 53 of this chapter, shall comply with the following:
 150.In § 73.55, revise paragraphs (a)(4) and (6), (i)(4)(iii), (l)(1), (l)(7)(ii), (p)(1)(i), (r)(2) and (r)(4)(iii) to read as follows: § 73.55
  (a) * * * (4) Applicants for an operating license under the provisions of part 50 or part 53 of this chapter or holders of a combined license under the provisions of part 52 or part 53 of this chapter shall implement the requirements of this section before fuel is allowed onsite (protected area). 
 (6) Applicants for an operating license under the provisions of part 50 or part 53 of this chapter, or holders of a combined license under the provisions of part 52 or part 53 of this chapter that do not reference a standard design certification or reference a standard design certification issued after May 26, 2009, shall meet the requirement of § 73.55(i)(4)(iii). 
 (i) * * * (4) * * * (iii) Applicants for an operating license under the provisions of part 50 of this chapter, or holders of a combined license under the provisions of part 52 of this chapter, or licensees under part 53 of this chapter that elect to demonstrate compliance with § 73.55, consistent with § 53.860(a)(2) of this chapter, shall construct, locate, protect, and equip both the central and secondary alarm stations to the standards for the central alarm station contained in this section. Both alarm stations shall be equal and redundant, such that all functions needed to satisfy the requirements of this section can be performed in both alarm stations. 
 (l) * * * (1) Commercial nuclear power reactors licensed under 10 CFR parts 50, 52, or 53 and authorized to use special nuclear material in the form of MOX fuel assemblies containing up to 20 weight percent PuO2
                             shall, in addition to demonstrating compliance with the requirements of this section, protect un-irradiated MOX fuel assemblies against theft or diversion as described in this paragraph. 
 (7) * * * (ii) Additional measures for the physical protection of un-irradiated MOX fuel assemblies containing greater than 20 weight percent PuO2
                             shall be determined by the Commission on a case-by-case basis and documented through license amendment in accordance with §§ 50.90 or 53.1510 of this chapter. 
 (p) * * * (1) * * * (i) Under §§ 50.54(x) and (y) or 53.740(h) of this chapter, the licensee may suspend any security measures under this section in an emergency when this action is immediately needed to protect the public health and safety and no action consistent with license conditions and technical specifications that can provide adequate or equivalent protection is immediately apparent. This suspension of security measures must be approved as a minimum by a licensed senior operator before taking this action. 
 (r) * * * (2) The licensee shall submit proposed alternative measure(s) to the Commission for review and approval under §§ 50.4 and 50.90, or §§ 53.040 and 53.1510 of this chapter before implementation. 
 (4) * * * (iii) Based on comparison of the costs of the alternative measures to the costs of demonstrating compliance with the Commission's requirements using the essential elements of §§ 50.109 or 53.1590 of this chapter, the costs of fully demonstrating compliance with the Commission's requirements are not justified by the protection that would be provided.
 151.In § 73.56, revise paragraph (a)(3) to read as follows: § 73.56
  (a) * * * (3) Each applicant for an operating license under the provisions of part 50 of this chapter, each holder of a combined license under the provisions of part 52 of this chapter, and applicants for an operating license or holders of a combined license under part 53 of this chapter that do not meet the requirements of § 53.860(a)(2) of this chapter, shall implement the 
                            ( print page 87119)
                            requirements of this section before fuel is allowed on site (protected area).
 152.In § 73.57, revise paragraph (a)(3) to read as follows: § 73.57
  (a) * * * (3) Before receiving its operating license under 10 CFR parts 50 or 53 or before the Commission makes its finding under §§ 52.103(g) or 53.1452(g) of this chapter, each applicant for a license to operate a nuclear power reactor (including an applicant for a combined license) or a non-power reactor may submit fingerprints for those individuals who will require unescorted access to the nuclear power facility or non-power reactor facility.
 153.In § 73.58, revise paragraph (a) to read as follows: § 73.58
  (a) Each operating nuclear power reactor licensee with a license issued under parts 50, 52, or 53 of this chapter shall comply with the requirements of this section.
 154.In § 73.67, revise paragraphs (d) introductory text and (f) introductory text to read as follows: § 73.67
  
