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

On January 14, 2019, the President signed the 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 of the Code of Federal Regulations (10 CFR) (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 10 CFR part 53 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” is 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 light water-cooled reactors and operational requirements for those types of reactors. This final rule responds to NEIMA by creating an alternative, technology-inclusive regulatory framework to accommodate licensing of future commercial nuclear plants, including advanced reactor designs that may not employ light-water technology. The new alternative requirements and implementing guidance 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 optionality and flexibility for licensing and regulating a variety of technologies and designs for commercial nuclear reactors.

NRC Advanced Reactor Readiness II. Discussion

A. Objective and Applicability

B. Need for Changes to the Existing Regulatory Framework

C. 10 CFR Part 53 Framework III. Part 53 Framework

Subpart A--General Provisions

A. Discussion of Definitions in 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 IV. Changes to Other Parts of 10 CFR Chapter I

10 CFR Part 26

A. Introduction

B. Changes to Part 26, Subparts A Through E and I

C. Requirements for Part 26, Subpart M

D. Changes to Part 26, Subpart N

E. 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

C. Appendix E to Part 50: Emergency Planning and Preparedness for Production and Utilization Facilities

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 V. Opportunities for Public Participation VI. Public Comment Analysis VII. Regulatory Flexibility Certification VIII. Regulatory Analysis IX. Backfitting and Issue Finality X. Cumulative Effects of Regulation XI. Plain Writing XII. Environmental Assessment and Final Finding of No Significant Environmental Impact XIII. Paperwork Reduction Act XIV. Executive Orders

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

B. Executive Order 14154: Unleashing American Energy

C. Executive Order 14192: Unleashing Prosperity Through Deregulation

D. Executive Order 14270: Zero-Based Regulatory Budgeting To Unleash American Energy XV. Congressional Review Act XVI. Criminal Penalties XVII. Voluntary Consensus Standards XVIII. Availability of Guidance XIX. Availability of Documents

In its “Policy Statement on the Regulation of Advanced Nuclear Power Plants,” dated July 8, 1986, the Commission stated that it considered the term “advanced” to apply to reactors that are significantly different from current (i.e., current in 1986) generation light-water reactors (LWRs) then under construction or in operation, and that “advanced” includes reactors that provide enhanced margins of safety or utilize simplified inherent or other innovative means to accomplish their safety functions. At the time, certain high temperature gas-cooled reactors, liquid metal reactors, and LWRs of innovative design were considered to be “advanced.” The 1986 policy statement provided the Commission's policy regarding the review of, and desired characteristics associated with, advanced reactors. The NRC updated this statement in the “Policy Statement on the Regulation of Advanced Reactors,” dated October 14, 2008 (Advanced Reactor Policy Statement).

The agency has undertaken many activities related to advanced reactors, including issuing an advance notice of proposed rulemaking titled “Approaches to Risk-Informed and Performance-Based Requirements for Nuclear Power Reactors,” dated May 4, 2006 (71 FR 26267). These efforts were often done in parallel, and sometimes interwoven, with the NRC's efforts to

improve risk-informed and performance-based approaches within the agency (e.g., the Commission's PRA policy statement, “Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities,” dated August 16, 1995 (60 FR 42622)).

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. These activities leveraged previous work described in NUREG-1860, “Feasibility Study for a Risk- Informed and Performance-Based Regulatory Structure for Future Plant Licensing,” published in 2007. 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 methodology is a technology-inclusive, risk-informed, and performance-based methodology developed for non-LWR designs. The LMP methodology 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 methodology 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 methodology 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 methodology.

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 the LMP methodology 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 methodology 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 methodology is technology-inclusive, the industry and NRC staff initially focused the LMP methodology'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, dated April 13, 2020, the NRC staff developed part 53 by building upon recent and ongoing activities such as the LMP methodology described in SECY-19- 0117. Such an approach supports implementing the NEIMA direction to establish a technology-inclusive framework as well as the 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. The LMP methodology 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 part 53 framework is largely based on the methodology described in SECY-19-0117 and includes a prominent role for PRA, other systematic risk evaluations (SREs), or a combination thereof.

The NRC has long recognized that the licensing and regulation of a variety of nuclear reactor technologies presents

challenges because the existing regulatory framework has evolved primarily to address the LWR designs that compose the current operating fleet. The NRC has had many interactions with designers of various reactor technologies under development, sometimes collectively referred to as advanced 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 has evaluated the ADVANCE Act, including how NRC regulations, such as 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 microreactors, nuclear reactor license application reviews, and nuclear fuel. Finally, in 2025, the President signed E.O. 14300, “Ordering the Reform of the Nuclear Regulatory Commission,” which builds on the provisions in the ADVANCE Act. E.O. 14300 will complement this rulemaking by providing additional mechanisms for streamlining the agency's efforts to provide an efficient licensing pathway for advanced reactors.

The requirements in part 53 support a wide variety of potential commercial nuclear reactor technologies. 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 Statements. The Commission reiterated its safety expectations for new reactors in the SRM for SECY-10-0121, “Modifying the Risk-Informed Regulatory Guidance for New Reactors,” dated March 2, 2011:

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

However, developing a regulatory framework that can accommodate a wide range of technologies while maintaining an acceptable level of safety presents significant regulatory challenges. The existing regulations have been developed over the course of decades and reflect changes to address events discovered through operating experience. As a result, the existing regulations have benefited from a focused and tailored treatment of safety issues as issues arose and evolved. In contrast, part 53 is being developed to accommodate technologies that, in some cases, lack significant operating experience. This lack of operating experience makes it challenging to develop technology- inclusive regulatory requirements when it is less well-known which issues may be more or less important to safety for any given set of technologies. To address these challenges, the NRC drew on well- developed approaches to licensing to produce a technology-neutral and robust regulatory framework. The regulatory framework uses PRAs, other SREs, or a combination thereof, to assess risks and focus on the issues most important to safety, help establish technical requirements, and manage operations. The framework builds on the LMP methodology, which is a technology-inclusive approach to licensing that leverages risk insights to provide applicants with significant design and operation flexibilities.

C. 10 CFR Part 53 Framework

This final 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. The major features of this final rule include the following:

(1) Technology-inclusiveness. This rule provides a broad and flexible regulatory framework that can be used for any reactor technology, any size reactor, and any reactor end use.

