# Licensing Requirements for Microreactors and Other Reactors With Comparable Risk Profiles > Briefs, arguments, decisions, and more. URL: https://www.frixlaw.com/law-library/documents/fr%3A2026-08550 ## Record - **Collection:** Federal Register - **Document type:** Proposed Rule - **Published:** May 1, 2026 - **Citation:** 91 FR 23628 ## Text NUCLEAR REGULATORY COMMISSION 10 CFR Parts 1, 2, 10, 11, 19, 20, 21, 25, 26, 30, 40, 50, 51, 57, 70, 72, 73, 74, 75, 95, 140, 150 [NRC-2025-0379] RIN 3150-AL36 Licensing Requirements for Microreactors and Other Reactors With Comparable Risk Profiles AGENCY: Nuclear Regulatory Commission. ACTION: Proposed rule; guidance; and request for comment. SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is proposing to amend its regulations to establish a risk-informed and performance-based regulatory framework for rapid licensing of new microreactors and other reactors with comparable risk profiles and for high-volume deployment of these reactors. The proposed rule would provide a flexible set of licensing pathways, reduce regulatory burden, and ensure that safety and security requirements remain commensurate with the potential hazards posed by these facilities. DATES: Comments must be submitted electronically using https://www.regulations.gov by 11:59 p.m. eastern time on June 15, 2026. ADDRESSES: Submit your comments, identified by Docket ID NRC-2025-0379, at https://www.regulations.gov. If your material cannot be submitted using https://www.regulations.gov, call or email the individuals listed in the FOR FURTHER INFORMATION CONTACT section of this document for alternate instructions. Do not include any personally identifiable information (such as name, address, or other contact information) or confidential business information that you do not want publicly disclosed. All comments are public records; they are publicly displayed exactly as received, and will not be deleted, modified, or redacted. Comments may be submitted anonymously. Follow the search instructions on https://www.regulations.gov to view public comments. You can read a plain language description of this proposed rule at https://www.regulations.gov/docket/NRC-2025-0379. For additional direction on obtaining information and submitting comments, see “Obtaining Information and Submitting Comments” in the SUPPLEMENTARY INFORMATION section of this document. FOR FURTHER INFORMATION CONTACT: George Tartal, Office of Nuclear Material Safety and Safeguards, telephone: 301-415-0016, email: George.Tartal@nrc.gov; Elijah Dickson, Office of Nuclear Reactor Regulation, telephone: 301-415-7647, email: Elijah.Dickson@nrc.gov; Michael Balazik, Office of Nuclear Reactor Regulation, telephone: 301-415-2856, email: Michael.Balazik@nrc.gov; and William Kennedy, telephone: 301-415-2313, email: William.Kennedy@nrc.gov. All are staff of the U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. SUPPLEMENTARY INFORMATION: Executive Summary A. Need for the Regulatory Action The purpose of this rulemaking is to safely expedite the licensing process for microreactors and other reactors with comparable risk profiles. This effort is consistent with, and implements direction in, the Accelerating Deployment of Versatile, Advanced Nuclear for Clean Energy Act of 2024 (Pub. L. 118-67, 138 Stat. 1448) (ADVANCE Act), and Executive Order (E.O.) 14300, “Ordering the Reform of the Nuclear Regulatory Commission” (90 FR 22587; May 29, 2025). Section 208 of the ADVANCE Act requires the NRC to develop “risk-informed and performance-based strategies and guidance to license and regulate microreactors.” The ADVANCE Act mandates that these strategies be incorporated into the existing regulatory framework, the technology-inclusive regulatory framework to be established through the rulemaking required by section 103(a)(4) of the Nuclear Energy Innovation and Modernization Act (Pub. L. 115-439, 132 Stat. 5572) (NEIMA), or a pending or new rulemaking by July 2027. On January 20, 2025, the President declared a National Energy Emergency in E.O. 14156, “Declaring a National Energy Emergency” (90 FR 8433; January 29, 2025), and stressed the need for a reliable, diversified, and affordable supply of energy. The President also issued E.O. 14154 (90 FR 8353; January 29, 2025), titled, “Unleashing American Energy,” with an objective of unleashing “America's affordable and reliable energy and natural resources.” On May 23, 2025, the President issued E.O. 14300. Section 5(e) of that E.O. directs the NRC to revise its regulations to “[e]stablish a process for high-volume licensing of microreactors and modular reactors, including by allowing for standardized applications and approvals and by considering to what extent such reactors or components thereof should be regulated through general licenses.” That E.O. set February 23, 2026, as the deadline for issuing this proposed rule, and the final rule must be issued by November 23, 2026. In developing this proposed rule, the NRC considered whether to establish the rule's scope within the amended non-power production or utilization facility (NPUF) licensing framework set out in the NRC's final rule, “Non-Power Production or Utilization Facility License Renewal,” issued on December 30, 2024 (89 FR 106234). That NPUF rulemaking was primarily intended to revise and streamline the license renewal process for facilities such as research and test reactors and medical isotope production facilities and was not designed to serve as a comprehensive licensing pathway for the high-volume deployment of microreactors. However, many of the design features and siting characteristics of NPUFs are expected to closely align with those reactors within the scope of this rulemaking. NPUFs are commonly located at national laboratories, private ventures, and universities, situated in both sparsely and densely populated areas. They operate over a broad range of thermal powers—up to tens of megawatts—with large thermal capacities and fuel designed with inherent safety features that enhance their stability and safety. The NRC considered amending part 50, “Domestic Licensing of Production and Utilization Facilities,” or part 52, “Licenses, Certifications, and Approvals For Nuclear Power Plants,” of title 10 of the Code of Federal Regulations (10 CFR), to provide for high-volume licensing of microreactors and other reactors with comparable risk profiles. The NRC didn't pursue amending part 52 or implementing a combined license approach in this proposed rule because the requirements for inspections, tests, analyses, and acceptance criteria (ITAAC) were designed for light water reactors (LWRs) (required by the Atomic Energy Act of 1954, as amended (AEA)) and the associated hearing on ITAAC closure could extend the licensing timeline. The NRC didn't pursue amending part 50 because the regulations in part 50 for commercial reactors were designed for large LWRs. The NRC also considered developing this proposed rule's scope within the framework of 10 CFR part 53, “Risk-Informed, Technology-Inclusive Regulatory Framework for Commercial Nuclear Plants.” Although part 53 provides a pathway to support licensing of microreactors, part 53 is designed to also cover large, complex reactors. The NRC decided to create a new part in 10 CFR chapter I that would be focused on rapid and high-volume licensing of microreactors and other reactors with comparable risk profiles. Therefore, the NRC developed a separate rulemaking that combines elements of the Commission's NPUF licensing approach in 10 CFR part 50 with elements from 10 CFR parts 52 and 53 to create proposed part 57, “Licensing Requirements for Microreactors and Other Reactors with Comparable Risk Profiles.” This proposed rule's framework would support rapid licensing of first-of-a-kind microreactors and other reactors with comparable risk profiles and high-volume deployment of these reactors through multiple licensing pathways, including the option for a general license to construct parts of these facilities. Collectively, the NRC's regulatory frameworks offer optionality and enable applicants to select licensing pathways that align with applicant-specific circumstances and deployment strategies. B. Major Provisions The primary provisions of this proposed rule would establish a risk-informed and performance-based regulatory framework for rapid and high-volume licensing of microreactors and reactors with comparable risk profiles. The proposed rule would provide flexible licensing pathways with streamlined requirements, as compared to the analogous requirements in part 50 and part 52, that would ensure safety and security requirements remain commensurate with the potential hazards posed by these facilities. Licensing and approval pathways would include a construction permit (CP) and an operating license (OL), a manufacturing license, a standard design approval, and provisions for affording regulatory finality to nuclear plant designs and essentially complete standardized operational programs. Applicants could combine in a single application requests for these licenses and approvals with requests for other licenses, approvals, and certifications for special nuclear material, byproduct material, transportation, and irradiated fuel storage to enable a broad spectrum of deployment models. The proposed rule is intended to expedite licensing reviews based on the statutory requirements of the AEA. E.O. 14300 directs the NRC to reach a final decision on an application to construct and operate a new reactor of any type within 18 months. This proposed licensing process should enable the NRC to issue an OL within 6-12 months after accepting an application, assuming that several factors beyond the NRC's control are met ( e.g., the application contains adequate information to allow the NRC to immediately docket the application and does not require the NRC to issue requests for additional information, the licensee completes timely construction, and any hearing contentions are expeditiously resolved). For a joint application for a CP and associated OL(s), the applicant would be required to submit final design information and complete operational programs at the time of application. The NRC would conduct a single, comprehensive safety review and potentially hold one adjudicatory hearing on the joint application. The Advisory Committee on Reactor Safeguards would review each joint application, focusing on aspects of the design that are unique, novel, and noteworthy. This proposed licensing framework would contain performance-based and risk-informed entry criteria consistent with design attributes that are necessary and essential for rapid, high-volume licensing of microreactors and other reactors with comparable risk profiles. Flexibilities in the proposed rule would include allowing a graded site characterization approach using existing site characterization data from Federal, State, or other organizations, provided that the data meets applicable NRC quality standards. Also, applicants would be able to define certain regulatory terms ( e.g., “basic component” and “safety-related”) and to limit the definition of “construction” to safety-related structures, systems, and components (SSCs), as defined in the proposed rule, or SSCs that would be relied upon to implement the proposed security requirements. The proposed rule would provide applicants with other flexibilities. Applicants could propose and justify an appropriate use of codes and standards as well as quality assurance programs tailored to the safety significance of the facility's SSCs. For environmental reviews, the proposed rule would permit the use of categorical exclusions under the National Environmental Policy Act, provided that specific conditions are met. The proposed rule would provide a general license for certain construction activities before issuance of a CP for an “nth-of-a-kind” facility ( i.e., a nuclear reactor or nuclear plant of a design that the NRC has already approved in a licensing proceeding) if certain conditions are met. The proposed rule would also provide alternative fitness-for-duty requirements for these licenses, as