 (d) 
                            *Fixed site requirements for special nuclear material of moderate strategic significance.*
                             Each licensee who possesses, stores, or uses quantities and types of special nuclear material of moderate strategic significance at a fixed site or contiguous sites, except as allowed by paragraph (b)(2) of this section and except those who are licensed to operate a nuclear power reactor pursuant to part 50 or part 53, provided that the special nuclear material is located within a protected area and protected under § 73.55 or § 73.100, shall: 
 (f) 
                            *Fixed site requirements for special nuclear material of low strategic significance.*
                             Each licensee who possesses, stores, or uses special nuclear material of low strategic significance at a fixed site or contiguous sites, except those who are licensed to operate a nuclear power reactor pursuant to part 50 or part 53, provided that the special nuclear material is located within a protected area and protected under § 73.55 or § 73.100, shall:
 155.In § 73.77, revise paragraphs (a), (b)(1), (c)(6) and (7) to read as follows: § 73.77
  (a) Each licensee subject to the provisions of §§ 73.54 or 73.110 shall notify the NRC Headquarters Operations Center via the Emergency Notification System (ENS), under paragraph (c) of this section: (1) Within one hour after discovery of a cyberattack that adversely impacted: (i) Safety-related or important-to-safety functions, security functions, or emergency preparedness functions (including offsite communications); or that compromised support systems and equipment resulting in adverse impacts to safety, security, or emergency preparedness functions within the scope of § 73.54; or, (ii) Functions performed by digital assets that would prevent a postulated fission product release resulting in offsite doses exceeding the values in § 53.210 of this chapter, or functions performed by digital assets used by the licensee for implementing the physical security requirements in § 53.860(a) of this chapter. (2) Within 4 hours: (i) After discovery of a cyberattack that could have caused an adverse impact to: (A) Safety-related or important-to-safety functions, security functions, or emergency preparedness functions (including offsite communications); or that could have compromised support systems and equipment, which if compromised, could have adversely impacted safety, security, or emergency preparedness functions within the scope of § 73.54; or, (B) Functions performed by digital assets that would prevent a postulated fission product release resulting in offsite doses exceeding the values in § 53.210 of this chapter, or functions performed by digital assets used by the licensee for implementing the physical security requirements in § 53.860(a) of this chapter. (ii) After discovery of a suspected or actual cyberattack initiated by personnel with physical or electronic access to digital computer and communication systems and networks within the scope of §§ 73.54 or 73.110. (iii) After notification of a local, State, or other Federal agency (
                            *e.g.,*
                             law enforcement, Federal Bureau of Investigation (FBI), etc.) of an event related to the licensee's implementation of their cybersecurity program for digital computer and communication systems and networks within the scope of §§ 73.54 or 73.110 that does not otherwise require a notification under paragraph (a) of this section. (3) Within 8 hours after receipt or collection of information regarding observed behavior, activities, or statements that may indicate intelligence gathering or pre-operational planning related to a cyberattack against digital computer and communication systems and networks within the scope of §§ 73.54 or 73.110. (b) 
                            *Twenty-four hour recordable events.*
                             (1) The licensee shall use the site corrective action program to record vulnerabilities, weaknesses, failures and deficiencies in their § 73.54 or § 73.110 cybersecurity program within 24 hours of their discovery. 
 (c) * * * (6) 
                            *Declaration of emergencies.*
                             Notifications made to the NRC for the declaration of an emergency class shall be performed in accordance with §§ 50.72 or 53.1630 of this chapter, as applicable. (7) 
                            *Elimination of duplication.*
                             Separate notifications and reports are not required for events that are also reportable under §§ 50.72 and 50.73 or §§ 53.1630 and 53.1640 of this chapter. However, these notifications should also indicate the applicable § 73.77 reporting criteria.
 156.Add Subpart J consisting of §§ 73.100 through 73.120 to read as follows: Subpart J—Security Requirements at Commercial Nuclear Plants73.100
                            Technology-inclusive requirements for physical protection of licensed activities at commercial nuclear plants against radiological sabotage.
                            73.110
                            Technology-inclusive requirements for protection of digital computer and communication systems and networks.
                            73.120
                            Access authorization program for commercial nuclear plants.
 Subpart J—Security Requirements at Commercial Nuclear Plants § 73.100
  (a) 
                                *Introduction.*
                                 (1) Each licensee that is licensed to operate a commercial nuclear plant under 10 CFR part 53 and elects to implement the requirements of this section must do so through its physical security plan, training and qualification plan, safeguards 
                                ( print page 87120)
                                contingency plan, and cybersecurity plan, referred to collectively hereafter as “security plans,” before initial fuel load into the reactor (or, for a fueled manufactured reactor, before initiating the physical removal of any one of the independent physical mechanisms to prevent criticality required under § 53.620(d)(1) of this chapter). (2) The security plans must identify, describe, and account for site-specific conditions that affect the licensee's capability to satisfy the requirements of this section. (b) 
                                *General performance objective and requirements.*
                                 (1) The licensee must establish, implement, and maintain a physical protection program and a security organization, which will have as their objective to provide reasonable assurance that activities involving special nuclear material are not inimical to the common defense and security and do not constitute an unreasonable risk to the public health and safety. (2) To satisfy the general performance objective of paragraph (b)(1) of this section, the physical protection program must protect against the design basis threat of radiological sabotage as stated in § 73.1. Specifically, the licensee must— (i) Ensure that the physical protection program capabilities to protect against the design basis threat of radiological sabotage are maintained at all times; and (ii) Provide defense in depth in achieving performance requirements through the integration of engineered systems, administrative controls, and management measures. (3) The physical protection program must be designed and implemented to achieve and maintain the reliability and availability of structures, systems, and components (SSCs) required for demonstrating compliance with the following performance requirements at all times: (i) Intrusion detection. The licensee must be capable of detecting attempted and actual unauthorized access to interior and exterior areas containing SSCs needed to implement safety and security functions. (ii) Intrusion assessment. The licensee must be capable of timely assessment for determining the cause of a detected intrusion. (iii) Security communication. The licensee must be capable of continuous security communications. Communication systems must account for design basis threats that can interrupt or interfere with continuity or integrity of communications. (iv) Security response. The physical protection program must be designed to provide timely security response to interdict and neutralize adversary attacks up to and including the design basis threat of radiological sabotage. The physical protection program must be designed to provide layers of security response, with each layer assuring that a single failure does not result in the loss of capability to neutralize the design basis threat adversary. Structures, systems, and components relied on for delay functions must be designed to allow for timely security responses to adversary attacks with adequate defense in depth. (A) The security response may rely on the use of onsite responders, law enforcement or other offsite armed responders, or a combination thereof, to fulfill the interdiction and neutralization functions required by paragraph (b)(3)(iv) of this section. A licensee relying entirely or partially on law enforcement or other offsite armed responders must— (
                                *1*) Maintain the capability to detect, assess, interdict, and neutralize threats as required by paragraphs (b)(3)(i), (b)(3)(ii), and (b)(3)(iv) of this section; (
                                *2*) Provide adequate delay to enable law enforcement or other offsite armed responders to fulfill the interdiction and neutralization functions for threats up to and including the design basis threat of radiological sabotage; (
                                *3*) Provide necessary information about the facility and make available periodic training to law enforcement or other offsite armed responders who will fulfill the interdiction and neutralization functions for threats up to and including the design basis threat of radiological sabotage; (
                                *4*) Fully describe in the safeguards contingency plan the role that law enforcement or other offsite armed responders will play in the licensee's protective strategy. The description must provide sufficient detail to enable the NRC to determine that the licensee's physical protection program provides reasonable assurance of adequate protection against threats up to and including the design basis threat of radiological sabotage; and (
                                *5*) Identify criteria and measures to compensate for the degradation or absence of law enforcement or other offsite armed responders and propose suitable compensatory measures that meet the requirements of paragraph (h)(3) of this section to address this degradation. (B) For licensees relying entirely or partially on law enforcement responders to fulfill the interdiction and neutralization functions required by paragraph (b)(3)(iv) of this section, the training and qualification requirements related to armed response personnel in paragraphs (c) and (e) of this section do not apply to law enforcement responders. The licensee shall continue to satisfy the performance evaluation requirements in paragraph (g) of this section for all armed response personnel, including law enforcement. (v) Protecting against land and waterborne vehicle bomb assaults. The licensee must be capable of protecting the plant against the design basis threat vehicle bomb assault. The methods that are relied on to protect against a design basis threat land vehicle and waterborne vehicle bomb assault must be designed to protect the reactor building and structures containing safety- or security-related systems, and components from explosive effects. (vi) Access control portals. The licensee must be capable of detecting and denying unauthorized access to persons and pass-through of contraband materials (
                                *e.g.,*
                                 weapons, incendiaries, explosives) to protected areas. (4) The licensee must meet the requirements related to target sets in § 73.55(f). (5) The licensee must identify and analyze site-specific conditions, including target sets, that may affect the physical protection program needed to implement the requirements of this section. The licensee must account for these conditions in demonstrating compliance with the requirements of this section. (6) The licensee must establish, implement, and maintain a performance evaluation program to assess the effectiveness of the licensee's implementation of the physical protection program to protect against the design basis threat of radiological sabotage. (7) The licensee must establish, implement, and maintain an access authorization program under § 73.56 and must describe the program in the physical security plan. (8) The licensee must establish, implement, and maintain a cybersecurity program under §§ 73.54 or 73.110 and must describe the program in the cybersecurity plan. (9) The licensee must establish, implement, and maintain an insider mitigation program and must describe the program in the physical security plan. (i) The insider mitigation program must monitor the initial and continuing trustworthiness and reliability of individuals granted or retaining unescorted access or unescorted access authorization to a protected or vital area, and implement defense-in-depth methodologies to minimize the potential for an insider (active, passive, or both) 
                                ( print page 87121)
                                to adversely affect, either directly or indirectly, the licensee's capability to protect against radiological sabotage. (ii) The insider mitigation program must integrate elements of— (A) The access authorization program under § 73.56; (B) The fitness-for-duty program under 10 CFR part 26; (C) The cybersecurity program under §§ 73.54 or 73.110; and (D) The physical protection program under this section. (10) The licensee must have the capability to track, trend, correct, and prevent recurrence of failures and deficiencies in the implementation of the requirements of this section. (11) Implementation of security plans and associated procedures must be coordinated with other onsite plans and procedures to preclude conflict during both normal and emergency conditions and ensure the adequate management of the safety and security interface. (12)(i) The licensee must ensure that the firearms background check requirements of § 73.17 of this part are met for all members of the security organization whose official duties require access to covered weapons or who inventory enhanced weapons. (ii) The provisions of this paragraph are only applicable to licensees subject to this section that are also subject to the firearms background check provisions of § 73.17 of this part. (c) 
                                *Security organization.*
                                 The licensee must establish and maintain a security organization that is staffed, trained, qualified, and equipped to implement the physical protection program under the requirements of this section. (1) The licensee must establish a management system for maintaining and implementing security policies and procedures to implement the requirements of this section and the security plans. (2) Implementing procedures must document the conduct of security operations, security design and configuration controls, maintenance, training and qualification, and contingency responses. (3) The licensee must— (i) Establish a process for the approval of designs, policies, processes, and procedures and changes by the individual with overall responsibility for the physical protection program; and (ii) Ensure that revisions and changes to the physical protection program and implementing policies, processes, and procedures satisfy the requirements of this section. (4) The licensee must retain, in accordance with § 73.70, all analyses, assessments, calculations, and descriptions of the technical basis for demonstrating compliance with the performance requirements of § 73.100(b). The licensee must protect these records in accordance with the requirements for protecting safeguards information in §§ 73.21 and 73.22. (5) The licensee may not permit any individual to implement any part of the physical protection program unless the individual has been trained, equipped, and qualified to perform their assigned duties and responsibilities in accordance with the training and qualification plan. (d) 
                                *Search requirements.*
                                 The licensee must establish and implement searches of individuals, vehicles, and materials to detect and prevent the introduction into the protected area of firearms, explosives, incendiary devices, or other items and material which could be used to commit radiological sabotage. (e) 
                                *Training and qualification program.*
                                 The licensee must establish and maintain a training and qualification program that ensures personnel who are responsible for the physical protection of the facility against radiological sabotage are able to effectively perform their assigned security-related job duties for implementing the requirements of this section and must describe the program in the training and qualification plan. (f) 
                                *Security reviews.*
                                 The licensee must establish and implement security reviews to assess the effectiveness of the implementation of the physical protection program. Security reviews must be performed by individuals independent of those personnel responsible for program management and any individual who has direct responsibility for implementing the onsite physical protection program. (1) The licensee must review each element of the physical protection program at a frequency commensurate with the importance or significance to safety of plant operations to ensure timely identification and documentation of vulnerabilities, improvements, and corrective actions. The objective of these reviews must be maintaining effective implementation of the engineered and administrative controls required to achieve the physical protection program functions and the management system required to implement programs and requirements in this section. (2) The licensee must establish and perform self-assessments to ensure the effective implementation of the physical protection program functions of detection, assessment, communication, delay, and interdiction and neutralization to protect against the design basis threat of radiological sabotage. The licensee must perform design verification and assessments of the capabilities of active and passive engineering systems relied on to protect against the design basis threat. (3) Reviews of the security program must include, but are not limited to, an audit of the effectiveness of the physical protection program, security plans, implementing procedures, cybersecurity programs, safety/security interface activities, the testing, maintenance, and calibration program, and response commitments by local, State, and Federal law enforcement authorities. (4) The results and recommendations of the onsite physical protection program reviews, management's findings regarding program effectiveness, and any actions taken as a result of recommendations from prior program reviews, must be documented in a report and must be maintained in an auditable form and available for inspection. (g) 
                                *Performance evaluation.*
                                 Licensee performance evaluations must include methods appropriate and necessary to assess, test, and challenge the integration of the physical protection program's functions to protect against the design basis threat, including measures to protect against cyberattack and engineered systems designed to protect against the design basis threat standalone ground vehicle bomb attack. (1) The licensee must establish the frequencies for performance evaluations commensurate with the security significance of the physical protection program. (2) The licensee must document processes and procedures for implementing the performance evaluations. The licensee must maintain records, including results, findings, and corrective actions identified during the performance evaluations. (h) 
                                *Maintenance, testing, and calibration and corrective actions.*
                                 (1) The licensee must ensure that security SSCs, including supporting systems, are inspected, tested, and calibrated for operability and performance at intervals necessary and sufficient to meet the requirements of this section. (2) The licensee must implement corrective actions to ensure resolution of identified vulnerabilities and deficiencies to meet the requirements of this section. (3) The licensee must establish and implement timely compensatory measures for degraded or inoperable security SSCs to meet the requirements of this section. Compensatory measures must provide a level of protection that is equivalent to the protection that was provided prior to the degradation or 
                                ( print page 87122)