(2) Risk-informed framework to support safety-focused decision- making. Part 53 provides a holistic, risk-informed framework that offers substantial flexibility in leveraging safety margins and focusing on design features and programmatic controls important to protecting public health and safety. The framework allows for explicit consideration of risk through the use of PRAs or other SRE techniques, or a combination thereof, to generate risk insights, and to assess and manage those risks. This approach departs from traditional deterministic methods, notably the use of the single-failure criterion, by enabling applicants to propose comprehensive risk metrics and associated risk performance objectives, appropriate systematic risk assessment techniques, and to demonstrate how their design and associated programmatic controls protect public health and safety.

(3) Performance-based approach. Part 53 is a performance-based framework that provides flexibility in establishing appropriate high- level safety objectives and demonstrating how a reactor design or specific commercial nuclear plant meets those objectives. Rather than prescribing specific methods or processes, the performance-based approach in part 53 promotes efficiency and innovation by allowing applicants to propose design features to meet safety objectives and achieve safety outcomes. This will support novel concepts such as leveraging functional containment concepts, alternative siting criteria for commercial nuclear reactors in relation to population centers, reduced staffing

levels, and remote operations, while eliminating traditional, prescriptive requirements, such as general design criteria and aircraft impact assessments.

(4) Licensing pathways that accommodate a broad spectrum of design maturities and deployment models. Part 53 provides several licensing options for applicants to choose from to meet their deployment model or business case needs, including the licenses, certifications, and approvals provided by parts 50 and 52. This final rule provides additional flexibility for manufacturing licenses (MLs), including the possible factory loading of fuel into manufactured reactors with appropriate features to prevent criticality for deployment to another location for operation.

(5) Operator licensing. Part 53 introduces the concept of self- reliant-mitigation facilities and the use of generally licensed reactor operators (GLROs) for those facilities. The allowance for GLROs provides flexibility for the types and locations of staffing needed under part 53.

(6) Efficiency. Part 53 provides opportunities to improve regulatory efficiency by including provisions for licensing first-of-a- kind proposals as well as provisions that benefit those proposing standardized and repetitious applications. Part 53 provides finality to designs for which an operating license has been issued to improve its incorporation into a standardized design approval or certification. Part 53 also provides for a risk-informed approach for managing plant equipment and programmatic controls that reduce the future need for regulatory approvals.

(7) Codes and standards. Part 53 does not incorporate by reference specific codes and standards as is done in Sec. 50.55a, “Codes and standards.” Instead, part 53 allows the use of generally accepted codes and standards to be tailored to the assessed safety significance of SSCs, such as the use of non-nuclear codes and standards for SSCs composed of commercial grade components.

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 complement the design and analysis requirements in subpart C. The performance-based approach adopted in part 53 also includes regulatory requirements that 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.

Subpart M of part 26 is new and is 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. Subpart M of part 26 is 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 requirements are not entirely equivalent because current subpart K of part 26 only applies during construction of the commercial nuclear plant, whereas subpart M of part 26 applies during construction, operation, and decommissioning. Furthermore, subpart M of part 26 allows 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.

Revisions to part 73 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 final rule makes conforming changes throughout 10 CFR chapter I, by adding “and part 53” where appropriate to account for the addition of part 53.

Subpart A provides the general provisions applicable to all applicants and licensees that are 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 AEA) and title II of the Energy Reorganization Act of 1974 (88 Stat. 1242). Subpart A includes 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 are largely equivalent to the general requirements in part 50 that are applicable to all part 50 applicants and licensees (specifically, Sec. Sec. 50.1 through 50.13) but reference the corresponding regulations in part 53 in place of references to part 50. A. Discussion of Definitions in Part 53

This final rule includes a definition section in Sec. 53.020. The definitions of most terms in Sec. 53.020 are equivalent to the corresponding terms defined in: (1) Sec. Sec. 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 Sec. 53.020, as noted previously, the NRC initially considered establishing the scope of part 53 as being for “advanced nuclear plants.” The preliminary proposed rule language defined “advanced nuclear plant” as “a utilization facility consisting of one or more advanced nuclear reactors” as defined in NEIMA. NEIMA defines the term “advanced nuclear reactor” as “a nuclear fission reactor or fusion machine, including a prototype plant (as defined in sections 50.2 and 52.1 of 10 CFR (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” is 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 https://www.regulations.gov.

The NRC allows the use of part 53 by any “commercial nuclear plant.” The use of the term “plant” versus “reactor,” as used in existing regulations (i.e., Sec. 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 Sec. 50.2. However, the phrase “in a self- supporting chain reaction” is not included in the definition of Commercial nuclear plant 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 Sec. 53.020 based on the definition of that term in Sec. 50.2 and refers to a “Commercial nuclear plant” as defined in Sec. 53.020.

The definition of “Construction” is different from the definition in Sec. 50.10. Because the regulatory framework in part 53 uses risk- informed, less prescriptive, and performance-based requirements as compared to part 50, the part 53 definition takes a different approach in determining what activities are prohibited without an NRC license. Under the part 53 approach, the definition of “Construction” specifies a variety of activities that are applicable to safety-related (SR) and non-safety-related but safety-significant (NSRSS) SSCs and are credited or relied upon for demonstrating compliance with safety criteria defined in subpart B of part 53 as well as SSCs necessary to comply with part 73 and onsite emergency facilities necessary to comply with Sec. 53.855. By listing the activities for SR and NSRSS SSCs that are credited or relied upon for demonstrating compliance with safety criteria defined in subpart B, this definition describes activities related to SSCs subject to some sort of special treatment, as that term is defined in Sec. 53.020. These special treatment requirements, which include quality assurance, design criteria, and programmatic controls, apply to safety-related SSCs and the set of non-safety-related SSCs for which a license is required to authorize construction activities. The latter category includes a facility's NSRSS SSCs. The non-safety- significant SSCs not subject to special treatment and NSRSS SSCs for which special treatments are limited to operational controls are, in general, identified as “commercial grade” and may be designed, procured, and installed in accordance with the usual practices employed for industrial plants. Importantly, under the part 53 definition, an SSC that falls outside the definition of construction may still be subject to the NRC's statutory authority during operations. In view of the foregoing, the definition of “Construction” in Sec. 53.020 is consistent with the provisions of the AEA related to construction permits, while simultaneously allowing activities related to SSCs that are commercial grade but which could still be subject to the NRC's jurisdiction during operations. This definition also includes the listed activities which are for SSCs necessary to comply with part 73 or onsite emergency facilities necessary to comply with Sec. 53.855. The inclusion of the listed activities which are for these SSCs is consistent with Sec. 50.10(a)(1)(v) and (vii), which include activities for corresponding SSCs. Including these activities in the definition of “Construction” is appropriate because, in both instances, part 53 points back to the relevant existing frameworks in part 73 and the relevant part 50 requirements, respectively, rather than creating an entirely new framework. Section 53.020 also adds 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 defines “Safety criteria” in terms of the plant- level performance-based metrics that are provided in Sec. Sec. 53.210 and 53.220. The term “Functional design criteria” is 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” refers to those SSCs needed to meet the safety criteria in Sec. 53.210. The term “Non-safety-related but safety-significant SSCs” means those SSCs that are not SR because they are not relied upon to perform any function necessary to demonstrate compliance with Sec. 53.210 but warrant special treatment because they are relied on to achieve adequate defense in depth or perform risk-significant functions. The term “Non-safety-significant SSCs” means those SSCs that are not SR or NSRSS.