well as require the development of a cybersecurity program using a consequence-based approach. C. Costs and Benefits The NRC prepared a draft regulatory analysis to determine the expected quantitative costs and benefits of this proposed rule and associated guidance as well as qualitative factors to be considered in the NRC's rulemaking decision. The conclusion from the analysis is that this proposed rule and associated guidance would result in net averted costs to the industry and the NRC of approximately $3.76 billion using a 7-percent discount rate and $11.84 billion using a 3-percent discount rate. As the number of applicants increases, so do the estimated averted costs. The draft regulatory analysis also considers qualitative factors, such as greater regulatory stability, predictability, and clarity to the licensing process. Another qualitative factor is promoting a performance-based regulatory framework that specifies requirements to be met and provides flexibility to an applicant or licensee regarding the information or approach needed to satisfy those requirements. For more information, please see the draft regulatory analysis (available in the NRC's Agencywide Documents Access and Management System (ADAMS) Accession No. ML26111A076). Table of Contents I. Obtaining Information and Submitting Comments A. Obtaining Information B. Submitting Comments II. Executive Order 14300: Ordering the Reform of the Nuclear Regulatory Commission III. Background A. Characteristics of Microreactors and Other Reactors With Comparable Risk Profiles B. Public Interest in Microreactors and Other Reactors With Comparable Risk Profiles IV. Discussion A. Need for an Alternative Regulatory Framework B. Description of Proposed Licensing Framework C. Utilization Facilities and General Licenses V. Part 57 Framework A. Discussion of Provisions in Proposed Part 57 B. Subpart A—General Provisions C. Subpart B—Eligibility D. Subpart C—Construction Permits and Operating Licenses E. Subpart D—Manufacturing Licenses F. Subpart E—Standard Design Approvals G. Subpart F—Reporting of Defects and Noncompliance H. Subpart G—Irradiated Fuel Storage, Decommissioning, and License Termination Requirements I. Subpart H—Maintaining and Revising Licensing Basis Information J. Subpart I—Transportation Package Design Certification K. Subpart J—Physical Security Requirements L. Subpart K—Categorical Exclusion M. Subpart L—Inspections N. Subpart M—Material Control and Accounting O. Subpart N—[Reserved] P. Subpart O—Enforcement Q. Subpart P—Operator Licensing and Human Factors R. Subpart Q—Reporting and Other Administrative Requirements VI. Changes to Other Parts of 10 CFR Chapter I A. Conforming Changes to 10 CFR Parts 1, 2, 10, 11, 19, 20, 21, 25, 26, 30, 40, 50, 51, 70, 72, 73, 74, 75, 95, and 150 B. 10 CFR Part 26 C. 10 CFR Part 73 D. 10 CFR Part 140 VII. Specific Requests for Comments VIII. Regulatory Flexibility Certification IX. Regulatory Analysis X. Backfitting and Issue Finality XI. Cumulative Effects of Regulation XII. Plain Writing XIII. Environmental Assessment and Proposed Finding of No Significant Environmental Impact A. Introduction B. Conforming Changes C. Environmental Impacts of the Proposed Action D. Environmental Impacts of the Alternative to the Proposed Agency Action E. Agencies and Persons Consulted F. Proposed Finding of No Significant Environmental Impacts G. Stakeholder Interactions H. Environmental Assessment References XIV. Paperwork Reduction Act XV. 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 E. Executive Order 14294: Fighting Overcriminalization in Federal Regulations XVI. Voluntary Consensus Standards XVII. Availability of Guidance XVIII. Public Meeting XIX. Availability of Documents I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2025-0379 when contacting the NRC about the availability of information for this action. You may obtain publicly available information related to this action by any of the following methods: • Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2025-0379. • NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly available documents online in the ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, at 301-415-4737, or by email to PDR.Resource@nrc.gov. For the convenience of the reader, instructions about obtaining materials referenced in this document are provided in the “Availability of Documents” section. • NRC's PDR: The PDR, where you may examine and order copies of publicly available documents, is open by appointment. To make an appointment to visit the PDR, please send an email to PDR.Resource@nrc.gov or call 1-800-397-4209 or 301-415-4737, between 8 a.m. and 4 p.m. eastern time, Monday through Friday, except Federal holidays. • Public Meeting: The NRC may conduct a public meeting to describe the proposed amendments and answer questions from the public on the proposed rule. If the NRC determines it will hold a public meeting, NRC will publish a notice of the location, time, and agenda of the meeting on the NRC's public meeting website within 10 calendar days of the meeting. Stakeholders should monitor the NRC's public meeting website for information about the public meeting at: https://www.nrc.gov/public-involve/public-meetings/index.cfm. B. Submitting Comments Comments must be submitted electronically using https://www.regulations.gov by 11:59 p.m. eastern time on June 15, 2026. Please include Docket ID NRC-2025-0379 in your comment submission. The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at https://www.regulations.gov as well as enter the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information. If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment into ADAMS. II. Executive Order 14300: Ordering the Reform of the Nuclear Regulatory Commission On May 23, 2025, President Donald J. Trump signed Executive Order (E.O.) 14300, “Ordering the Reform of the Nuclear Regulatory Commission.” Section 5, “Reforming and Modernizing the NRC's Regulations,” requires the NRC to undertake a review and wholesale revision of its regulations and guidance documents as guided by the policies set forth in section 2 of the E.O. This rulemaking addresses section 5(e), which requires the NRC to “[e]stablish a process for high-volume licensing of microreactors and modular reactors, including by allowing for standardized applications and approvals and by considering to what extent such reactors or components thereof should be regulated through general licenses.” III. Background A. Characteristics of Microreactors and Other Reactors With Comparable Risk Profiles The microreactors and other reactors with comparable risk profiles that would be licensed under this proposed rule would be commercial nuclear reactors under section 103, “Commercial Licenses,” of the Atomic Energy Act of 1954, as amended (AEA). Due to their expected small sizes, low power levels, potential mobility, and simplicity of operation compared to the current fleet of operating power reactors, microreactors and other reactors with comparable risk profiles may be useful, for example, for remote communities, non-electric industrial processes, military bases, maritime applications, disaster relief, and other applications where a grid connection is unreliable or nonexistent. Microreactors and other reactor concepts with comparable risk profiles encompass a wide variety of reactor designs, including fuel forms, coolant types, and power levels. These concepts often incorporate inherent and passive safety design features that distinguish them from the large light water reactors in the current operating fleet. Fuel forms vary widely, from traditional light water reactor fuel assemblies to advanced fuels such as tri-structural isotropic (TRISO) particles, metallic fuels, and liquid fuels. Coolants include water, liquid metals ( e.g., sodium, lead), inert gases ( e.g., helium), and various molten salts. Power outputs range from only a few kilowatts to several tens of megawatts, and designs may operate in either a fast or thermal neutron spectrum. These diverse technical approaches reflect the industry's pursuit of reactor systems optimized for specific missions, operational environments, and market applications. Based on input from stakeholders (see section III.B, “Public Interest in Microreactors and Other Reactors with Comparable Risk Profiles,” of this document), the NRC anticipates that microreactors and other reactors with comparable risk profiles would rely heavily on standardization of design features and mass production to simplify licensing and deployment. Some reactors may be “self-contained” in that they would incorporate the reactor, shielding, and balance of plant in one or several transportable containers and require minimal site preparation or construction activities at the deployment site. Other designs may consist of a nuclear reactor that would be fabricated in a manufacturing facility and then incorporated into or connected to the permanent structures and systems of a nuclear plant constructed at the deployment site, such as a reactor building and power conversion equipment. The NRC understands that deployment models for microreactors and other reactors with comparable risk profiles would include various activities involving NRC licensing, certification, or approval. These activities may include designing reactors, manufacturing at a manufacturing facility, loading fuel at a manufacturing facility, operating the reactors for testing at a manufacturing facility, transporting fueled reactors to deployment sites (loaded with unirradiated or irradiated fuel), operating the reactors for the production of electrical or heat energy at the deployment sites, replacing reactors at the deployment sites, transporting reactors away from the deployment sites at the end of their useful lives, decommissioning or refurbishing and refueling reactors at locations away from the deployment sites, and re-deploying refurbished reactors to deployment sites. Some microreactors and other reactors with comparable risk profiles may also use more “traditional” approaches, including constructing the reactor in its entirety, loading fuel, or performing operational testing at the deployment site. This proposed rule would provide processes and requirements that would enable all these potential deployment models. B. Public Interest in Microreactors and Other Reactors With Comparable Risk Profiles The NRC recognizes the public interest in the development and deployment of microreactors and other reactors with comparable risk profiles. For several years, the NRC has conducted advanced reactor stakeholder meetings to facilitate open communication between the agency, industry, and the public regarding regulatory policy, licensing pathways, and technical issues related to advanced reactors. These meetings covered a wide range of topics, including safety and security considerations, fuel qualification and transportation, siting and environmental review, emergency preparedness, quality assurance approaches, risk-informed and performance-based regulatory methods, and lessons learned from the licensing of non-power production or utilization facilities (NPUFs). Stakeholders have also discussed and presented strategies for streamlining licensing processes to accommodate the anticipated high licensing volumes associated with modular and transportable reactor concepts. In addition to these public meetings, the NRC has received letters and formal reports from a broad spectrum of interested parties, including non-governmental organizations, policy organizations representing both the nuclear industry and public interest groups, national laboratories, and Federal, State, and local governmental entities. These submissions have provided perspectives on technical design features, operational considerations, safety analysis methodologies, environmental impacts, workforce development, and policy objectives for advanced reactor deployment. Many communications have highlighted the potential for microreactors to support energy resilience, remote power applications, industrial process heat, and national security missions. A recurring