                                inoperability of the security structures, systems, or components. (4) The licensee must document processes and procedures and maintain records for implementing the corrective actions, compensatory measures, and maintenance, inspection, testing, and calibration of security SSCs. (i) 
                                *Suspension of security measures.*
                                 (1) The licensee may suspend implementation of affected requirements of this section in accordance with § 53.740(h) of this chapter under the following conditions: (i) In an emergency, when action is immediately needed to protect the public health and safety; and (ii) During severe weather, when the suspension of affected security measures is immediately needed to protect the personal health and safety of personnel. (2) Suspended security measures must be reinstated as soon as conditions permit. (3) The suspension of security measures must be reported and documented in accordance with the provisions of §§ 73.1200 and 73.1205. (j) 
                                *Records.*
                                 (1) The Commission may inspect, copy, retain, and remove all reports, records, and documents required to be kept by Commission regulations, orders, or license conditions, whether the reports, records, and documents are kept by the licensee or a contractor. (2) The licensee must maintain all records required to be kept by Commission regulations, orders, or license conditions, until the Commission terminates the license for which the records were developed and must maintain superseded portions of these records for at least 3 years after the record is superseded, unless otherwise specified by the Commission. (3) If a contracted security force is used to implement the onsite physical protection program, the licensee's written agreement with the contractor must be retained by the licensee as a record for the duration of the contract. (4) Review and audit reports must be available for inspection, for a period of 3 years.
 § 73.110
  (a) Each licensee that is licensed to operate a commercial nuclear plant under 10 CFR part 53 and elects to implement the requirements of this section must establish, implement, and maintain a cybersecurity program that is commensurate with the potential consequences resulting from cyberattacks, up to and including the design basis threat as described in § 73.1. The cybersecurity program must provide reasonable assurance that digital computer and communication systems and networks are adequately protected against cyberattacks that are capable of causing the following consequences: (1) Adversely impacting the functions performed by digital assets that would prevent a postulated fission product release resulting in offsite doses exceeding the values in § 53.210 of this chapter. (2) Adversely impacting the functions performed by digital assets used by the licensee for implementing the physical security requirements in § 53.860(a) of this chapter. (b) To protect digital computer and communication systems and networks associated with the functions described in paragraphs (a)(1) and (2), the licensee must— (1) Analyze the potential consequences resulting from cyberattacks on digital computer and communication systems and networks and identify those assets that must be protected to demonstrate compliance with paragraph (a) of this section; and (2) Implement the cybersecurity program in accordance with paragraph (d) of this section. (c) The licensee must comply with the requirements in § 73.54(a)(2) for the systems and networks identified in paragraph (b)(1) of this section in a manner that is commensurate with the potential consequences resulting from cyberattacks. (d) The cybersecurity program must be designed in a manner that is commensurate with the potential consequences resulting from cyberattacks through the following steps: (1) Implement security controls to protect the assets identified under paragraph (b)(1) of this section from cyberattacks, commensurate with their safety and security significance; (2) Apply and maintain defense-in-depth protective strategies to ensure the capability to detect, delay, respond to, and recover from cyberattacks capable of causing the consequences identified in paragraph (a) of this section; (3) Mitigate the adverse effects of cyberattacks capable of causing the consequences identified in paragraph (a) of this section; and (4) Ensure that the functions of protected assets identified under paragraph (b)(1) of this section are not adversely impacted due to cyberattacks. (e) The licensee must implement the following requirements in a manner that is commensurate with the potential consequences resulting from cyberattacks: (1) As part of the cybersecurity program, the licensee must comply with the requirements in § 73.54(d)(1), (2), and (4), and must ensure that modifications to assets, identified under paragraph (b)(1) of this section are evaluated before implementation to ensure that the cybersecurity performance objectives identified in paragraph (a) of this section are maintained. (2) The licensee must establish, implement, and maintain a cybersecurity plan that implements the cybersecurity program requirements of this section. (i) The cybersecurity plan must describe how the requirements of this section will be implemented and must account for the site-specific conditions that affect implementation. (ii) The cybersecurity plan must include measures for incident response and recovery for cyberattacks. The cybersecurity plan must include the analysis identified under paragraph (b)(1) of this section and describe how the licensee will— (A) Apply and maintain defense-in-depth protective strategies as required in paragraph (d)(2) of this section; (B) Maintain the capability for timely detection and response to cyberattacks; (C) Mitigate the consequences of cyberattacks; (D) Correct exploited vulnerabilities; and (E) Restore affected systems, networks, and/or equipment affected by cyberattacks. (3) The licensee must develop and maintain written policies and implementing procedures to implement the cybersecurity plan. Policies, implementing procedures, and other supporting technical information used by the licensee need not be submitted for Commission review and approval as part of the cybersecurity plan but are subject to inspection by NRC staff on a periodic basis. (4) The licensee must establish and implement cybersecurity reviews to assess the effectiveness of the implementation of the cybersecurity program. (i) The licensee must review each element of the cybersecurity program at a frequency commensurate with the importance or significance to safety of plant operations to ensure timely identification and documentation of vulnerabilities, improvements, and corrective actions. (ii) Cybersecurity reviews must be performed by individuals independent 
                                ( print page 87123)
                                of those personnel responsible for program management and any individual who has direct responsibility for implementing the cybersecurity program. (iii) The licensee must establish and perform self-assessments to ensure the effective implementation of the cybersecurity program. (iv) The results and recommendations of the cybersecurity program reviews, management's findings regarding program effectiveness, and any actions taken as a result of recommendations from prior program reviews, must be documented in a report and must be maintained in an auditable form and available for inspection. (5) The licensee must retain all records and supporting technical documentation required to demonstrate compliance with the requirements of this section as a record until the Commission terminates the license for which the records were developed and must maintain superseded portions of these records for at least three (3) years after the record is superseded, unless otherwise specified by the Commission.
 § 73.120
  (a) 
                                *Introduction and scope.*
                                 Each applicant for an operating license or a holder of a combined license under 10 CFR part 53 must establish, maintain, and implement an access authorization program before initial fuel load into the reactor (or, for a fueled manufactured reactor, before initiating the physical removal of any one of the independent physical mechanisms to prevent criticality required under § 53.620(d)(1) of this chapter). The requirements in this section apply to licensees satisfying the criterion in § 53.860(a)(2)(i) of this chapter. (b) 
                                *Applicability.*
                                 (1) The following individuals must be subject to an access authorization program under this section: (i) Any individual to whom a licensee intends to grant unescorted access to a commercial nuclear plant protected area, vital area, or controlled access area where licensed material is used or stored; (ii) Any individual whose duties and responsibilities permit the individual to take actions by electronic means, either on site or remotely, that could adversely impact the licensee's or applicant's operational safety, security, or emergency preparedness; (iii) Any individual who has responsibilities for implementing a licensee's or applicant's protective strategy, including armed security force officers, alarm station operators, and tactical response team leaders but not including Federal, State, or local law enforcement personnel; and (iv) The licensee or applicant access authorization program reviewing official or contractor or vendor access authorization program reviewers. (2) The licensee or applicant may subject other individuals, including employees of a contractor or a vendor who are designated in access authorization program procedures, to an access authorization program that demonstrates compliance with the requirements of this section. (c) 
                                *General performance objectives and requirements.*
                                 Each licensee's or applicant's access authorization program under this section must demonstrate that the individuals who are specified in paragraph (b) of this section are trustworthy and reliable, such that they do not constitute an unreasonable risk to public health and safety or the common defense and security. The licensee's access authorization program must maintain the capabilities for demonstrating compliance with the following performance requirements: (1) 
                                *Background investigation.*
                                 (i)(A) Licensees and applicants must ensure that any individual seeking initial unescorted access or to maintain unescorted access is subject to a background investigation. (B) Background investigations must include the program elements contained under § 37.25 of this chapter and must also include a credit history evaluation. (ii) Background investigations must include fingerprinting and an FBI identification and criminal history records check in accordance with § 37.27 of this chapter. (iii) Licensees must have the informed and signed consent of the subject individual to initiate a background investigation. This consent must include authorization to share personal information with other individuals or organizations as necessary to complete the background investigation. A signed consent must be obtained prior to any reinvestigation. The subject individual may withdraw his or her consent at any time. Licensees must inform the individual that— (A) If an individual withdraws his or her consent, the licensee may not initiate any elements of the background investigation that were not in progress at the time the individual withdrew his or her consent; and (B) The withdrawal of consent for the background investigation is sufficient cause for denial or termination of unescorted access authorization. (2) 
                                *Behavioral observation.*
                                 Licensees, applicants, contractors, and vendors must ensure the access authorization program includes provisions that the individuals specified in paragraph (b) of this section are subject to behavioral observation. (i) Each person subject to behavioral observation must communicate to the licensee or applicant observed behaviors or activities of individuals that may constitute an unreasonable risk to the health and safety of the public and common defense and security. (ii) Behavioral observation must include visual observation, in person or remotely by video, to detect and promptly report to plant supervision any concerns arising from behavioral observation, including, but not limited to, concerns related to any questionable behavior patterns or activities of others. (3) 
                                *Self-reporting of legal actions.*
                                 Licensees or applicants must inform personnel who are granted and who maintain unescorted access of their responsibilities to self-report to plant supervision legal actions taken by a law enforcement authority or court of law against the individual that could result in incarceration or a court order or that requires a court appearance, including but not limited to an arrest, an indictment, the filing of charges, or a conviction, but excluding minor civil actions or misdemeanors such as parking violations or speeding tickets, for any individual who has applied for unescorted access or who maintains unescorted access. (4) 
                                *Unescorted access.*
                                 Licensees or applicants must grant unescorted access only after the licensee has verified an individual is trustworthy and reliable. A list of persons currently approved for unescorted access to a protected area, vital area, or controlled access area must be maintained at all times. Unescorted access determinations must be reviewed annually by the reviewing official. Licensees and applicants must complete an FBI criminal history record check update for each individual maintaining unescorted access, within 10 years of the last review. (5) 
                                *Termination of unescorted access.*
                                 Licensees and applicants must promptly terminate unescorted access when this access is no longer required or a reviewing official determines an individual is no longer trustworthy and reliable in accordance with this section. (6) 
                                *Determination basis for access.*
                                 (i) The licensee's or applicant's reviewing official must determine whether to permit, deny, unfavorably terminate, maintain, or administratively withdraw an individual's unescorted access based on an evaluation of all of the 
                                ( print page 87124)
                                information collected to demonstrate compliance with the requirements of this section. (ii) Licensees and applicants must provide individuals subject to this section, prior to any final adverse determination, the right to complete, correct, and explain information obtained as a result of the licensee's background investigation pursuant to § 37.23(g) of this chapter. (iii) The licensee's or applicant's reviewing officials are the only individuals authorized to make unescorted access determination decisions. Each licensee or applicant must name one or more individuals to be reviewing officials pursuant to the requirements of § 37.23(b)(2) of this chapter. (7) 
                                *Review procedures.*
                                 Review procedures must be established in accordance with § 37.23(f) of this chapter, to include provisions for the notification in writing of individuals who are denied unescorted access or who are unfavorably terminated. (8) 
                                *Protection of information.*
                                 Licensees, applicants, contractors, or vendors must establish and maintain a system of files and procedures in accordance with § 37.31 of this chapter, to ensure personal information is not disclosed to unauthorized persons. (9) 
                                *Access authorization reviews and corrective action.*
                                 Licensees and applicants must develop, implement, and maintain procedures for conduct of access authorization reviews and corrective actions in accordance with § 37.33 of this chapter to ensure the continuing effectiveness of the access authorization program and to ensure that the access authorization program and program elements are in compliance with the requirements of this section. Each licensee and applicant must be responsible for the continuing effectiveness of the access authorization program, including access authorization program elements that are provided by the contractors or vendors, and the access authorization programs of any of the contractors or vendors that are accepted by the licensee or applicant. (10) 
                                *Records.*
                                 Licensees, applicants, and contractors or vendors must document the processes and procedures for maintaining records used or created to establish an individual's trustworthiness and reliability or to document access determinations. Licensees, applicants, and contractor or vendors must— (i) Retain documentation regarding the trustworthiness and reliability of individual employees for 3 years from the date the individual no longer requires unescorted access; (ii) Retain a copy of the current access authorization program procedures as a record for 3 years after the procedure is no longer needed. If any portion of the procedure is superseded, retain the superseded material for 3 years after the record is superseded; and (iii) Retain the list of persons approved for unescorted access for 3 years after the list is superseded or replaced. Records maintained in any database(s) must be available for NRC review.
 157.In § 73.1200, revise paragraphs (a) introductory text, (c)(1) introductory text, (e)(1) introductory text, (e)(3) and (4), (g)(1) introductory text, (o)(5)(i) and (o)(6)(i), (r) and (s) to read as follows: § 73.1200
  (a) 
                            *15-minute notifications—facilities.*
                             Each licensee subject to the provisions of § 73.20, § 73.45, § 73.46, § 73.51, § 73.55, or § 73.100 must notify the NRC Headquarters Operations Center, as soon as possible but within 15 minutes after— 
 (c) * * * (1) Each licensee subject to the provisions of §§ 73.20, 73.45, 73.46, 73.50, 73.51, 73.55, 73.60, 73.67, or 73.100 must notify the NRC Headquarters Operations Center as soon as possible but no later than 1 hour after the time of discovery of the following significant facility security events involving— 
 (e) * * * (1) Each licensee subject to the provisions of §§ 73.20, 73.45, 73.46, 73.50, 73.51, 73.55, 73.60, 73.67, or 73.100 must notify the NRC Headquarters Operations Center within 4 hours after time of discovery of the following facility security events involving— 
 (3)(i) An event involving a law enforcement response to the facility that could reasonably be expected to result in public or media inquiries and that does not otherwise require a notification under paragraphs (a) through (h) of this section, or in other NRC regulations such as § 50.72(b)(2)(xi) or § 53.1630(b)(2)(v) of this chapter. (ii) As an exemption, licensees need not report law enforcement responses to minor incidents, such as traffic accidents. (4) For licensees subject to the provisions of § 73.55 or § 73.100 of this part, an event involving the licensee's suspension of security measures. 
 (g) * * * (1) Each licensee subject to the provisions of § 73.20, § 73.45, § 73.46, § 73.50, § 73.51, § 73.55, § 73.60, § 73.67, or § 73.100 must notify the NRC Headquarters Operations Center within 8 hours after time of discovery of the following facility security program failures involving— 
 (o) * * * (5) * * * (i) Licensees must establish the requested continuous communications channel once the licensee has completed other required notifications under this section, § 50.72 of this chapter, appendix E to part 50 of this chapter, § 53.1630 of this chapter, § 70.50 of this chapter; or § 72.75 of this chapter; as appropriate. 
 (6) * * * (i) Licensees must establish the requested continuous communications channel once the licensee or the movement control center has completed other required notifications under this section, § 50.72 of this chapter, appendix E to part 50 of this chapter, § 53.1630 of this chapter, § 70.50 of this chapter; § 72.75 of this chapter; or requested assistance from the LLEA, as appropriate. 
 (r) 
                            *Declaration of emergencies.*
                             Licensees notifying the NRC of the declaration of an emergency class must do so in accordance with §§ 50.72, 53.1630, 63.73, 70.50, and 72.75 of this chapter, as applicable. (s) 
                            *Elimination of duplication.*
                             Licensees with notification obligations under paragraphs (a) through (h), (m), and (n) of this section and §§ 50.72, 53.1630, 63.73, 70.50, and 72.75 of this chapter may notify the NRC of events in a single communication. This communication must identify each regulation under which the licensee is reporting.
 158.In § 73.1205, revise paragraph (b)(2) to read as follows: § 73.1205
  