The term “Programmatic controls” means administrative measures that govern human action in implementing programs and operating, monitoring, and maintaining SSCs and equipment of a commercial nuclear plant.

The terms “Design-basis accidents,” “Anticipated event sequences,” “Unlikely event sequences,” and “Very unlikely event sequences” are defined to be different types of “Licensing-basis events” and are also largely equivalent to the LMP methodology's definitions of DBAs, anticipated operational occurrences (AOOs), design-basis events (DBEs), and beyond-design-basis events, respectively. The term “Design-basis accidents” is 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 are derived from the unlikely event sequences from the PRA, a type of SRE, other SREs, or a combination thereof, 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 1x10\-2\/plant-year and greater are classified as anticipated event sequences. Within the LMP methodology, infrequent event sequences with mean frequencies of 1x10\-4\/plant-year to 1x10\-2\/plant-year are classified as unlikely event sequences. “Very unlikely event sequences” are less likely to occur than unlikely event sequences. Within the LMP methodology, rare event sequences with frequencies of 5x10\-7\/plant-year to 1x10\-4\/plant- year are classified as very unlikely event sequences. While the terminology for these event sequences creates some differences between part 53 and the LMP methodology, part 53 uses 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. Finally, although the term “event sequence” is often used in the context of a PRA, that term is used generically in part 53 and does not imply the use of a specific type of SRE, such as a PRA. The sections of this preamble related to subparts B and C provide additional discussion of LBEs.

Section 53.020 includes a definition of “Special treatment” to explain that it means those requirements, such as quality assurance (QA), design criteria, and programmatic controls, that are taken beyond the procurement, installation, and maintenance of commercial grade products. Routine commercial practices may include the use of selected consensus codes and standards that are cited in applications to support the identification of special treatments that may go beyond what would otherwise be required by those selected commercial codes and standards. The special treatments increase confidence 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 subpart C. Structures, systems, and components designated as SR 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 Sec. 53.210.

To maintain alignment with definitions in part 52, the NRC has added a definition of early site permit (ESP). The NRC proposed definitions for “Consensus code or standard” and “probabilistic risk assessment” but is not including a definition for these terms in this final rule because these terms were determined not to be essential for the framework and including the definitions could introduce issues with consistency given alternative definitions developed by other organizations. B. Other General Provisions

Section 53.040 governs written communications and how applications and other required information must be submitted to the NRC. These requirements are equivalent to those in Sec. 50.4.

Section 53.050 establishes 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 are equivalent to those in Sec. 50.5.

Section 53.060 prohibits 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 prescribes a procedure for seeking a remedy for employees who believe they have been discriminated against for engaging in such protected activities. These requirements are equivalent to those in Sec. Sec. 50.7 and 52.5.

Section 53.070 governs the completeness and accuracy of information provided to the NRC. These requirements are equivalent to those in Sec. Sec. 50.9 and 52.6.

Section 53.080 governs exemptions from the requirements of the regulations in part 53. These requirements are equivalent to those in Sec. Sec. 50.12 and 52.7.

Paragraphs (a) through (d) of Sec. 53.090 establish requirements for standards that the NRC will consider in determining whether a construction permit (CP), operating license (OL), ESP, combined license, or ML under part 53 will be issued to an applicant. These requirements are equivalent to those in Sec. Sec. 50.40, 50.42, 50.43 and 50.22, respectively. Requirements equivalent to those in Sec. Sec. 50.41 and 50.21 are not included in part 53 because they apply to Class 104 licenses, and part 53 does not apply to those licenses.

Section 53.100 requires that no license issued under part 53 may cover activities that are not under or within the jurisdiction of the United States. These requirements are equivalent to those in Sec. 50.53.

Section 53.110 states that licensees and applicants are not 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 are equivalent to those in Sec. 50.13.

Section 53.115 establishes requirements for rights related to SNM. These requirements are equivalent to those in Sec. 50.54(b) and (c).

Section 53.117 establishes 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 are equivalent to those in Sec. 50.54(d).

Section 53.120 establishes requirements for information collection requirements that have received Office of Management and Budget (OMB) approval. These requirements are equivalent to those in Sec. 50.8.

Subpart B, “Technology-Inclusive Safety Requirements,” provides technology-inclusive safety criteria that serve as performance standards for the subsequent performance-based requirements used throughout part 53. Subsequent subparts 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 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 is built upon the safety criteria in Sec. 53.210, “Safety criteria for design-basis accidents.” The more detailed requirements for those SSCs are then 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 are governed by subpart E, “Construction and Manufacturing Requirements,” and subpart F, “Requirements for Operation.”

Requirements for NSRSS SSCs warranting special treatment are

likewise determined under Sec. 53.220, “Safety criteria for licensing-basis events other than design-basis accidents,” in subpart B and Sec. 53.460, “Safety categorization and special treatment,” in subpart C. Regulatory requirements related to the NSRSS SSCs are distinguished from the regulatory requirements for SR SSCs throughout part 53. Part 53 affords more flexibility to applicants and licensees regarding how NSRSS SSCs are used in the design and maintained during plant operations, as compared to SR SSCs.