theme in both the stakeholder discussions and the written correspondence has been the need for the NRC to develop a clear, predictable, and efficient regulatory framework that supports rapid licensing of new microreactors and other reactors with comparable risk profiles and high-volume deployment of these reactors. Several stakeholders emphasized that when a microreactor applicant demonstrates low radiological consequences at the site boundary in the unlikely event of an accident, the NRC should allow the use of a licensing approach similar to that established for NPUFs. Stakeholders have noted that such an approach—appropriately adapted for microreactors—would leverage proven regulatory structures, align safety requirements with actual risk, and reduce unnecessary regulatory burden while maintaining the NRC's safety and security standards. IV. Discussion A. Need for an Alternative Regulatory Framework Rapid and high-volume deployment of microreactors and modular reactors is needed to support national policy and market demand. The Nuclear Energy Innovation and Modernization Act seeks to streamline licensing and reduce regulatory uncertainty for advanced reactor designs. The Accelerating Deployment of Versatile, Advanced Nuclear of Clean Energy Act requires the NRC to develop “risk-informed and performance-based strategies and guidance to license and regulate microreactors.” Executive Orders promote the development of domestic energy supplies to meet the increasing demand for electricity and direct the NRC to conduct this rulemaking. Market demand for baseload power has resulted in business cases for high-volume deployment of microreactors and modular reactors in markets where traditional large-scale nuclear power plants are impractical or uneconomical. This proposed rule is needed to establish a regulatory framework specifically tailored to rapid licensing of first-of-a-kind microreactors and other reactors with comparable risk profiles and high-volume deployment of these reactors. The use cases for such reactors support energy resilience, remote power applications, and industrial process heat. The proposed framework would be based on simplified safety requirements and would maximize the benefits of standardization. The proposed processes and requirements in this rule would enable shorter licensing timeframes that require fewer resources than those supported by existing regulations for nuclear power reactors in part 50 and part 52, which were designed for stationary, large light water reactors (LWRs). This proposed alternative regulatory framework is also needed to address Presidential and Congressional direction and stakeholder feedback. B. Description of Proposed Licensing Framework This proposed rule is complementary to and shares several features with part 53, “Risk-Informed, Technology-Inclusive Regulatory Framework for Commercial Nuclear Plants.” The part 53 rule features a risk analysis approach that accommodates licensing all reactor technologies, including microreactors and large, complex reactors. To complement this broad scope approach, proposed part 57 would rely on streamlined safety requirements to focus on simpler license applications and rapid licensing reviews of new reactors with less complex designs and operational characteristics and low potential radiological consequences. The major provisions and features of this proposed part 57 rule include the following: 1. Rapid Licensing Through Streamlined and Focused Safety Requirements This proposed rule would provide a pathway to enable rapid licensing through streamlined and focused safety requirements, for microreactors and other reactors with comparable risk profiles. The proposed rule would leverage the simplified designs, limited nuclear inventory, and overall low risk profiles of these facilities to establish the necessary and sufficient regulatory requirements to provide for reasonable assurance of adequate protection. This approach would enable shorter licensing timeframes by streamlining the information needed to be prepared by applicants and reviewed by the NRC. The applicant would be required to submit final design information and complete operational programs in a joint application for a construction permit (CP) and associated operating licenses (OLs). The NRC would conduct a single, comprehensive safety review and potentially hold one adjudicatory hearing on the joint application. Time and resource savings would be achieved for qualifying “first-of-a-kind” and “nth-of-a-kind” designs without any adverse impact on safety and security. 2. High Volume Licensing This proposed rule would enable high volume licensing based on standardization of reactor designs and operational programs. An applicant would have the option to request a single CP and any number of OLs for any number of nuclear reactors of essentially the same design to be built at one or more specific sites or within designated large geographical areas. Multiple applicants for essentially the same design would have the option to reference common non-site-specific information, and the NRC could consolidate some aspects of the licensing proceedings. 3. Rapid Deployment This proposed rule would provide options for issuance of a CP to include approval of the final reactor design and operational programs, address siting and environmental requirements for large geographical areas or multiple specific sites, and satisfy requirements for mandatory and adjudicatory hearings if an applicant provided all necessary information in a joint application for a CP and associated OL(s). This could support licensing reactor operation within days of site selection for time-critical deployment, depending on the simplicity of onsite construction activities. 4. Multiple Licensing Pathways The proposed rule would provide several licensing options for applicants to choose from to meet their deployment model or business case needs, including a joint application for a CP and associated OL(s), which would allow for deployment of reactors and approval of standard designs; a manufacturing license (ML), which would allow for approval and manufacture of standardized designs and approval of operational programs; and a standard design approval (SDA), which would allow for approval of entire reactor designs or major portions thereof. Applicants would be able to combine requests for these types of licenses and approvals with requests for license(s), approvals, and certifications under other regulations in a single application to holistically address their deployment strategies. 5. Request for Generic Finality An applicant may include in its joint application for a CP and associated OL(s) a request for generic finality. Matters resolved in a proceeding on the application for issuance of the CP and associated OL(s) for which the applicant has requested and the Commission has granted generic finality would be considered resolved in proceedings on other joint applications under proposed part 57 that reference the approved CP or associated OL(s). For joint applications for “nth-of-a-kind” nuclear reactors and nuclear plants that reference CPs and associated OL(s) afforded generic finality, the scope of licensing proceedings would be reduced to site- and applicant-specific information. 6. Manufacturing License Provisions The proposed rule would include the use of features to prevent criticality to allow reactors to be fabricated, fueled, and tested at a manufacturing facility before being transported to an operating site. This proposed rule would also allow ML applicants to request and the NRC to afford finality to the entire nuclear plant design and operational programs, thereby reducing the scope of proceedings on joint application for a CP and associated OL(s) that reference the ML to site- and applicant-specific information. 7. Categorical Exclusions The proposed rule would permit the use of categorical exclusions from the requirement for the NRC to prepare an environmental assessment or environmental impact statement under the National Environmental Policy Act (NEPA), provided that specific conditions are met. 8. General Licensee for Construction This proposed rule would establish a general license under which an applicant that files a joint application for a CP and associated OL(s) for a “nth-of-a-kind facility” could begin construction activities before the issuance of a CP, provided that certain conditions are met. 9. Alternative to 10 CFR Part 100 Siting Requirements The proposed rule would allow a graded site characterization approach with use of existing site characterization data from Federal, State, or other organizations, provided that the data meets applicable NRC quality standards. 10. Applicant Defined Definitions The definitions of many terms in this proposed rule would be equivalent to the corresponding terms defined in §§ 21.3, 50.2, and 52.1, all entitled “Definitions,” and other NRC regulations. However, given the variety of microreactor and other reactor designs with comparable risk profiles, flexibility is proposed to allow applicants to redefine applicable definitions to support their specific design and licensing basis needs, provided that such redefinitions are justified and supported by the applicant's safety analysis. 11. Codes or Standards The proposed rule would allow applicants to propose, with adequate justification, the use of codes and standards appropriate for their reactor design and not incorporate by reference the specific codes and standards in 10 CFR 50.55a, “Codes and standards.” 12. Quality Assurance Program The proposed rule would not impose quality assurance requirements under the existing regulations in appendix B, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,” to 10 CFR part 50. Instead, the proposed rule would allow the applicant to choose an industry-approved quality assurance program, similar to the approach taken in American National Standards Institute/American National Standard ANSI/ANS-15.8-1995 (R2018), “Quality Assurance Program Requirements for Research Reactors.” 13. Operational Programs Information related to operational programs concerning facility operation could be standardized to facilitate fleet-wide deployment of a microreactor or other reactor with comparable risk profile. These standardized operational programs could be designed to be administered onsite or at a corporate or institutional level. Standard operational programs such as emergency preparedness and security plans would receive finality, to the extent practicable, for future applicants that reference those approvals. 14. Remote Monitoring, Remote Operation, and Autonomous Operation This proposed rule would include provisions for applicants to specify design features for monitoring and operating a nuclear reactor from outside the site boundary and for autonomous performance of operations and safety functions. The NRC has posed a question in this proposed rule to obtain stakeholder feedback on remote operations and autonomous operations. 15. Operator Licensing and Human Factors This proposed rule would adjust staffing, training, personnel qualifications, and human factors engineering requirements, and would include provisions for general licenses for reactor operators, to reflect the expectation that the role of operators would be reduced for microreactors and other facilities with comparable risk profiles as compared to the current fleet of large LWRs. 16. Flexible Processes for Changes This proposed rule includes provisions for ML holders and holders of OLs that reference reactors manufactured under MLs to combine applications for license amendments or to make changes to the facility as described in the final safety analysis report (FSAR) without an amendment. Under certain conditions, holders of OLs for manufactured reactors would be able to implement the same changes approved by amendment to an ML without requesting amendments to their OLs that reference the ML. This would eliminate duplication of applications for NRC review of changes to manufactured reactors, including changes that might be made for improving safety or operational reliability. 