 (b) * * * (2)(i) Licensees subject to § 50.73 or § 53.1640 of this chapter must prepare the written follow-up report on NRC Form 366. (ii) Licensees not subject to § 50.73 or § 53.1640 of this chapter must prepare the written follow-up report in a letter format.
 159.In § 73.1210, revise paragraphs (a)(1) and (b)(3)(i) to read as follows: § 73.1210
  (a) * * * (1) Licensees with facilities or shipment activities subject to the provisions of § 73.20, § 73.25, § 73.26, § 73.27, § 73.37, § 73.45, § 73.46, § 73.50, § 73.51, § 73.55, § 73.60, § 73.67, or § 73.100, must record the physical security events and conditions adverse to security that are specified in paragraphs (c) through (f) of this section. 
 (b) * * * (3)(i) Licensees must record these physical security events and conditions adverse to security in either a stand-alone safeguards event log or as part of the licensee's corrective action program, as specified under the applicable quality assurance program provisions of parts 50, 52, 53, 60, 63, 70, and 72 of this chapter, or both.
 160.In § 73.1215, revise paragraph (d)(1) introductory text to read as follows: § 73.1215
  
 (d) * * * (1) For licensees subject to the provisions of §§ 73.20, 73.45, 73.46, 73.50, 73.51, 73.55, 73.60, 73.67, or 73.100, the licensees must report activities they assess are suspicious. Examples include, but are not limited to, the following:
 161.In appendix B to part 73, revise *Definitions*
                         introductory text to read as follows: Appendix B to Part 73—General Criteria for Security Personnel 
 Definitions Terms defined in parts 50, 53, 70, and 73 of this chapter have the same meaning when used in this appendix. 
 PART 74—MATERIAL CONTROL AND ACCOUNTING OF SPECIAL NUCLEAR MATERIAL
 162.The authority citation for 10 CFR part 74 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 53, 57, 161, 182, 223, 234, 1701 (42 U.S.C. 2073, 2077, 2201, 2232, 2273, 2282, 2297f); Energy Reorganization Act of 1974, secs. 201, 202 (42 U.S.C. 5841, 5842); 44 U.S.C. 3504 note. 163.In § 74.31, revise paragraph (a) introductory text to read as follows: § 74.31
  (a) 
                            *General performance objectives.*
                             Each licensee who is authorized to possess and use more than one effective kilogram of special nuclear material of low strategic significance, excluding sealed sources, at any site or contiguous sites subject to control by the licensee, other than a production or utilization facility licensed pursuant to parts 50, 53, or 70 of this chapter, or operations involved in waste disposal, shall implement and maintain a Commission approved material control and accounting system that will achieve the following objectives:
 164.In § 74.41, revise paragraph (a) introductory text to read as follows: § 74.41
  (a) 
                            *General performance objectives.*
                             Each licensee who is authorized to possess special nuclear material (SNM) of moderate strategic significance or SNM in a quantity exceeding one effective kilogram of strategic special nuclear material in irradiated fuel reprocessing operations other than as sealed sources and to use this material at any site other than a nuclear reactor licensed pursuant to parts 50 or 53 of this chapter; or as reactor irradiated fuels involved in research, development, and evaluation programs in facilities other than irradiated fuel reprocessing plants; or an operation involved with waste disposal, shall establish, implement, and maintain a Commission-approved material control and accounting (MC&A) system that will achieve the following performance objectives:
 165.In § 74.51, revise paragraph (a) introductory text to read as follows: § 74.51
  (a) 
                            *General performance objectives.*
                             Each licensee who is authorized to possess five or more formula kilograms of strategic special nuclear material (SSNM) and to use such material at any site, other than a nuclear reactor licensed pursuant to parts 50 or 53 of this chapter, an irradiated fuel reprocessing plant, an operation involved with waste disposal, or an independent spent fuel storage facility licensed pursuant to part 72 of this chapter shall establish, implement, and maintain a Commission-approved material control and accounting (MC&A) system that will achieve the following objectives:
 PART 75—SAFEGUARDS ON NUCLEAR MATERIAL—IMPLEMENTATION OF SAFEGUARDS AGREEMENTS BETWEEN THE UNITED STATES AND THE INTERNATIONAL ATOMIC ENERGY AGENCY
 166.The authority citation for 10 CFR part 75 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 53, 63, 103, 104, 122, 161, 223, 234, 1701 (42 U.S.C. 2073, 2093, 2133, 2134, 2152, 2201, 2273, 2282, 2297f); Energy Reorganization Act of 1974, sec. 201 (42 U.S.C. 5841); Nuclear Waste Policy Act of 1982, secs. 135, 141 (42 U.S.C. 10155, 10161); 44 U.S.C. 3504 note. 167.In § 75.4, revise the introductory text and paragraph (6) of the definition for “Facility” to read as follows: § 75.4
  