The collective set of performance-based requirements in part 53 are sufficient, if met, for the NRC to make the findings required to grant an application for a utilization facility under section 182 of the AEA 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 is similar to existing NRC regulations, which the Commission has said on many occasions do not specifically define “adequate protection.” However, compliance with NRC regulations may be presumed to assure adequate protection at a minimum. The requirements throughout part 53 that support demonstrating compliance with Sec. 53.220 are 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 AEA.

Consistent with historical practice, sections 182 and 161 of the AEA are cited as authorizing legislation within this final rule. However, specific language from the AEA is not incorporated into the safety objectives or safety criteria in part 53. This is because, again consistent with historical practice, the NRC is not defining “adequate protection” through the individual safety requirements in part 53. Rather, part 53 enables the NRC to make its required findings under the AEA 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 provides safety criteria for DBAs that are required to be identified under Sec. 53.240 and analyzed under Sec. 53.450(f) in subpart C of part 53. Subsequent sections in part 53 require that the SSCs relied upon to demonstrate compliance with the criteria in Sec. 53.210 be classified as SR. The use of SR SSCs and the 25 rem reference values for potential radiological consequences aligns with traditional deterministic approaches for LWRs from Sec. Sec. 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 Sec. 50.34(a)(1)(ii)(D)(1) and Sec. 52.79(a)(1)(vi)(A) is included in Sec. 53.210 to explain that the use of the 25 rem value 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 that is 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 provides a logical structure supporting the identification and treatment of SR SSCs and establishing the corresponding functional design criteria for those SSCs.

Section 53.220 provides safety criteria for LBEs other than DBAs that are required to be identified under Sec. 53.240 and analyzed under Sec. 53.450(e) in subpart C. Whereas Sec. 53.210 and the related requirements for SR SSCs provide that a defined success path exists for DBAs, the safety criteria for LBEs other than DBAs establish the connections between SSC design, human actions, and programmatic controls and a broader set of potential internal and external hazards. These safety criteria also address defense-in-depth matters such as a balanced consideration of prevention and mitigation.

The safety criterion in Sec. 53.220(b) includes 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 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 pre-established, 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 requirements for analyses in Sec. 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 do 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, is one of several performance standards used in subpart B. The use of multiple performance standards, including deterministic criteria and defense-in- depth measures, reflects an integrated decision-making process similar to that described in RG 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,” Revision 3. The NRC's approval of using a comprehensive risk metric or set of metrics with associated risk performance objectives is not, by itself, an indicator of adequate protection. Rather, the comparison of comprehensive risk metrics to associated risk performance objectives that are acceptable to the NRC is part of a suite of regulatory requirements that,

when considered holistically, form the basis for the NRC's decision- making. This is analogous to the approach used for plants licensed under part 50 and part 52, where no single regulatory requirement governs whether a plant is “safe enough.”

The RG 1.233, “Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors,” describes an example of an acceptable approach for identifying and analyzing LBEs under part 50 and part 52, including the use of the quantitative health objectives (QHOs) stated in the NRC's policy statement, “Safety Goals for Nuclear Power Plant Operation,” dated August 4, 1986 (51 FR 28044), as corrected and republished August 21, 1986 (51 FR 30028) (Safety Goals Policy Statement), as acceptable performance objectives for comprehensive risk metrics. The use of comprehensive risk metrics, such as the individual early fatality risk (IEFR) and the individual latent cancer fatality risk (ILCFR), and associated risk performance objectives, such as the QHOs, from the Safety Goals Policy Statement, could form the basis for one approach to meet Sec. 53.220(b). The requirement for comprehensive risk metrics, in combination with the other requirements in subparts B and C, brings 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 provides a logical performance objective to support the risk management approaches in the various subparts comprising 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 Sec. 53.220(b) can 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 final rule establish acceptable risks, not acceptable deaths.

Applicants under 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 previously discussed, including the use of surrogate measures for use in specific analyses to satisfy the requirements in Sec. 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 will, as appropriate, review novel approaches for comprehensive risk metrics and associated risk performance objectives proposed by applicants, industry organizations, or standard development organizations and will engage stakeholders during the development of the related regulatory guidance or specific licensing actions.

Section 53.230 requires safety functions needed to ensure that the safety criteria under Sec. Sec. 53.210 and 53.220 can be met if an assumed LBE were to occur at a commercial nuclear plant. Section 53.230 specifies 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) can 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 final 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.

Section 53.240 requires 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 Sec. 53.450 are below the safety criteria identified under Sec. Sec. 53.210 and 53.220 and that SSCs, personnel, and programs address the safety functions from Sec. 53.230. Including a high-level performance requirement related to the identification of LBEs to address appropriate risk- informed combinations of malfunctions of plant SSCs, human errors, facility hazards, and the effects of external hazards and analysis thereof in subpart B reflects the historical and continuing importance of evaluating unplanned events as part of the licensing of commercial nuclear plants. Section 53.240 requires identification and analysis of LBEs under Sec. 53.450 using a PRA, other SREs, or a combination thereof. An example of acceptable methods of using PRAs to identify and assess LBEs is the methodology in RG 1.233, as discussed in RG 1.254, “Technology-Inclusive Identification of Licensing Events for Commercial Nuclear Plants.”

Section 53.250 establishes 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 necessitates this specific requirement to ensure defense in depth is provided. The requirements in this section state that no single engineered design feature, human action, or programmatic control, no matter how robust, should be exclusively relied upon to address the range of LBEs other than DBAs. The requirement under Sec. 53.250(c) is different from the single failure criterion described in appendix A to part 50. The Sec. 53.250(c) requirement does not allow the safety analysis to exclusively rely upon a

single engineered design feature, human action, or programmatic control to address the range of LBEs other than DBAs (i.e., ranging from very unlikely event sequences to anticipated event sequences). In contrast, the single failure criterion under appendix A to part 50 relates, in part, to the failure of a component to perform its intended safety function, regardless of whether that component was exclusively relied upon to address the range of LBEs. This means the requirement under Sec. 53.250(c) does not strictly disallow single failures, as defined in appendix A to part 50, because a component could experience such a single failure and, if it is not otherwise exclusively relied upon to address the range of LBEs other than DBAs, its failure alone does not preclude being able to satisfy Sec. 53.250(c). In that regard, Sec. 53.250 allows for greater flexibility such that other measures could be taken to ensure appropriate defense in depth without needing to accommodate single failures, as defined in appendix A to part 50. The phrase “engineered design feature” does not preclude the possible crediting of inherent characteristics within the design and analysis for commercial nuclear reactors. While defense in depth is only assessed for LBEs other than DBAs, the need to ensure dedicated success paths for DBAs contributes to the overall defense in depth for each commercial nuclear plant under part 53.