17. Readiness for Operation Finding This proposed rule would provide for the NRC to authorize reactor operation upon finding that reactor construction conforms to the approved design and license requirements instead of using inspections, tests, analyses, and acceptance criteria under 10 CFR part 52, which could delay this authorization. 18. Fitness-for-Duty Program Flexibility This proposed rule would allow an applicant to propose an FFD program of its own specification if operator action would not be required to maintain the reactor within the criterion of proposed § 57.25(a) or a credible operator or maintenance error could not result in exceeding that criterion. 19. Resident Inspectors The NRC does not anticipate stationing a full-time resident inspector at facilities licensed under this framework. Instead, this proposed rule would rely on targeted inspections and performance oversight. 20. Transportation The proposed rule would add a provision that allows for a risk methodology to be used for evaluating normal and/or accident conditions in the event that an applicant cannot meet the testing and performance requirements of 10 CFR part 71, “Packaging and Transportation of Radioactive Material.” 21. Decommissioning and License Termination The NRC is proposing the flexibility for applicants to develop decommissioning plans as part of the initial licensing process. This approach would offer greater flexibility, given the variety of design and operational strategies being considered. The proposed decommissioning framework primarily builds on the NPUF model while incorporating elements from the power reactor framework. This proposed rule consists of several major components, including a new part 57, revisions to 10 CFR parts 26, “Fitness for Duty Programs,” and 73, “Physical Protection of Plants and Materials,” and conforming changes throughout 10 CFR chapter I to refer to part 57 where appropriate. C. Utilization Facilities and General Licenses E.O. 14300 directed the NRC to consider regulating microreactors or their components through general licenses. Stakeholders also have expressed interest in the possibility of the NRC using general licenses for these reactors or redefining “utilization facility” to exclude some nuclear reactors from the licensing requirements in section 103 of the AEA. The NRC considered these potential alternative approaches for high-volume licensing and regulation of nuclear reactors or fleets of reactors in developing this proposed rule. The NRC proposes that using a general license for regulation of construction activities for certain structures, systems, and components of nuclear reactors or nuclear plants would be the most practicable approach under this proposed rule. The NRC considered whether it would be practicable to exclude certain reactors that would otherwise be licensed under proposed part 57 from the definition of “utilization facility” and regulate them under a different regulatory framework. The pertinent portions of the definition of “utilization facility” in section 11(cc) of the AEA are the following: “(1) any equipment or device, except an atomic weapon, determined by rule of the Commission to be capable of making use of special nuclear material in such quantity as to be of significance to the common defense and security, or in such manner as to affect the health and safety of the public . . .; or (2) any important component part especially designed for such equipment or device as determined by the Commission.” The AEA definition of a utilization facility allowed the Atomic Energy Commission (AEC), the NRC's predecessor, to determine by rulemaking which equipment or devices met the criteria for a utilization facility. By connecting the definition of a utilization facility to the quantity of special nuclear material involved and the manner the material is used, and that material's potential impact on the common defense and security and public health and safety, Congress ensured that the AEC's regulatory authority would encompass facilities whose operation involves radiological safety and security. The AEC promulgated a definition of “utilization facility” in 1956, now set forth at 10 CFR 50.2 and proposed for part 57, that was limited to “any nuclear reactor other than one designed or used primarily for the formation of plutonium or [uranium-233].” The AEC also defined “nuclear reactor” as an apparatus, other than an atomic weapon, designed or used to sustain nuclear fission in a self-supporting chain reaction. This definition, also part of this proposed rule, implements both criteria of the AEA's “utilization facility” definition. An apparatus designed or used to sustain nuclear fission in a self-supporting chain reaction meets the first criterion—capable of making use of special nuclear material (SNM) in such quantity as to be of significance to the common defense and security. Several current examples show that even a quantity of SNM less than what is required to support a self-sustaining fission reaction in a nuclear reactor is significant to the common defense and security. The U.S. Department of Energy Order 474.2A, “Nuclear Material Control and Accountability,” requires that quantities of uranium-235 or plutonium of 1 gram or larger are subject to that order and require material control and accounting and security programs. Additionally, the NRC defines a quantity of uranium-235 (contained in enriched uranium) in excess of 1 kilogram as being at least Category III material requiring material control and accounting and security requirements. Finally, the International Atomic Energy Agency's Nuclear Security Recommendation on Physical Protection of Nuclear Material and Nuclear Facilities states that a mass as small as 1 kilogram of uranium-235 (contained in enriched uranium) needs to be subject to physical security requirements. These examples are relevant to this proposed rule because all reactors that would be licensed under this proposed rule—each one an apparatus designed or used to sustain nuclear fission in a self-supporting chain reaction—would require more than these minimum amounts of SNM to operate. An apparatus designed or used to sustain nuclear fission in a self-supporting chain reaction also meets the second criterion in the AEA definition of utilization facility—capable of making use of SNM in such manner as to affect the health and safety of the public. Decades of reactor licensing, including research reactors with power levels ranging from a few watts to several tens of megawatts, have shown that the use of SNM for self-sustaining fission reactions is capable of affecting public health and safety. Direct radiation from fission reactions, the creation and potential release of radioactive byproducts, and improperly-controlled (or uncontrolled) self-sustaining fission reactions can all affect public health and safety. Improper control of a self-sustaining fission reaction can cause significant and potentially very rapid increases in radiation levels, temperatures, and pressures, which is why the NRC requires appropriate regulatory controls that are different than those for devices that use SNM in other manners, such as a subcritical assembly for physics experiments or a neutron source for providing the initial neutrons needed to safely start up a nuclear reactor. These other devices have not typically been considered utilization facilities. The NRC anticipates that any nuclear reactor that would be licensed under proposed part 57 to use SNM for self-sustaining fission reactions for commercial purposes would clearly require controls to provide reasonable assurance of adequate protection of public health and safety. The AEA definition of “utilization facility” requires that only the safety prong or security prong of the definition be met. The discussion of the safety and security prongs in this document suggests that any nuclear reactor would meet both prongs and constitute a utilization facility under the definition in the AEA, thereby warranting regulation by the NRC as such, consistent with the responsibilities and authorities conferred to the NRC by the AEA. The Commission has used its regulatory authority under sections 103 and 182(a) of the AEA to require technical specifications for utilization facilities to provide reasonable assurance of adequate protection of public health and safety. The NRC would continue to do so under this proposed rule. The NRC considered whether it would be practicable to use the authority provided to the Commission by section 109(a) of the AEA to “issue general licenses for domestic activities required to be licensed under section [101 of the AEA] if the Commission determines in writing that such general licensing will not constitute an unreasonable risk to the common defense and security.” The AEA limits this authority “to those utilization and production facilities which are so determined by the Commission pursuant to section [11(cc)(2)] of [the AEA].” Section 11(cc) of the AEA is the definition of utilization facility, and section 11(cc)(2) of the AEA is “any important component part especially designed for [a utilization facility as defined in section 11(cc)(1) of the AEA] as determined by the Commission.” Thus, the NRC can issue a general license for any important component part especially designed for a utilization facility. The Commission proposes to use this authority to issue a general license in proposed § 57.45(d) for construction activities, subject to conditions in proposed § 57.45(d)(1) through (6) that would ensure that the general license would only be for any important component part especially designed for a utilization facility, not constitute an unreasonable risk to the common defense and security, and provide for adequate protection of the health and safety of the public. The proposed general license would potentially enable shorter deployment timeframes and is described in detail in section V.D of this document. The NRC also considered whether it could include in proposed part 57 a general license for regulation of an entire utilization facility, meaning a utilization facility as defined in section 11(cc)(1) of the AEA. However, the AEA provides the NRC with the authority to issue general licenses only for utilization facilities as defined in section 11(cc)(2) of the AEA, meaning any important component part especially designed for an entire utilization facility. Therefore, in developing proposed part 57, the NRC did not consider general licensing of an entire utilization facility as viable under the current statutory structure. Instead, the proposed rule would include a licensing framework under section 103 of the AEA that would reduce the number of licensing actions, resources for their completion, and required NRC oversight associated with deployment of individual reactors or nuclear plants or fleets of such facilities, as described in section IV.B of this document. V. Part 57 Framework A. Discussion of Provisions in Proposed Part 57 Proposed part 57 is comprised of subparts A through Q. These subparts would provide performance criteria and would be organized to specify requirements to demonstrate compliance with those performance criteria throughout the major stages of the life cycle of microreactors and reactors with comparable risk profiles. The performance-based approach proposed in part 57 also would include regulatory requirements that would allow applicants to use a flexible and graded approach to the performance of safety functions based on the role of a particular structure, system, or component and limiting its impact on assessed radiological consequence to the public. Proposed subpart P of part 26 would be new and would be largely consistent with the fitness-for-duty (FFD) requirements in current subpart K, “FFD Programs for Construction,” of part 26 supplemented by select requirements from subparts A through I, N, and O of part 26. These requirements are designed to ensure program effectiveness, maintain protections afforded to individuals subject to the FFD program, and align with FFD program implementation by parts 50 and 52 licensees. The proposed requirements