 Unless otherwise defined in this section, the terms defined in §§ 40.4, 50.2, 53.020, and 70.4 of this chapter have the same meaning when used in this part. 
 *Facility*
                             means: 
 (6) Any plant or location where the possession of more than 1 effective kilogram of nuclear material is licensed pursuant to 10 CFR part 40, 50, 53, 60, 61, 63, 70, 72, 76, or 150 of this chapter or an Agreement State license.
 PART 95—FACILITY SECURITY CLEARANCE AND SAFEGUARDING OF NATIONAL SECURITY INFORMATION AND RESTRICTED DATA
 168.The authority citation for 10 CFR part 95 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 145, 161, 223, 234 (42 U.S.C. 2165, 2201, 2273, 2282); Energy Reorganization Act of 1974, sec. 201 (42 U.S.C. 5841); 44 U.S.C. 3504 note; E.O. 10865, as amended, 25 FR 1583, 3 CFR, 1959-1963 Comp., p. 398; E.O. 12829, 58 FR 3479, 3 CFR, 1993 Comp., p. 570; E.O. 12968, 60 FR 40245, 3 CFR, 1995 Comp., p. 391; E.O. 13526, 75 FR 707, 3 CFR, 2009 Comp., p. 298. 169.In § 95.5, revise the definition for “ *License*
                        ” to read as follows: § 95.5
  