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

Section 53.270 provides for the protection of plant workers and establishes a level of safety based on requirements in 10 CFR part 20, which limit occupational dose.

This subpart provides requirements for the design of commercial nuclear plants and the supporting analyses, including the analyses of LBEs, to demonstrate that the performance standards in subpart B can be satisfied. The sections within subpart C reflect the overall hierarchy throughout part 53, which covers: (1) plant-level safety criteria (Sec. Sec. 53.210 and 53.220); (2) safety functions (Sec. 53.230) needed to demonstrate compliance with the safety criteria; (3) design features (Sec. 53.400), human actions, and programmatic controls needed to fulfill the safety functions; and (4) functional design criteria (Sec. Sec. 53.410 and 53.420) that must be defined for each design feature relied upon to demonstrate the safety criteria (Sec. Sec. 53.210 and 53.220) are met. Subpart C also contributes 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 Sec. 53.210. NSRSS 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 establishes a requirement that design features be provided for each commercial nuclear plant to satisfy the safety criteria and fulfill safety functions from subpart B during LBEs. Other sections in subpart C, 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 requires that functional design criteria be defined for safety-related design features relied upon to demonstrate that the consequences from DBAs would be below the criteria in Sec. 53.210 through analyses performed under Sec. 53.450(f), which includes insights from both PRAs and deterministic analyses. Other sections within part 53 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 consideration of human actions and programmatic controls in the identification of special treatments associated with the design of SR SSCs will contribute to ensuring that a commercial nuclear plant licensed under part 53 would meet the safety criteria in Sec. 53.210.

Section 53.415 requires 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, Sec. 53.415 requires that SR SSCs remain capable of performing the safety functions stated in Sec. 53.230 for which they are credited up to the design-basis external hazard levels as determined under Sec. 53.510. As used in Sec. 53.415 and subpart D of part 53, a hazard level refers 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 will 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 requires 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 are described in Sec. 53.450(e), which requires that those events be identified and assessed using a PRA, other SREs, or a combination thereof, together with other generally accepted approaches for systematically evaluating engineered systems. The SSCs determined to be safety significant (i.e., either SR or NSRSS) will have associated special treatment requirements as specified in Sec. 53.460. Special treatment is defined in subpart A of part 53 and generally refers to measures (e.g., quality assurance, testing, monitoring) taken beyond normal commercial practices related to the procurement, installation, and maintenance of commercial grade products to provide confidence that SR and NSRSS SSCs will perform under the service conditions and with the reliability assumed in the analysis under Sec. 53.450(e) and will comply with the applicable functional design criteria. Such normal commercial practices include the use of consensus codes and standards, as identified in an application, to support the identification of special treatments that include measures that may go beyond the use of commercial codes and standards. 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 safety criteria in subpart B. Therefore, designers and licensees for commercial nuclear plants are 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 Sec. 53.220(b).

Section 53.425 establishes 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 provides similar requirements for design features and related functional design criteria for protection of plant workers to meet the safety criteria in part 20. Section 53.425 provides applicants and licensees flexibility to define design objectives for design features related to controlling liquid, gaseous, and solid wastes as required under part 20. The design objective will assist designers, applicants, and licensees in performing the evaluations of possible reductions in public dose from routine effluents when considering costs and other factors.

The requirements in Sec. Sec. 53.425 and 53.430 for design features and functional design criteria to support radiation protection activities have parallels in existing regulations such as Sec. 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 addresses various design requirements that warrant specific mention to ensure that the design features required by Sec. 53.400 comply with the functional design criteria required by Sec. Sec. 53.410 and 53.420. These requirements will be met through design practices, consideration of testing and operating experience, and various assessments of LBEs and other potential challenges to commercial nuclear plants. Discussion of some of the key design requirements included in section 53.440 follow.

(1) Sec. 53.440(a): An essential element to ensuring a proposed design can comply with the performance criteria in part 53 is that the ability of design features to fulfill their safety functions is demonstrated by a combination of analyses, test programs, prototype testing, and operating experience. This requirement closely aligns with the language in Sec. 50.43(e) and is included in part 53 as the same foundational requirement. In addition, Sec. 53.440(a) requires 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 requirement corresponds to the existing requirement under Sec. 50.34(f)(3)(i) that was developed in response to the 1979 accident at Three Mile Island Nuclear Generating Station.

(2) Sec. 53.440(b): The design and licensing of commercial nuclear plants should use generally accepted consensus codes and standards for design features classified as safety-related. 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 will 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.

(3) Sec. 53.440(c): The design requirements in subpart C require the materials used for SR and NSRSS SSCs to be qualified for their service conditions over the design life of the SSC as appropriate to satisfy the special treatments established for the SSC under Sec. 53.460.

(4) Sec. 53.440(d): The requirements in Sec. 53.440 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.

(5) Sec. 53.440(e) and (f): The design requirements in subpart C 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. Under Sec. 53.440(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. This approach ensures considerations are given for safety and security together throughout the plant's lifetime, including the design process and prior to implementing changes to plant configurations, to ensure risks are effectively managed. The implementation of a security strategy and design features early in the design process has the potential to be more efficient and cost-effective rather than implementing these features after the plant has been designed and constructed.

(6) Sec. 53.440(g) and (h): Specific design requirements will ensure that commercial nuclear reactors under part 53 have the capability to achieve and maintain subcriticality and long-term cooling. The requirements are 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.

(7) Sec. 53.440(i): The design, analysis, and development of programmatic controls under part 53 will consider the number of reactor units and other significant inventories of radioactive materials contributing to the risks to public health and safety. This reflects the definition of “Commercial nuclear plant” in subpart A and reinforces that the evaluation of LBEs is performed on a plant-wide basis. This aspect of part 53 is different from parts 50 and 52, which generally define safety requirements on the assumption of events involving only individual reactor units.