would not be entirely equivalent with requirements in current subpart K of part 26 because the latter only applies during construction of the nuclear plant, whereas proposed subpart P of part 26 would apply during construction and operation. Furthermore, proposed subpart P of part 26 would allow the use of a variety of biological specimens for drug testing as well as innovative technologies for drug and alcohol screening and testing that are not described or allowed by the requirements in subparts A through K, N, and O of part 26, except under limited conditions. Proposed part 57 would also include a technology-inclusive consequence-based approach for physical security and emergency preparedness for nuclear plants. The NRC used operating experience to propose additional regulatory flexibility for a part 57 licensee's implementation of security requirements. This proposed rule would also propose changes to part 73 for a technology-inclusive approach to cybersecurity. The proposed provisions for these operational programs are based on meeting the proposed entry criteria for part 57. In addition, this proposed rule would make conforming changes throughout 10 CFR chapter I, by adding “and part 57” or similar language where appropriate to account for the addition of the proposed part 57. B. Subpart A—General Provisions Subpart A would provide the general provisions applicable to all applicants and licensees under proposed part 57. Subpart A would include provisions on purpose, scope, definitions, written communications, deliberate misconduct, employee protections, completeness and accuracy of information, information collection requirements, exemptions, standards for review, jurisdictional limits, attacks and destructive acts, rights related to SNM, license suspension and rights of recapture, backfitting and issue finality, the Advisory Committee on Reactors Safeguards, combining licenses, and filing of applications. 1. Definitions in Proposed Part 57 This proposed rule would provide its own definitions section in proposed § 57.3, “Definitions.” The definitions of many terms in proposed § 57.3 would be equivalent to the corresponding terms defined in §§ 21.3, 50.2, 52.1, and other NRC regulations. However, given the variety of microreactor and other reactor designs with comparable risk profiles, proposed § 57.3 would provide flexibility by allowing applicants to redefine applicable definitions to support their specific design and licensing basis needs, provided that such redefinitions are justified and supported by the applicant's safety analysis. Definitions established by the application would not require an exemption from proposed part 57. The flexibility to provide new definitions would extend only to definitions defined in proposed part 57 and not to those terms defined by statute, such as “special nuclear material.” Specific proposed definitions are further explained in the following paragraphs. The NRC proposes to include a definition of “Autonomous operation” in part 57 that would provide the means for applicants to present information regarding the performance of operational and safety functions without reliance on human intervention, external command, or active control system input under normal operations and accident conditions. The design of the microreactor with inherent safety features and active structures, systems, and components (SSCs) would govern what design functions need to be executed and/or monitored during normal, off-normal and accident conditions. The proposed definition of “Certified fuel handler” would mean a non-licensed operator who is responsible for decisions on the safe conduct of decommissioning activities, safe handling and storage of spent fuel as defined in 10 CFR 72.3, “Definitions,” and appropriate response to plant emergencies. The certified fuel handler would need to be qualified in accordance with a fuel handler training program that meets the same requirements as training programs for non-licensed operators required by proposed § 57.420, “Training and qualification for non-licensed personnel.” The proposed definition of “Consensus code or standard” would be based on the use of these terms in the National Technology Transfer and Advancement Act of 1995 (NTTAA) (Pub. L. 104-113) and the Office of Management and Budget (OMB) Circular No. A-119, “Federal Participation in the Development and Use of Voluntary Consensus Standards and in Conformity Assessment Activities.” As required by NTTAA, the NRC undertakes the following activities: (i) consults with voluntary consensus standards bodies; (ii) participates with voluntary consensus bodies in the development of consensus standards; and (iii) uses consensus standards to carry out the NRC's policy objectives. The proposed definition of “Construction” is slightly different than the current definition in existing § 50.10, “License required; limited work authorization.” The proposed definition would differ from the current § 50.10 definition in that it would apply to only safety-related SSCs (as defined in proposed part 57) and SSCs relied upon to implement the proposed security requirements. The proposed definition of “Control room” would provide a means for remote monitoring and/or remote operation outside the site boundary where actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions. The proposed definition of “Decommission” would be slightly different than the definition in § 50.2. The proposed definition would also include permanent removal of an individually licensed nuclear reactor. The proposed definition of “Defense in depth” would provide a philosophy of designing a nuclear facility that includes two or more independent and redundant layers of defense in the design of a facility and its operating procedures to compensate for uncertainties such that no single layer of defense, no matter how robust, is exclusively relied upon. Defense in depth includes, but is not limited to, the use of access controls, physical barriers, redundant and diverse safety functions, and emergency response measures. The proposed definition of “Design bases” would be the information that identifies the specific functions to be performed by an SSC of a facility, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design. These values may be (1) restraints derived from generally accepted “state-of-the-art” practices for achieving functional goals, or (2) requirements derived from analysis (based on calculation and/or experiments) of the effects of a postulated accident for which an SSC must meet its functional goals. The proposed definition of “Design features” would be the active and passive SSCs and inherent characteristics of those SSCs that contribute to limiting the total effective dose equivalent (TEDE) to individual members of the public during normal operations and prevent or mitigate the consequences of design basis accidents. The proposed definition of “Fission product release” would be the amount and composition of radioactive material released to the environment, after accounting for any retention of radionuclides provided by reactor design features. The proposed definition of “Fuel” would be SNM or source material, discrete elements that physically contain SNM or source material, and homogeneous mixtures that contain SNM or source material, intended to or used to create power in a nuclear reactor. The proposed definition of “Licensing basis information” would be the information contained in regulations, orders, licenses, certifications, or approvals issued by the NRC for a nuclear plant licensed under proposed part 57 and that information submitted to the NRC by an applicant or licensee in a safety analysis report, program description, or other licensing-related document required under proposed part 57. The proposed definition of “Manufactured reactor” would be the essential portions of a nuclear reactor that are manufactured under an ML and subsequently incorporated into a nuclear plant under a construction permit issued under subpart C of proposed part 57. The proposed definition of “Manufacturing license” would be a license issued under subpart D of proposed part 57 that authorizes the production of manufactured reactors but not their construction, installation, or operation. The proposed definition of “Programmatic controls and operational programs” would be administrative procedures that govern human action in implementing programs and operating, monitoring, and maintaining SSCs and equipment of a nuclear plant. Programmatic controls could be standardized to facilitate fleet-wide deployment of a microreactor. These standardized operational programs could be designed to be administered on site or at a corporate or institutional level. Implementation milestones for each operational program would need to be described depending on whether the program will be implemented all at once or on a phased basis. The proposed definition of “Quality assurance” (QA) would be planned and systematic actions during design, construction, and modification necessary to provide adequate confidence that the SSC will perform satisfactorily in service. The proposed definition of “Remote monitoring” would mean observing plant data from a location outside of the site boundary. Remote monitoring does not include the performance of any operator actions necessary to manipulate the reactor to protect the public health and safety ( i.e., remote operations). However, remote monitoring could be used to access real-time data needed to perform other functions that protect the public health and safety, such as emergency preparedness or security. The ability to protect the public would be dependent upon having accurate and timely access to the plant-monitored parameter data. Wireless communication could be used to support remote monitoring. The proposed definition of “Remote operation” would be to command and control the reactor from a location outside of the site boundary. Industry has indicated that the design of a microreactor with inherent safety features and active SSCs would govern what design functions need to be executed and/or monitored during normal, off-normal, and accident conditions. The proposed definition of “Safe shutdown” would be bringing the nuclear reactor to safe, stable conditions specified in plant technical specifications when the reactor is under design basis accident conditions with loss of emergency power and offsite power. The proposed definition of “Safety function” would be the purpose served by a design feature, human action, or programmatic control to prevent or mitigate unplanned events and thereby demonstrate compliance with requirements in proposed part 57 for limiting risks to public health and safety. Safety functions could be performed by any combination of the elements supported by the safety analysis and could be specified at the plant level or at the level of a particular barrier or system. Multiple plant-level safety functions would be assumed to apply to all reactor designs based on established requirements and historical practices. These fundamental safety functions would include the control of reactivity, removal of heat, and limiting the release of radioactive materials. The protection of a specific barrier or system that contributes to meeting plant-level safety criteria could also be referred to as a safety function. The proposed definition of “Safety-related structures, systems and components” is slightly different than the definition in § 50.2. Whereas the § 50.2 definition refers to “events,” the proposed definition would refer to “accidents.” Design basis accidents bound events. Also, where the § 50.2 definition refers to a reactor coolant pressure boundary, the proposed definition would be technology neutral because some reactor designs under proposed part 57 may not operate at pressure. The proposed definition of “Source term” would be the magnitude and mix of the radionuclides released from the fuel, expressed as fractions of the fission product inventory in the fuel, as well as their physical and chemical form, and the timing of their release. The source term would be developed by the applicant when performing the maximum hypothetical accident (MHA) or maximum credible accident (MCA) methodology. This source term would then be analyzed with site parameter information to demonstrate compliance with the accident dose-based entry criterion in proposed § 57.25(a). The proposed definition of “Special nuclear material” would be (1) plutonium, uranium-233, uranium enriched in the isotope-233 or in the isotope-235, and any other material that the Commission, pursuant to the provisions of section 51 of the AEA, determines to be SNM, but does not include source material; or (2) any material artificially enriched by any of the foregoing, but does not include source material. 