 *License*
                             means a license issued under 10 CFR part 50, 52, 53, 54, 60, 63, 70, or 72.
 § 95.39
 170.In § 95.39(a), remove “part 52” and add in its place “parts 52 or 53.” PART 140—FINANCIAL PROTECTION REQUIREMENTS AND INDEMNITY AGREEMENTS
 171.The authority citation for 10 CFR part 140 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 161, 170, 223, 234 (42 U.S.C. 2201, 2210, 2273, 2282); Energy Reorganization Act of 1974, secs. 201, 202 (42 U.S.C. 5841, 5842); 44 U.S.C. 3504 note. 172.In § 140.2, revise paragraphs (a)(1) and (2) to read as follows: § 140.2
  (a) * * * (1) To each person who is an applicant for or holder of a license issued under 10 CFR part 50, 52, 53, or 54 to operate a nuclear reactor, and (2) With respect to an extraordinary nuclear occurrence, to each person who is an applicant for or holder of a license to operate a production facility or a utilization facility (including an operating license issued under part 50 or part 53 of this chapter and a combined license under part 52 or part 53 of this chapter), and to other persons indemnified with respect to the involved facilities.
 173.Revise § 140.10 to read as follows: § 140.10
  This subpart applies to each person who is an applicant for or holder of a license issued under 10 CFR parts 50, 53 or 54 to operate a nuclear reactor, or is the applicant for or holder of a combined license issued under 10 CFR parts 52, 53, or 54, except licenses held by persons found by the Commission to be Federal agencies or nonprofit educational institutions licensed to conduct educational activities. This subpart also applies to persons licensed to possess and use plutonium in a plutonium processing and fuel fabrication plant.
 174.In § 140.11, revise paragraph (b) to read as follows: § 140.11
  
 (b) In any case where a person is authorized under 10 CFR parts 50, 52, 53, or 54 to operate two or more nuclear reactors at the same location, the total primary financial protection required of the licensee for all such reactors is the highest amount which would otherwise be required for any one of those reactors; provided, that such primary financial protection covers all reactors at the location.
 175.In § 140.12, revise paragraph (c) to read as follows: § 140.12
  
 (c) In any case where a person is authorized under 10 CFR parts 50, 52, 53, or 54 to operate two or more nuclear reactors at the same location, the total financial protection required of the licensee for all such reactors is the highest amount which would otherwise be required for any one of those reactors; provided, that such financial protection covers all reactors at the location.
 176.Revise § 140.13 to read as follows: § 140.13
 10 CFR part 52
 Each holder of a 10 CFR part 50 or 10 CFR part 53 construction permit, or a holder of a combined license under parts 52 or 53 of this chapter before the date that the Commission had made the finding under §§ 52.103(g) or 53.1452(g) of this chapter, who also holds a license under part 70 of this chapter authorizing ownership, possession and storage only of special nuclear material at the site of the nuclear reactor for use as fuel in operation of the nuclear reactor after issuance of either an operating license under 10 CFR part 50 or 53, or a combined license under 10 CFR part 52 or 53, shall, during the period before issuance of a license authorizing operation under 10 CFR part 50 or 53, or the period before the Commission makes the finding under § 52.103(g) or § 53.1452(g) of this chapter, as applicable, have and maintain financial protection in the amount of $1,000,000. Proof of financial protection shall be filed with the Commission in the manner specified in § 140.15 before issuance of the license under part 70 of this chapter.
 177.In § 140.20, revise paragraphs (a)(1)(i) and (ii) to read as follows: § 140.20
  (a) * * * (1)(i) The effective date of the license (issued under part 50 or part 53 of this chapter) authorizing the licensee to operate the nuclear reactor involved; or (ii) The date that the Commission makes the finding under §§ 52.103(g) or 53.1452(g) of this chapter; or
 PART 150—EXEMPTIONS AND CONTINUED REGULATORY AUTHORITY IN AGREEMENT STATES AND IN OFFSHORE WATERS UNDER SECTION 274
 178.The authority citation for 10 CFR part 150 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 11, 53, 81, 83, 84, 122, 161, 181, 223, 234, 274 (42 U.S.C. 2014, 2201, 2231, 2273, 2282, 2021); Energy Reorganization Act of 1974, sec. 201 (42 U.S.C. 5841); Nuclear Waste Policy Act of 1982, secs. 135, 141 (42 U.S.C. 10155, 10161); 44 U.S.C. 3504 note. 179.In § 150.15, revise paragraphs (a)(7)(iii) and (a)(8) to read as follows: § 150.15
  (a) * * * (7) * * * (iii) Greater than Class C (GTCC) waste, as defined in part 72 of this chapter, in an ISFSI or an MRS licensed under part 72 of this chapter; the GTCC waste must originate in, or be used by, a facility licensed under parts 50, 52, or 53 of this chapter. (8) Greater than Class C waste, as defined in part 72 of this chapter, that originates in, or is used by, a facility licensed under parts 50, 52, or 53 of this chapter and is licensed under part 30 and/or part 70 of this chapter.
 PART 170—FEES FOR FACILITIES, MATERIALS, IMPORT AND EXPORT LICENSES, AND OTHER REGULATORY SERVICES UNDER THE ATOMIC ENERGY ACT OF 1954, AS AMENDED
 180.The authority citation for 10 CFR part 170 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 11, 161(w) (42 U.S.C. 2014, 2201(w)); Energy Reorganization Act of 1974, sec. 201 (42 U.S.C. 5841); 42 U.S.C. 2215; 31 U.S.C. 901, 902, 9701; 44 U.S.C. 3504 note. 181.In § 170.3, revise the definitions for “Manufacturing License,” “Part 55 Reviews,” “Power reactor,” and “Special projects” to read as follows: § 170.3
  