(8) Sec. 53.440(k): The inclusion of a specific requirement for design features and related functional design criteria, including associated programmatic controls or a combination thereof, 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 is similar to the existing requirements in 10 CFR part 70, “Domestic Licensing of Special Nuclear Material,” that address both potential radiological and chemical hazards for licensed materials at fuel cycle facilities.

(9) Sec. 53.440(l): These provisions 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 Sec. 20.1406.

(10) Sec. 53.440(m): This design requirement provides a technology-inclusive equivalent to the requirements in Sec. 50.68 by including options for commercial nuclear plants to either have a monitoring system capable of detecting a criticality as described in Sec. 70.24 or to have restrictions on SNM

handling and storage that would prevent inadvertent criticality events.

(11) Sec. 53.440(n): The design needs 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.

One notable exclusion from the design requirements in the part 53 proposed rule is an explicit requirement to consider and address the potential impact of a large, commercial aircraft, as is currently required of parts 50 and 52 applicants under Sec. 50.150, “Aircraft impact assessment.” When the Commission promulgated the aircraft impact final rule on June 12, 2009 (74 FR 28112), it noted that “the impact of a large aircraft on the nuclear power plant is regarded as a beyond-design-basis event” and it was “the NRC's view that effective mitigation of the effects of events causing large fires and explosions (including the impact of a large, commercial aircraft) can be provided through operational actions,” which were covered by other requirements. In light of this view, the Commission stated that “the mitigation of the effects of aircraft impacts through design should be regarded as a safety enhancement which is not necessary for adequate protection.” In the Regulatory Analysis that accompanied the aircraft impact rule, the NRC quantified the costs of the rule, but did not quantify the benefits of the rule, stating that the “benefits of the final rule can be evaluated only on a qualitative basis.” The NRC concluded that the key benefit of the rule was “improvement in knowledge.” The Commission acknowledged that “it is difficult to quantify the safety enhancement gained through implementation of the aircraft impact rule,” but stated that “the NRC nevertheless believes that the cost of performing the assessment and incorporating the results into the design . . . is justified in view of the increased safety provided by implementation of the aircraft impact rule.”

It has been over 15 years since the promulgation of the aircraft impact rule in 2009. Events like the terrorist attacks of September 11, 2001, are now much less likely due to significant increases in security at commercial aviation facilities as well as hardened access to aircraft cockpits. In addition, it is not clear that the Commission's previous belief that the cost of implementation of the aircraft impact rule was justified by the increase in safety provided by the rule would hold true for future reactors licensed under part 53. As stated previously, the NRC concluded that the key benefit of the rule was “improvement in knowledge” achieved by performing the aircraft impact assessment. It's worth noting that licenses issued under parts 50 and 52 were largely based on deterministic analyses of the safety of the facility relying on the General Design Criteria. The technical requirements in part 50 were supplemented over the years to address specific beyond-design-basis events, such as the loss of large areas of the plant due to fires and explosions. In contrast, under part 53, applicants will be required to perform a comprehensive assessment of their reactor design to identify potential failures, susceptibility to internal and external hazards, and other contributing factors that could pose a risk to public health and safety. The spectrum of events and hazards considered will include those that have traditionally been considered design-basis events and those that have been considered beyond-design-basis events. Although part 53 does not include prescriptive requirements to assess a licensing-basis event comprising an intentional act that could cause large fires or explosions, it does require applicants to assess a full spectrum of unplanned events, to include anticipated events, unlikely events, and very unlikely events. The NRC believes that the systematic evaluations of internal hazards, external hazards, and security threats under part 53 and part 73 sufficiently address the potential loss of large areas of the plant due to explosions or fire currently addressed under Sec. 50.155(b)(2).

Therefore, part 53 applicants will have considered how to mitigate the broader potential plant impacts that may result from an event such as the impact of a large aircraft. As a result, applicants and licensees under part 53 will have substantially more information about the design of their facilities than applicants and licensees did before the promulgation of the aircraft impact rule. Accordingly, the “improvement in knowledge” to be gained by requiring a separate assessment of the impact of a large commercial aircraft under part 53 is expected to be significantly less than the improvements in knowledge for part 50 or 52 applicants the Commission estimated when it promulgated the aircraft impact rule. Because the potential impact of beyond-design-basis events are considered in other ways under part 53, the NRC concludes that the cost of performing a separate aircraft impact assessment and incorporating the results into the design of a commercial nuclear plant licensed under part 53 would not be justified. For these reasons, this final rule does not contain requirements for applicants to assess the impact of a large, commercial aircraft on the design of the facility.

Section 53.450 establishes analysis requirements and centers upon the use of a PRA, other SREs, or a combination thereof with other generally accepted approaches for systematically evaluating engineered systems. The use of PRA, other SREs, or a combination thereof as a key component in the analysis requirements for part 53 reflects the decades of improvements in the use of such methodologies and their increasing use 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. This policy statement also acknowledges the variability in the characteristics of events considered and the associated complexity of engineered systems related to different regulatory activities and that risk-informed analysis techniques of varying complexity may be employed to yield meaningful insights and results. In that regard, the use of PRA, other SREs, or a combination thereof under part 53 needs to be commensurate with the complexity of the analyzed systems and their behaviors, with consideration of all aspects of operations. 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 maintains 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, other SREs, or a combination thereof will be used together with other techniques in part 53 to identify and categorize LBEs, classify SSCs, evaluate defense in depth, and inform the appropriate special treatments for SSCs. This increased role for PRA and SREs

necessitates that they be developed, performed, and maintained in accordance with NRC approved standards and practices (see Sec. 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 must be qualified for the range of conditions being simulated across a wide range of unplanned events. These analyses must use realistic approaches and address uncertainties associated with states of knowledge, modeling, and performance of SSCs.