2. Other General Provisions Proposed § 57.4, “Written communications,” would govern written communications and how applications and other required information must be submitted to the NRC. These requirements would be equivalent to those in § 50.4, “Written communications.” Proposed § 57.5, “Deliberate misconduct,” would establish requirements for enforcement action to which a licensee, an applicant, or a licensee's or applicant's contractor or subcontractor, or an employee of any of them, may be subject for engaging in deliberate misconduct. These requirements would be equivalent to those in § 50.5, “Deliberate misconduct.” Proposed § 57.6, “Employee protection,” would prohibit discrimination against an employee of a holder or applicant for an NRC license, permit, or SDA, or a contractor or subcontractor of a holder or applicant for an NRC license, permit, or SDA for engaging in certain protected activities. Proposed § 57.6 also would prescribe a procedure for seeking a remedy for employees who believe they have been discriminated against for engaging in such protected activities. These requirements would be equivalent to those in §§ 50.7 and 52.5, both entitled “Employee protection.” Proposed § 57.7, “Completeness and accuracy of information,” would govern the completeness and accuracy of information provided to the NRC. These requirements would be equivalent to those in §§ 50.9 and 52.6, both entitled “Completeness and accuracy of information.” Proposed § 57.8, “Information collection requirements: OMB approval,” would establish requirements for information collection requirements and OMB approval. These requirements would be equivalent to those in § 50.8, “Information collection requirements: OMB approval.” Proposed § 57.9, “Specific exemptions,” would govern exemptions from the requirements of the regulations in proposed part 57. These requirements would be equivalent to those in §§ 50.12 and 52.7, both entitled “Specific exemptions.” Proposed § 57.11, “Jurisdictional limits,” would require that no license or SDA issued under proposed part 57 would cover activities that are not under or within the jurisdiction of the United States. These requirements would be equivalent to those in § 50.53, “Jurisdictional limitations.” Proposed § 57.12, “Attacks and destructive acts,” would state that licensees, holders of standard design approvals, and applicants for licenses and standard design approvals would not be required to provide design features or other measures for the specific purpose of protection against the effects of attacks and destructive acts by enemies of the United States directed against the facility or deployment of weapons incident to U.S. defense activities. These requirements would be equivalent to those in § 50.13, “Attacks and destructive acts by enemies of the United States; and defense activities.” Proposed § 57.13, “Rights related to special nuclear material,” would establish requirements for rights related to SNM. These requirements would be equivalent to those in § 50.54(b) and (c). Proposed § 57.14, “License suspension and rights of recapture,” would establish requirements for license suspension and rights of recapture of the material or control of the facility in a state of war or national emergency declared by Congress. These requirements would be equivalent to those in § 50.54(d). Proposed § 57.15, “Agreement limiting access to Classified Information,” would address requirements for agreements limiting access to classified information and would be equivalent to § 50.37, “Agreement limiting access to Classified Information.” Proposed § 57.16, “Backfitting and issue finality,” would address backfitting requirements by providing requirements that would be equivalent to those in § 50.109, “Backfitting,” and issue finality requirements by providing requirements that would be equivalent to those in §§ 52.83(a), 52.145, “Finality of standard design approvals; information requests,” and 52.171, “Finality of manufacturing licenses; information requests.” An exception is that proposed § 57.16(c) would not include an equivalent requirement to § 52.171(b)(2), which requires the Commission to determine that departures will comply with the requirements in § 52.7 and that the special circumstances for the departure would outweigh any decrease in safety that may result from the reduction in standardization caused by the departure. Proposed § 57.16(c) would instead require the joint application for the referencing CP and OL(s) to include analysis of departures from the design characteristics, site parameters, terms and conditions, or approved design of the nuclear reactor, nuclear plant, or manufactured reactor. Proposed § 57.16(c) would also specify that analysis would not be required for departures from any operational programs or requirements approved with the referenced CP, OL, or ML that are not material to the adequacy of the design, if the joint application includes proposed alternative operational programs or requirements. Under proposed § 57.16(c), all departures would be subject to litigation in the same manner as other issues in the CP or OL, which would be equivalent to § 52.171(b)(2). Proposed § 57.17, “Referral to the Advisory Committee on Reactor Safeguards (ACRS),” would address referral to the Advisory Committee on Reactor Safeguards (ACRS) and would be equivalent to §§ 50.58, “Hearings and report of the Advisory Committee on Reactor Safeguards,” 52.141, “Referral to the Advisory Committee on Reactor Safeguards (ACRS),” and 52.165, “Referral to the Advisory Committee on Reactor Safeguards (ACRS).” Proposed § 57.18, “Combining licenses; elimination of repetition; relationships between subparts,” would address combining applications and would be equivalent to §§ 50.31, “Combining applications,” 50.52, “Combining licenses,” and 52.8, “Combining licenses; elimination of repetition.” Proposed § 57.18 would also provide clarity about various combinations of licenses and contents of related applications that would enable various high-volume deployment strategies. While proposed part 57 clearly outlines the licensing framework for combining licenses for multiple reactors, multiple sites, manufacturing, possession of special nuclear material, and other deployment activities, this licensing framework largely exists under other parts of 10 CFR chapter I, such as parts 50, 52, and 53. Proposed § 57.18(a)(1) would include a provision for applications that would be filed under proposed part 57 by one or more applicants for licenses to construct and operate nuclear reactors or nuclear plants of essentially the same design to be located at different sites, to refer to a single FSAR. This proposed provision would be similar to the provisions in appendix N to part 50, “Standardization of Nuclear Power Plant Designs: Permits To Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites.” Proposed § 57.18(a)(2) would include a provision that an applicant may include in one application for a CP and associated OL(s) for a nuclear reactor or nuclear plant under proposed part 57 information for multiple sites at which the applicant proposes to construct and operate the reactor or plant. This proposed provision would allow for licensing construction and operation of a single nuclear reactor or nuclear plant at multiple locations over its lifetime, such as for operational testing at a manufacturing facility and power operation at a deployment site. Proposed § 57.18(a)(3) would require an application under proposed part 57 for multiple types of permits, licenses, or certifications to clearly indicate to which permit, license, or certification information in the application pertains. This proposed requirement would facilitate the NRC's review of the application by ensuring that the NRC would apply the appropriate proposed requirements ( e.g., standards of review, issuance, hearings, finality, etc.) to the information in the application. Proposed § 57.18(a)(4) would include provisions for holders of OLs that reference the same ML to combine among themselves, or with the holder of the ML, applications for license amendments under proposed § 57.310, “Amendment of license.” This proposed provision would potentially decrease the overall resources that would be required for applicants and the NRC for identical requests for amendments to multiple licenses as opposed to separate filings and reviews of each application for amendment. Proposed § 57.18(a)(5) would specify that an applicant may include in a single joint application a request for a CP for any number of nuclear reactors of essentially the same design that would be built at a specific site and requests for OLs for those reactors, provided that the application would state the earliest and latest dates for completion of the construction of each nuclear reactor as would be required by proposed § 57.55(g) and would include the information that would be specified in proposed § 57.60(a)(4). This proposed provision would potentially reduce applicant and NRC resources related to licensing a nuclear plant at which multiple nuclear reactors of essentially the same design would be operated over its lifetime, including replacement reactors. Proposed § 57.18(b), (d), and (e) would include provisions for incorporating by reference information contained in previous applications, statements, or reports filed with the Commission and applicable Commission approvals issued under part 50 or 52; referencing a standard design approval, CP, OL, ML, or combination thereof, that would be issued under proposed part 57; and referencing a relevant U.S. Department of War or U.S. Department of Energy authorization for a utilization facility that has been tested and that has demonstrated the ability to function safely, respectively. These provisions would allow applicants and the NRC to minimize duplication of previous efforts in filing and reviewing applications under proposed part 57. Proposed § 57.18(c) would continue the Commission's practice of combining multiple authorizations for a licensee under various parts of 10 CFR chapter I into one license based on the Commission's authority under section 161(h) of the AEA to combine NRC licenses. Proposed § 57.19, “Filing of application,” would address filing of applications and would be equivalent to §§ 50.30, “Filing of application; oath or affirmation,” 52.135, “Filing of applications,” and 52.155(a). Proposed § 57.19(f) would require an applicant for licenses to construct and operate one or more nuclear reactors under subpart C of proposed part 57 to file a joint application for a CP and associated OL(s). Proposed § 57.19(f) would also require that the joint application include the information specified in proposed §§ 57.55, “Content of applications; general information,” and 57.60, “Content of applications; technical information,” and be complete enough to permit all evaluations necessary for the issuance of the requested CP and the associated OL(s) upon the NRC making the finding required by proposed § 57.100(b)(1) ( i.e., the finding that construction has been substantially completed). The joint application would permit the NRC to use the regulations in § 2.105(c) to specify in the notice of proposed issuance of the CP that on completion of construction and the NRC making the finding that would be required by proposed § 57.100(b)(1), the associated OL(s) would be issued without further prior notice, thus streamlining the process