 *Manufacturing license*
                             means a license under subpart F of part 52 of this chapter or subpart H of part 53 of this chapter to manufacture a nuclear power reactor(s) to be operated at sites not identified in the license application. 
 *Part 55 Reviews*
                             as used in this part means those services provided by the Commission to administer requalification and replacement examinations and tests for reactor 
                            ( print page 87127)
                            operators licensed under 10 CFR part 55 or 53 of the Commission's regulations and employed by part 50 or 53 licensees. These services also include related items such as the preparation, review, and grading of the examinations and tests. 
 *Power reactor*
                             means a nuclear reactor designed to produce electrical or heat energy licensed by the Commission under the authority of section 103 or subsection 104b of the Act, and under the provisions of §§ 50.21(b), 50.22, or part 53 of this chapter. 
 *Special projects*
                             means specific services provided by the Commission for which fees are not otherwise specified in this chapter. This includes, but is not limited to, contested hearings on licensing actions directly related to U.S. Government national security initiatives (as determined by the NRC), topical report reviews, early site reviews, waste solidification activities, activities related to the tracking and monitoring of shipment of classified matter, services provided to certify licensee, vendor, or other private industry personnel as instructors for 10 CFR part 55 or 53 reactor operators, reviews of financial assurance submittals that do not require a license amendment, reviews of responses to Confirmatory Action Letters, reviews of uranium recovery licensees' land-use survey reports, and reviews of §§ 50.71 or 53.1545 of this chapter Final Safety Analysis Reports. Special projects does not include activities otherwise exempt from fees under this part. It also does not include those contested hearings for which a fee exemption is granted in § 170.11(a)(2), including those related to individual plant security modifications.
 182.In § 170.12, revise paragraph (d)(1)(v) to read as follows: § 170.12
  
 (d) * * * (1) * * * (v) 10 CFR 50.71 or 53.1545 final safety analysis reports;
 § 170.21
 183.In § 170.21, in footnote 1 remove the phrase “( *e.g.,*10 CFR 50.12, 10 CFR 73.5)” and add in its place the phrase “(
                        *e.g.,*10 CFR 50.12, 10 CFR 53.080, 10 CFR 73.5)”. 184.Revise § 170.41to read as follows: § 170.41
  If the Commission determines that an applicant or a licensee has failed to pay a prescribed fee required in this part, the Commission will not process any application and may suspend or revoke any license or approval issued to the applicant or licensee. The Commission may issue an order with respect to licensed activities that the Commission determines to be appropriate or necessary to carry out the provisions of this part, parts 30, 31, 32 through 35, 40, 50, 53, 61, 70, 71, 72, 73, and 76 of this chapter, and of the Act.
 PART 171—ANNUAL FEES FOR REACTOR LICENSES AND FUEL CYCLE LICENSES AND MATERIALS LICENSES, INCLUDING HOLDERS OF CERTIFICATES OF COMPLIANCE, REGISTRATIONS, AND QUALITY ASSURANCE PROGRAM APPROVALS AND GOVERNMENT AGENCIES LICENSED BY THE NRC
 185.The authority citation for 10 CFR part 171 continues to read as follows: Authority:
                        Atomic Energy Act of 1954, secs. 11, 161(w), 223, 234 (42 U.S.C. 2014, 2201(w), 2273, 2282); Energy Reorganization Act of 1974, sec. 201 (42 U.S.C. 5841); 42 U.S.C. 2215; 44 U.S.C. 3504 note. 186.Revise § 171.3 to read as follows: § 171.3
  The regulations in this part apply to any person holding an operating license for a test reactor or research reactor issued under part 50 of this chapter, and to any person holding an operating license for a power reactor licensed under 10 CFR part 50 or 53, or a combined license issued under 10 CFR part 52 or 53, that has provided notification to the U.S. Nuclear Regulatory Commission (NRC) that the licensee has successfully completed power ascension testing. The regulations in this part also apply to any person holding a materials license as defined in this part, a Certificate of Compliance, a sealed source or device registration, a quality assurance program approval, and to a Government agency as defined in this part. Notwithstanding the other provisions in this section, the regulations in this part do not apply to uranium recovery and fuel facility licensees until after the Commission verifies through inspection that the facility has been constructed in accordance with the requirements of the license.
 187.In § 171.5, revise the definitions for “Operating license,” and “Power reactor” to read as follows: § 171.5
  
 *Operating license*
                             means having a license issued under §§ 50.57 or 53.1387 of this chapter. It does not include licenses that only authorize possession of special nuclear material after the Commission has received a request from the licensee to amend its licensee to permanently withdraw its authority to operate or the Commission has permanently revoked such authority. 
 *Power reactor*
                             means a nuclear reactor designed to produce electrical or heat energy and licensed by the Commission under the authority of section 103 or subsection 104b of the Atomic Energy Act of 1954, as amended, and under the provisions of §§ 50.21(b) or 50.22, or part 53 of this chapter.
 188.In § 171.15, revise paragraphs (a), (b)(2)(iii), (c)(1), and (d)(1) to read as follows: § 171.15
  (a) Each person holding an operating license for one or more non-power production or utilization facilities under 10 CFR part 50 that has provided notification to the NRC of the successful completion of startup testing; each person holding an operating license for a power reactor licensed under 10 CFR part 50 or a combined license under 10 CFR part 52, or an operating license or combined license for a commercial nuclear plant under 10 CFR part 53, that has provided notification to the NRC of the successful completion of power ascension testing; each person holding a 10 CFR part 50 or 52, power reactor license, or a 10 CFR part 53 commercial nuclear plant license that is in decommissioning or possession only status, except those that have no spent fuel onsite; and each person holding a 10 CFR part 72 license who does not hold a 10 CFR part 50, 52, or 53 license and provides notification under § 72.80(g) of this chapter, shall pay the annual fee for each license held during the Federal fiscal year in which the fee is due. This paragraph (a) does not apply to test or research reactors exempted under § 171.11(b). (b) * * * (2) * * * (iii) Generic activities required largely for NRC to regulate power reactors (
                            *e.g.,*
                             updating part 50, part 52, or part 53 of this chapter, operating the Incident Response Center, new reactor regulatory infrastructure). The base annual fee for operating power reactors does not 
                            ( print page 87128)
                            include generic activities specifically related to reactor decommissioning. (c)(1) The FY 2022 annual fee for each power reactor holding a 10 CFR part 50 operating license or combined license issued under 10 CFR part 52 or part 53 that is in a decommissioning or possession-only status and has spent fuel onsite, and for each independent spent fuel storage 10 CFR part 72 licensee who does not hold a 10 CFR part 50 or part 53 operating license, or a 10 CFR part 52 or part 53 combined license, is $227,000. 
 (d)(1) Each person holding an operating license for an SMR issued under 10 CFR part 50 or part 53, or a combined license issued under 10 CFR part 52 or part 53, that has provided notification to the NRC of the successful completion startup testing, shall pay the annual fee for all licenses held for an SMR site. The annual fee will be determined using the cumulative licensed thermal power rating of all SMR units and the bundled unit concept, during the fiscal year in which the fee is due. For a given site, the use of the bundled unit concept is independent of the number of SMR plants, the number of SMR licenses issued, or the sequencing of the SMR licenses that have been issued.
 189.In § 171.17, revise paragraphs (a) introductory text, (a)(1)(ii), and (a)(2) to read as follows: § 171.17
  (a) Reactors, 10 CFR part 72 licensees who do not hold 10 CFR part 50, 52, or 53 licenses, and materials licenses with annual fees of $100,000 or greater for a single fee category. The NRC will base the proration of annual fees for terminated and downgraded licenses on the fee rule in effect at the time the action is official. The NRC will base the determinations on the proration requirements under paragraphs (a)(2) and (3) of this section. (1) * * * (ii) The annual fees for new licenses for non-power production or utilization facilities, 10 CFR part 72 licensees who do not hold 10 CFR part 50, 52, or 53 licenses, and materials licenses with annual fees of $100,000 or greater for a single fee category for the current FY, that are subject to fees under this part and are granted a license to operate on or after October 1 of a FY, are prorated on the basis of the number of days remaining in the FY. Thereafter, the full annual fee is due and payable each subsequent FY. (2) 
                            *Terminations.*
                             The base operating power reactor annual fee for operating reactor licensees or the annual fee for small modular reactor licensees, who have requested amendment to withdraw operating authority permanently during the FY will be prorated based on the number of days during the FY the license was in effect before docketing of the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel or when a final legally effective order to permanently cease operations has come into effect. The spent fuel storage/reactor decommissioning annual fee for reactor licensees who permanently cease operations and have permanently removed fuel from the site during the FY will be prorated on the basis of the number of days remaining in the FY after docketing of both the certifications of permanent cessation of operations and permanent removal of fuel from the site. The spent fuel storage/reactor decommissioning annual fee will be prorated for those 10 CFR part 72 licensees who do not hold a 10 CFR part 50, 52, or 53 license who request termination of the 10 CFR part 72 license and permanently cease activities authorized by the license during the FY based on the number of days the license was in effect before receipt of the termination request. The annual fee for materials licenses with annual fees of $100,000 or greater for a single fee category for the current FY will be prorated based on the number of days remaining in the FY when a termination request or a request for a possession-only license is received by the NRC, provided the licensee permanently ceased licensed activities during the specified period. The annual fee for non-power production or utilization facilities will be prorated based on the number of days remaining in the FY when the authorization to operate the facility has been permanently removed from the license during the FY.
 Dated: October 7, 2024. For the Nuclear Regulatory Commission. Carrie Safford, Secretary of the Commission.