While industry consensus PRA standards and PRA peer review processes endorsed in RGs 1.200 and 1.247 remain acceptable for developing a PRA, they are not regulatory requirements and an application under part 53 need not follow every aspect of the applicable consensus PRA standard. Existing processes for defining the scope and capability of a PRA supporting an application offer flexibility in determining the degree to which the PRA needs to be developed and may be informed by other factors such as design complexity and the needed degree of realism and level of detail, consistent with the use of the PRA with SREs and the 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 regulatory requirements for the specific uses of analyses under Sec. 53.450(b). The specific uses of analyses in Sec. 53.450(b) are to inform LBE selection; inform classification of SSCs according to safety significance; evaluate adequacy of defense in depth; identify and assess all plant operating states with a potential for uncontrolled release of radioactivity to the environment; 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; and inform the establishment and updating of appropriate measures for plant operations, including availability controls, to ensure configurations and special treatments for SR SSCs and NSRSS SSCs provide the capabilities, availability, and reliability consistent with satisfying the high-level safety criteria in Sec. 53.220.

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.

For risk contributors that are excluded from PRA logic models or PRA screening processes and are otherwise analyzed by an SRE--also referred to as supplementary analyses--the NRC plans to develop guidance for determining the acceptability of such SREs.

Section 53.450(c) requires periodic maintenance and upgrading of the PRA, other SREs, or a combination thereof 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 Sec. 53.220 and the evaluation criteria of Sec. 53.450(e)(2). The periodic maintenance of the PRA, other SREs, or a combination thereof is also 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, other SREs, or a combination thereof and other requirements in part 53 will 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.

Section 53.450(d) provides requirements for the qualification of the analytical codes used in modeling the physical behavior of plant systems and that those codes must be qualified for the range of conditions for which they are to be used.

The categories of LBEs used in part 53 include anticipated event sequences, unlikely event sequences, and very unlikely event sequences. The unlikely event sequences 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 Sec. 53.450(e), the analyses of LBEs other than DBAs will be used not only to show the performance criteria of Sec. 53.220 are satisfied but also to show that evaluation criteria defined for each LBE or category of LBEs are satisfied. Such evaluation criteria for specific LBEs or categories of LBEs are 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. An applicant's or licensee's defining of evaluation criteria under Sec. 53.450(e) and the risk performance objectives under Sec. 53.220(b) are also part of the integrated approach within part 53 where the analyses from subpart C are used for decisions on design, siting, and operations. As an example, an applicant or licensee could propose to justify siting proposals by defining their evaluation criteria such that the calculated consequences for an individual at the exclusion area boundary are less than the total effective dose equivalent (TEDE) values used in graded approaches to assessing population densities under subpart D. Another requirement for the Sec. 53.450(e) analyses is that the methodology must 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 maintains an important role for a deterministic analysis of DBAs in the performance criteria of Sec. 53.210 and the related analytical requirements in Sec. 53.450(f). The analysis of DBAs will 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 set forth in this section, DBAs must 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 Sec. Sec. 53.760 through 53.795.

While the DBAs analyzed under part 53 are 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 part 53. In part 53, the role of the

DBA analysis is 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 Sec. 53.210. The overall control of risks posed by commercial nuclear plants under part 53 is provided by the analyses of and measures taken for both DBAs and other LBEs, including very unlikely event sequences. This contrasts with the traditional deterministic approach in part 50 wherein the analyses of DBEs such as DBAs were used to provide bounding assessments, to 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(g) includes provisions to ensure that analyses are performed to support the design requirements of Sec. 53.440(e) on fire protection and Sec. 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 analysis requirements related to fire protection 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 establishes criteria for the safety classification of SSCs and determination of appropriate special treatments. As noted in subpart A, the term “Special treatments” is defined to mean those items, such as measures taken to satisfy functional design criteria, quality assurance, and programmatic controls, that provide assurance that certain SSCs will provide defense in depth or perform risk- significant functions. These requirements also provide confidence that the SSCs will perform under the service conditions and with the reliability credited in the analysis performed in accordance with Sec. 53.450 to satisfy the safety criteria in Sec. Sec. 53.210 and 53.220. The terminology used in part 53 includes the following categories for SSC classification: (1) SR; (2) NSRSS; and (3) non-safety significant. Requirements for SR SSCs are defined in other sections of part 53 and include using TSs for controls during operation and the application of quality assurance requirements from appendix B to part 50.

Requirements for NSRSS SSCs include the need to identify necessary special treatments such as performance measures on reliability. Licensees will generally be afforded flexibility in maintaining and changing special treatments for SSCs categorized as NSRSS. Non-safety- significant SSCs will be addressed under normal licensee programs for commercial grade equipment and typical industry practices for general plant design and maintenance. Safety-related SSCs 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 Sec. 53.220 and the evaluation criteria in Sec. 53.450(e).

Section 53.480 establishes seismic design considerations. This section relates to the safety criteria in subpart B, the analytical requirements related to external hazards in Sec. 53.450, and subpart D, “Siting Requirements.” For licenses issued under part 53, this section in subpart C will 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 Sec. 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 Sec. 53.450 must address seismic hazards as well as other external hazards. The expected responses of SSCs to a range of seismic events must be included in the analyses when ensuring that the safety criteria defined under Sec. 53.220 will 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 in part 53 states requirements for the siting of commercial nuclear plants and serves the role provided by 10 CFR part 100, “Reactor Site Criteria,” for nuclear reactors licensed under parts 50 and 52. As reflected in Sec. 53.500, the reason for establishing siting requirements remains 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.

Section 53.510 requires 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 are used in Sec. 53.415, which requires 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 need to account for uncertainties and variabilities in data, models, and methods used to characterize those hazards. Existing approaches can 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 are reflected in Sec. 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 Sec. 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,” is 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 needs to be sufficient to inform a site-specific design (e.g., a CP or custom combined license (COL)) or confirm the use of a standard design for a commercial nuclear plant under Sec. 53.480 and other sections of subpart C. A risk-informed approach can use several design-basis ground motions (DBGMs) to assess SSCs in various SDCs (i.e., one DBGM per SDC). Section 53.510(d) states 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 requires applicants to identify and assess site characteristics related to topics that include meteorology, geology, hydrology, or other areas in the design and analyses required under subpart C.