for issuance of the associated OL(s) and reducing the timeframe for licensing. C. Subpart B—Eligibility The NRC based the development of the proposed part 57 framework on existing licensing practices for non-power and other utilization facilities that, by design and operational characteristics, present low risks of radiological consequences. These characteristics have designers approach safety by emphasizing accident prevention with inherent self-limiting reactivity feedback mechanisms and passive safety systems for heat and decay heat removal without reliance on complex active safety systems. The NRC used these characteristics to create a set of requirements to determine which applicants would be eligible to use proposed part 57. Located in proposed §§ 57.25, “Applicability,” and 57.30, “Design criteria attributes,” these proposed requirements are termed “entry criteria” and “design criteria attributes,” respectively. Given the wide range of reactor types and their functional characteristics, this proposed rule would emphasize the “attributes” of microreactors and other reactors with comparable risk profiles. Rather than defining these reactors in terms of thermal power level, this attribute-based approach would describe microreactors and other reactors with comparable risk profiles in terms of their functional characteristics, such as the capability to prevent or mitigate accidents without active systems or operator intervention. By doing so, the NRC recognizes that reactors with inherently safe design features and more favorable safety profiles may appropriately be designed with higher power levels than other reactor designs. The first eligibility criterion would be a dose-based acceptance value. The second eligibility criterion would be an upper limit on the amount of fuel. These eligibility criteria are intended to screen in reactor designs that are smaller, simpler, and more conducive to rapid, high-volume licensing. These eligibility criteria would be supported by six design criteria attributes. These design criteria attributes emphasize the features of inherently and passively safe reactors that make them secure and protective against radiological harm. These attributes include (1) reactivity control, (2) heat removal, (3) fission product retention, (4) shielding, (5) radioactive effluents control, (6) security by design. If an applicant for a reactor design does not meet these criteria, they can apply for a license under a different regulatory framework. 1. Dose-Based Entry Criterion A dose-based entry criterion under accident conditions would be used to inform the analysis of postulated accidents and the development of safety measures so that, in the unlikely event of an accident, there is assurance that no acute radiation-related harm will result to any member of the public. The Commission has found the use of a dose-based entry criterion to be adequate for facility siting and design purposes based on decades of extensive experience in the criterion's application and in recognition of the assumptions and considerations applied within the radiological consequence analyses. While the dose-based entry criterion would be computed in terms of dose, it is a figure of merit used to characterize the minimum requirements for design, fabrication, construction, testing, operational limits, and performance for safety-related SSCs. The numerical value of the criterion does not represent acceptable or actual public exposures received during normal and emergency conditions, which are primarily controlled by 10 CFR part 20, “Standards for Protection Against Radiation,” and through emergency planning. An applicant would be required to demonstrate their reactor design meets the 1 rem (10 millisieverts (mSv)) TEDE dose-based entry criterion in proposed § 57.25(a), and the NRC has found that the maximum hypothetical and maximum credible accident methodologies would be acceptable means of providing this demonstration. These methodologies are associated with a fission product release accompanying damage to fission product retention barriers, maximum allowable leak rates, a postulated single failure of any safety-related SSCs, conservative site meteorological dispersion characteristics, and an individual member of the public presumed to be at the location of maximum cumulative dose in the unrestricted area without protective actions. By demonstrating under these conservative assumptions that, in the unlikely event of an accident, the dose to the maximally exposed individual member of the public in the unrestricted area would remain below the accident dose acceptance criterion, there is reasonable assurance that actual accidents would not result in acute offsite doses. Historically, NRC licensing processes have relied on deterministic bounding analyses that, while conservative, may impose unnecessary siting, design, and operational constraints on advanced reactor designs with inherent and highly reliable passively safe reactor technologies. The Commission recognizes the need for flexibility in how applicants define their licensing basis to reflect the diversity of microreactors and other reactor designs with comparable risk profiles. Proposed part 57's inclusion of both the MHA and MCA methodologies provides risk-informed and performance-based regulatory pathways that align the applicant's safety analysis scope with the complexity and safety characteristics of their design. Proposed part 57 distinguishes between the MHA and the MCA with respect to the amount of analytical rigor necessary to justify the derived source term. By distinguishing between the MHA and MCA approaches, the Commission would allow applicants to tailor the scope and depth of their accident analyses to their design and business model needs while continuing to ensure safety. The source term defines the magnitude and mix of the radionuclides released from the fuel, expressed as fractions of the fission product inventory in the fuel, as well as their physical and chemical form, and the timing of their release. The applicant would utilize their MHA or MCA source term to establish the site boundary and determine the level of design, qualification, testing, and maintenance of SSCs necessary to show with reasonable assurance that the radiological consequences at the site boundary are below the 1 rem TEDE entry criterion of proposed § 57.25(a). Depending on the desired level of analysis, applicants may select either the MHA or MCA approach. The MHA approach can demonstrate safety through a postulated accident scenario, often highly conservative, which assumes a severe release of radioactive material consistent with physical laws, regardless of probability. This MHA analysis does not rely on detailed risk-informed assessment methodologies, thereby reducing analytical complexity for reactors with few to no active systems or self-limiting physical phenomena. The MHA approach may be desirable for applicants that are willing to accept additional conservatism by leveraging simplified analyses that are less time and resource intensive. Although the MHA may not necessarily reflect a realistic or credible sequence of events, it represents a bounding case to support subsequent safety decisions. If an applicant does not wish to accept the conservatisms associated with the MHA approach, further analyses would need to be performed to support an MCA approach. The MCA approach excludes certain physically unrealistic or excessively conservative assumptions, focusing instead on events that are credible given the technology, safety systems, and plant operating conditions. The MCA analysis can leverage a variety of modern risk-informed methodologies to credibly quantify events and consequences, providing a rational basis for a smaller site boundary and focused SSC categorization and potentially reducing the number of components subject to the more stringent safety requirements. Two identical reactor designs could, in principle, yield different site boundary distances and safety classifications depending on whether their analyses employ the MHA or MCA methodology. Under the MHA approach, conservative bounding assumptions, such as postulated worst-case system failures and maximum radionuclide release, would produce a larger source term necessitating a greater site boundary and broader safety classification of SSCs. In contrast, an MCA analysis that quantifies system performance and reliability could justify a smaller, more realistic source term and a correspondingly smaller site boundary and narrower safety classification. Both outcomes would be acceptable under proposed part 57's consequence-based framework because each would provide reasonable assurance that offsite radiological consequences remain below the 1 rem TEDE entry criterion. The preferred approach would likely depend on the scope and depth of analysis the applicant wishes to undertake. Applicants would need to be clear on which approach is being applied, and analyses would have to be supported by appropriate and sufficient technical justifications. The NRC is providing flexibility on how the TEDE dose-based entry criterion would be met in recognition of the need for expedited licensing and deployment of the types of facilities on which proposed part 57 is focused. Including both the MHA and MCA methodologies supports the Commission's regulatory modernization goals by encouraging innovation in reactor design while maintaining a consistent safety objective. Furthermore, this graded approach would enable efficient licensing reviews by aligning analytical rigor with risk significance without diminishing safety assurance. Under this proposed framework, applicants should discuss their plans for use of an MHA or MCA with the NRC staff prior to submittal of an application. This would ensure there is common understanding of the applicant's approach and would allow for resolution of any issues before development of a complete application. 2. Fuel Mass Limit The premise of this proposed rule is to establish regulatory requirements commensurate with the low hazards posed by facilities that would be licensed under proposed part 57. These requirements would be justified by the use of a dose-based entry criterion applied to the results of a maximum hypothetical or maximum credible accident that assesses siting and the performance of safety-related SSCs. This would also be true for large LWRs with a very large site boundary. However, many of the traditional requirements that the NRC considered when creating this proposed rule have historically provided defense in depth to address unlikely events that may exceed analyzed releases. Traditional requirements include the Commission's historical treatment of severe accidents based on lessons learned from operating large LWRs. Examples of these regulations include: 10 CFR 50.46, “Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors,” for assessing large-break loss of coolant accidents; 10 CFR 50.155, “Mitigation of beyond-design-basis events,” for flexible mitigation strategies for beyond-design-basis events; and several part 52 requirements for severe accident design features. The fuel mass limit entry criteria would deterministically screen reactor designs without additional performance-based acceptance criterion or severe accident analysis to assess events beyond which SSCs could be challenged. The fuel mass limit entry criteria would be established to provide additional defense in depth for these very unlikely events by limiting the amount of decay heat that may necessitate the need for active cooling systems and overall material available for release, further limiting the potential for causing acute health effects to the public. However, the NRC has proposed a question in this proposed rule, asking whether, in lieu of applying a deterministic material limit on the quantity of SNM, the NRC should apply an alternative performance-based acceptance criterion such as an adiabatic heat rate threshold, beyond which SSCs could be challenged. To assist in developing a quantitative basis for such a limit, the NRC reviewed and evaluated the quantities of SNM in the cores of several reactor types. In evaluating the quantities of SNM, the NRC determined the quantities of uranium (U) and plutonium (Pu). This includes the following isotopes: 1 U-233, U-234, U-235, U-236, U-238, Pu-236, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Pu-244. For technological neutrality, the mass criteria would also include thorium isotopes, because thorium can be used as a breeding material in thermal spectrum breeder reactors. None of the reactors considered in the evaluation included this technology, but there have been early indications of industry interest in pursuing this concept. 