# Footnotes

1.
    
                             On December 17, 1982, the NRC issued “Manufacturing License ML-1 to Offshore Power Systems for the manufacture of a maximum of eight floating nuclear plants,” dated September 30, 1982, but the project was subsequently canceled. Back to Citation
 2.
    
                             SECY-22-0072, “Proposed Rule: Alternative Physical Security Requirements for Advanced Reactors,” dated August 2, 2022. Back to Citation
 3.
    
                             A “merchant plant” is a plant licensed to a non-rate-regulated entity (
                            *e.g.,*
                             a nonutility) that engages in the business of production, manufacturing, generating, buying, aggregating, marketing, or brokering electricity for sale at wholesale or for retail sale to the public. Back to Citation
 4.
    
                             The NRC uses the term “operation” in its part 26 discussion to focus on human performance, namely the necessity of individuals to operate, maintain, surveil, and protect the facility and respond to operational transients and unlikely event sequences. Back to Citation
 5.
    
                             Controlled access area and vital area are defined in § 73.2, “Definitions.” Back to Citation
 6.
    
                             See 
                            *e.g.,*
                             Irene Wærø, Ragnar Rosness, and Stine Skaufel Kilska, “Human performance and safety in Arctic environments,” SINTEF (2018). Back to Citation
 7.
    
                             See 
                            *e.g.,*
                             Russell Mannion and Carl Thompson, “Systematic biases in group decision-making: implications for patient safety,” 
                            *International Journal for Quality I Health Care,*
                             Vol. 26, No. 6 (2014): 606-612 (arguing that small group dynamics in healthcare teams produce systematic biases in group decision-making because healthcare professionals may be reticent to vocalize concerns they have about quality of care). Back to Citation
 8.
    
                             See 
                            *e.g.,*
                             Wærø, Rosness, and Kilska (arguing that groupthink leads teams to “develop shared rationalizations that bolster a proposed choice, rather than examining alternative options and identifying the risks associated with the proposed choice”). See also David Hofmann and Adam Stetzer, “A Cross-Level Investigation of Factors Influencing Unsafe Behaviors and Accidents,” 
                            *Personnel Psychology,*
                             Vol. 49 (1996) (finding that in a study of fatal accidents involving offshore oil rigs, in the absence of standard operating procedures, workers “equated normal work methods (
                            *i.e.*
                             what everyone else does) with safe and/or ideal work methods,” revealing that the groupthink phenomena will further cement modes of work that do not reflect safety protocols in small groups that lack strong norms around workplace safety and tacitly reward short-cuts that prioritize efficiency over safety). Back to Citation
 9.
    
                             Mannion and Thompson, “Systematic biases in group decision-making: implications for patient 
                            
                            safety,” 
                            *International Journal for Quality I Health Care,*
                             Vol. 26, No. 6 (2014): 606-612. Back to Citation
 10.
    
                             The Drug Enforcement Administration classifies drugs, substances, and certain chemicals used to make drugs into five (5) distinct categories, depending upon the drug's acceptable medical use and the drug's abuse or dependency potential. These categories appear as Schedules I through V of section 202 of the Controlled Substances Act (21 U.S.C. 812). Schedule I drugs have a high potential for abuse, have no currently accepted medical uses in treatment in the United States, and lack accepted safety for use under medical supervision. At the other end of the classification scheme, Schedule V drugs have the least potential for abuse among the five categories of drugs, have a currently accepted medical use in treatment in the United States, and abuse of the drug may lead to limited physical dependence or psychological dependence. For more information, see 
                            *https://www.dea.gov/​drug-information/​drug-scheduling.* Back to Citation
 11.
    
                             By “well established” the NRC means that there are Federal, State, and non-governmental organizations with reputable and scientifically based resources available for a licensee or other entity to use in its procedures or training to inform individuals of the physiological indications of alcohol impairment or intoxication. Back to Citation
 12.
    
                             The IMP must monitor the initial and continuing trustworthiness and reliability of individuals granted or retaining unescorted AA to a protected or vital area and implement defense-in-depth methodologies to minimize the potential for an insider to adversely affect, either directly or indirectly, the licensee's capability to protect against radiological sabotage. Back to Citation
 13.
    
                             PNNL, Technical Letter Report, “The Use of Electronic Communications to Perform Determinations of Fitness,” dated August 2017. Back to Citation
 14.
    
                             The NRC's security regulations for commercial nuclear power reactors have historically considered onsite armed responders to be the only acceptable method for interdicting and neutralizing threats up to and including the DBT of radiological sabotage. The proposed rule would permit advanced power reactor licensees to use any interdiction and neutralization method, which would be an extension of the Commission's position in SRM-SECY-17-0100, “Security Baseline Inspection Program Assessment Results and Recommendations for Program Efficiencies,” dated October 8, 2018, and SRM-SECY-20-0070, “Technical Evaluation of the Security Bounding Time Concept for Operating Nuclear Power Plants,” dated June 6, 2024. Under the proposed rule, a licensee would retain the responsibility to detect, assess, interdict, and neutralize threats up to and including the DBT of radiological sabotage, but would be able to rely on law enforcement or other offsite armed responders as a method for fulfilling the required interdiction and neutralization capabilities. For licensees that choose to rely on law enforcement to fulfill these capabilities, the proposed rule would not create any NRC regulatory jurisdiction over, or requirements for, law enforcement. In SRM-SECY-23-0021, “Proposed Rule: Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors (RIN 3150-AK31),” dated March 4, 2024, the Commission approved a similar approach to defend against radiological sabotage. Back to Citation
 15.
    
                             Section 70.23(a)(5). Back to Citation


[FR Doc. 2024-23434 Filed 10-23-24; 8:45 am] BILLING CODE 7590-01-P