Section 53.530 sets requirements for population-related considerations and largely maintains requirements and definitions similar to those currently in part 100 for an exclusion area, low population zone, and population center distance. The NRC recognizes that some applicants may propose to essentially collapse the exclusion area and low population zone to the site boundary. This approach would rest on a demonstration that the calculated consequences of DBAs remain below the dose guidelines used in Sec. 53.210, which are the same as those in the existing regulations in parts 50, 52, and 100. The definitions in Sec. 53.020 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 is maintained in part 53 by using the current language from part 100 as one option under Sec. 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.” The NRC recognizes that safety, environmental, economic, or other factors may justify siting commercial nuclear plants in areas with higher population densities or within a densely populated center containing more than about 25,000 residents. Therefore, an option is included within Sec. 53.530 for such sites to be proposed using assessments of additional societal risks associated with siting a reactor in areas of higher population density (e.g., potential increases in population dose or economic consequences from reactor accidents) in comparison to the societal benefits of a specific site (e.g., ability to use existing infrastructure for a retired fossil fuel power plant). Site-related requirements in part 20 (restricted area) and part 73 (protected and owner-controlled areas) remain applicable to commercial nuclear plants licensed under part 53.

Section 53.540 requires that site characteristics be appropriately considered in other activities such as the design and analysis performed under subpart D of part 53 and the emergency planning and security programs under subpart F of part 53.

The part 53 language establishes construction and manufacturing requirements in subpart E. The language for construction-related activities largely reflects current requirements in part 50 without any fundamental changes. Limited changes were 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 language for requirements for manufacturing activities largely mirrors those for construction-related activities. However, the 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 is not addressed by these subparts.

Section 53.600 establishes the overall construction and manufacturing requirements for CPs, OLs, COLs, MLs, and limited work authorizations (LWAs). This section connects the construction and manufacturing requirements to the safety criteria, quality assurance requirements, and other requirements located in other subparts. These requirements require that construction and manufacturing activities be managed and conducted such that when combined with associated design features and programmatic controls, the constructed plant will satisfy the relevant requirements in subpart B.

Section 53.605 establishes requirements for the reporting of defects and instances of noncompliance during construction. This section provides equivalent requirements to those in Sec. 50.55(e).

Section 53.610(a) establishes the requirement to have in place a well-defined command and control structure to manage construction activities. The requirements generally reflect current requirements, with an emphasis on the quality assurance programs for complying with the requirements in appendix B to part 50. Section 53.610(a)(6) requires programmatic controls for implementing special treatment for NSRSS SSCs to align with requirements in other subparts in part 53. The section also refers to other NRC regulations to address matters such as requirements to have an 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) provides requirements governing construction activities, including the equivalent of the requirement in Sec. 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) requires 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 are 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) addresses 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) includes 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.

\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.

The part 53 sections in subpart E for manufacturing and in subpart H for licensing matters maintain requirements largely equivalent to those in part 52 for MLs. The NRC approval of a standard design and related manufacturing processes, coupled with a stable workforce and established procedures, has the potential for maintaining and even improving the quality and consistency of manufacturing, as compared to the traditional method of constructing reactors onsite by a variety of contractors and subcontractors.

Subpart E includes requirements that 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, Sec. 53.620 establishes 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 includes 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 requirements in Sec. 53.620(b) that 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 is also incorporated into this 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) provides 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, subpart E refers 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 included in areas such as having a fire protection program, an emergency plan, training programs, and procedures to minimize contamination.

The most significant change for MLs in part 53 as compared to MLs under part 52 relates to Sec. 53.620(d) in subpart E and the associated licensing provisions in subpart H. These provisions 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 or OL. The first requirement in Sec. 53.620(d) establishes limitations on when a license under part 70 would authorize the loading of fuel into a reactor manufactured under an ML. The regulation requires the manufactured reactor to be configured during its loading, storage, and transport with features to prevent criticality and that those features be specified in the ML. The requirement provides flexibility because of the potential variety of reactor designs, the variety of possible measures to prevent criticality, and the range of possible conditions associated with the loading, storage, and transport of manufactured reactors. For example, the features to prevent criticality that could be considered individually and collectively to address possible adverse conditions include the reactivity control systems in place to support operations, inherent features of the fuel and materials within a manufactured reactor, and temporary measures or physical mechanisms (e.g., neutron poisons) for specific circumstances and conditions, such as during transport. This requirement contributes to the NRC's longstanding practice of requiring defense in depth for preventing accidents in any facility dealing with SNM, including requirements in Sec. 70.64 for certain part 70 licensees to adhere to the “double contingency principle.”

The requirements to have in place features 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 features to prevent criticality in the part 53 requirements will reasonably ensure that a manufactured reactor does not become critical over a range of possible conditions. With the requirements for features to prevent criticality under part 53 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, these measures will reasonably ensure that the manufactured reactor is not capable of operations, thereby obviating the need for a COL under Sec. Sec. 53.1416 and 53.1440 to authorize fuel loading. Additionally, this approach focuses 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, will be regulated primarily under the part 70 license. The reference to the requirements in subpart H of part 70 in Sec. 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 10 CFR part 51, “Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions,” 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.

Section 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) includes 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 Sec. 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 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, Sec. 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 physical security program for fueled manufactured reactors requires 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 requirements 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 necessary for physical security, radiation monitoring, or criticality prevention.

The regulations in part 53 covering the activities related to the storage, movement, and loading of fresh fuel into a manufactured reactor in the manufacturing facility likewise refer to the applicable regulations in part 70. Section 53.620(d) also requires 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 final 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) only allows the transport or removal of a manufactured reactor or portions of a manufactured reactor for either (1) delivery to a domestic site for which the Commission has issued a COL or CP authorizing the construction of a commercial nuclear plant using a manufactured reactor under the specific ML, or (2) export in accordance with 10 CFR part 110, “Export and Import of Nuclear Equipment and Material.” This requirement is similar to the limitations in Sec. 52.153. An additional paragraph in Sec. 53.620(e) provides 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. As noted previously, Sec. 53.620(e) includes an additional provision that allows a manufactured reactor or portions of a manufactured reactor to be removed from the place of manufacture for export in accordance with part 110, which represents another difference from the similar provision in Sec. 52.153.

Section 53.620(f) includes requirements for the acceptance and installation of a manufactured reactor at the site of a commercial nuclear plant. The requirements reference the construction requirements in Sec. 53.610 to govern the integration of the manufactured reactor into the construction of a commercial nuclear plant. Other requirements in the section address required receipt inspections and verification that interface requirements between the manufactured reactor and the balance of the commercial nuclear plant have been met.