1 None of the evaluated non-LWRs included thorium, so they had negligible amounts of U-233. In conducting this evaluation, the NRC considered a spectrum of reactor technologies, including several non-LWR designs, two small modular pressurized water reactors (PWRs) and one small modular boiling water reactor (BWR), and several representative large LWRs. The purpose of this evaluation was to understand the similarities and differences between these reactor technologies and inform an entry criterion that facilitates high-volume licensing of microreactors. The assessment compared these reactor technologies, the SNM masses, type and kinds of engineered safety features, and accident response characteristics. To perform this evaluation, the NRC considered several sources of publicly available information covering a range of reactor types and power levels. The evaluation included several non-LWRs of various reactor types and fuel forms ( e.g., TRISO, metal, oxide, and molten salt) and coolants ( e.g., gas, molten salt, liquid metal, water). The power range of these designs spans from approximately 5 megawatts thermal (MW th ) to about 2250 MW th . The assessment also included small modular and large LWRs to gain a sense of the differences in SNM quantities between the non-LWR and small LWR designs currently in development versus the quantities in the currently operating large LWR commercial fleet. The power reactor range for the large LWRs spans from approximately 2600 MW th to about 4400 MW th . The quantities of SNM vary by reactor technology. For each reactor technology, the NRC calculated SNM quantities at the beginning and end of an operating cycle based on published core and fuel parameters and operational characteristics. To perform the calculation, the NRC utilized the Oak Ridge National Laboratory SCALE code system. The SCALE code system is a widely used modeling and simulation suite for nuclear safety analysis and design. Results of these calculations found that the large LWR SNM quantities at the beginning of an operating cycle ranged from approximately 71 metric tons heavy metal (MTHM) 2 for a PWR to 154 MTHM for a BWR. At the end of an operating cycle, these quantities range from approximately 69 to 148 MTHM, respectively. Except for a large molten salt reactor, which had an SNM quantity of approximately 43 MTHM, the remaining reactors at the beginning of an operating cycle had SNM quantities no greater than 9.3 MTHM and at the end of an operating cycle, or equilibrium, SNM quantities no greater than 8.7 MTHM. 2 MTHM is a unit used to define the mass of SNM where that material may include more than uranium ( i.e., when plutonium is included). One metric ton of heavy metal equates to 1000 kg of uranium, plutonium, or both. For a reactor containing entirely uranium fuel, 1 MTHM = 1 MTU. Table 1 compares various reactor types by the amount of SNM, in terms of MTHM, each contains by cycle period. Table 1 provides the reactor name, fuel type, percent fuel enrichment, and cycle period for which each of the SNM quantities were estimated as beginning of life (BOL), continuous refueling (cont.), equilibrium (equil.), beginning of equilibrium cycle (BOEC), and end of equilibrium cycle (EOEC). The BOL are conditions of the reactor core at initial startup after fresh fuel loading. The end of life (EOL) describes the conditions of the reactor core at the end of its useful fuel cycle, when fuel burnup or reactivity limits have been reached. Some reactor designs operate continuously. For continually refueled systems, SNM inventories are given as equilibrium conditions. For these designs, the BOEC is a state of the reactor core at the start of a cycle once equilibrium operating conditions have been established. Likewise, the EOEC is a state of the reactor core operating on a continuous refueling cycle at the end of a typical equilibrium operating cycle, after equilibrium burnup has occurred. Uranium dioxide (UO 2 ) is a ceramic oxide fuel made from uranium dioxide powder, pressed into pellets, and sintered for LWRs. TRISO fuel consists of spherical uranium kernels, usually of uranium dioxide or uranium oxycarbide, coated with multiple layers of pyrolytic carbon and silicon carbide, which act as a miniature containment system. Metallic alloy fuel in a compact form is composed of uranium (U), transuranics (TRU), and 10 weight percent (wt. %) zirconium (Zr) (U-TRU-10Zr Metal Fuel). Molten salt fuel is a liquid fuel salt mixture consisting of lithium fluoride (LiF), beryllium fluoride (BeF 2 ), and uranium tetrafluoride (UF 4 ) (LiF-BeF 2 -UF 4 ). BILLING CODE 7590-01-P EP01MY26.006 BILLING CODE 7590-01-C Reactor safety profiles vary significantly between technologies due to differences in fuel type, coolant, operating characteristics, and reliance on active versus intrinsic and passive safety systems. Traditional large LWRs have large inventories of SNM and operate at higher power levels, power densities, and operating pressures than the other reactors studied. These features present more complex accident scenarios, and the reactor design relies on multiple engineered safety systems, active cooling, and robust containment structures to manage accident conditions. Accident analyses for large LWRs frequently require a high level of analytical rigor, including the use of sophisticated probabilistic risk assessment methodologies and computational tools to characterize plant responses and overall risk profiles. While appropriate for complex, high-power facilities, this level of analysis is resource intensive and not well suited to the streamlined processes needed to support high-volume licensing. In contrast, many advanced non-LWR designs incorporate inherent safety features—such as low-pressure operation, high thermal capacities, and strong negative reactivity feedbacks—that reduce the likelihood and severity of accidents. Also, small LWRs, while similar in technology to large LWRs, generally benefit from reduced core power levels and power density, fission product inventories, and simpler system layouts, leading to more straightforward accident analyses. As such, these non-LWR and small LWR risk profiles can demonstrate the designs' low consequence without a very large site boundary and without extensive reliance on probabilistic risk assessment methods. These safety features and relatively small sizes and source terms as compared to large LWRs lend themselves to licensing and manufacturing standardization, which makes these types of reactors more conducive to efficient, high-volume licensing. To understand the various reactor technology safety profiles, the NRC reviewed several published scientific studies, NRC's preliminary safety evaluation reports, and environmental review documents. The review focused on identifying common design attributes among these reactors—such as strong negativity reactivity feedback, robust fuel forms, higher thermal margins, and passive heat removal—that inherently limit transient and accident progression. The NRC found non-LWR designs and microreactors are often designed with large thermal capacities that allow them to dissipate operational and decay heat passively for relatively long periods of time without the need for active systems or operator action. These designs also feature large shutdown reactivity margins and other intrinsic safety characteristics that provide strong inherent barriers to accident progression. As a result, their overall safety behavior can be well understood without relying on sophisticated probabilistic or risk assessment methodologies, since the fundamental design attributes themselves demonstrate a robust ability to prevent and mitigate accidents that previous large LWR designs have traditionally been designed to accommodate. Accordingly, these designs do not necessarily have the need for traditional containments as there is a reduced likelihood of events occurring requiring such mitigation features. Furthermore, these designs would not warrant precautionary protective measures to respond to emergencies. Instead, as a final layer of defense in depth, licensees could rely on a risk-informed approach to emergency planning. Based on its evaluation of SNM inventories and safety characteristics of non-LWRs, small LWRs, and representative large LWRs, the NRC concluded that the establishment of a defined SNM material limit would be technically justified as an entry criterion to proposed part 57. This material limit would be defined as a total inventory of thorium, uranium, and plutonium contained in the nuclear reactor not to exceed 10 metric tons. The evaluation showed that designs within the material limit would likely have inherent and passive safety features and exhibit favorable safety profiles despite variations in core design and thermal power levels. Together, these insights support the NRC's determination that a numerical material limit that is risk-informed due to inherent and passive design features could be part of an appropriate regulatory threshold to using a licensing approach to enable rapid and efficient licensing of microreactors and other reactor designs with comparable risk profiles. 3. Design Criteria Attributes The design criteria attributes in proposed § 57.30—reactivity control, heat removal, fission product retention, shielding, radioactive effluent control, and security by design—are rooted in the fundamental principles of nuclear safety and radiation protection. • Reactivity Control—The reactor would need to be able to safely control the power level in normal operation, shut down quickly if needed, and stay safely shut down. The reactor would be required to have a natural “braking” effect: when temperatures rise, the power level automatically falls (net negative reactivity feedback). Also, if the fuel would be loaded into the reactor at a manufacturing facility, then the reactor design would need to have built-in protections to prevent the reactor from unplanned criticality. • Heat Removal—Even after the reactor is shut down, heat keeps being produced. The design would be required to have highly reliable, passive systems to keep the reactor cool and within safe temperature limits, even if the main cooling system fails during events like power loss or earthquakes. • Fission Product Retention—Barriers like the fuel itself and the reactor vessel can retain radioactive materials during both normal operations and accident conditions. The design would need to keep temperatures and pressures well below the limits these barriers can handle. • Shielding—The reactor would need strong, durable shielding to protect workers and the public from radiation, including during transportation. The design also would have to account for heat that builds up in shielding and the removal of the heat if needed. • Radioactive Effluents Control—The reactor would be required to meet limits for any radioactive gases, liquids, or solid wastes it would release, and have monitoring and handling systems that protect people and the environment. • Security by Design—Where possible, the design itself should address security risks, using built-in engineering and physical protection features instead of relying only on procedural measures. D. Subpart C—Construction Permits and Operating Licenses Proposed subpart C would provide requirements related to applications for NRC licenses to construct and operate utilization facilities for commercial or industrial purposes under part 57. The AEA calls these licenses “construction permits” and “operating licenses,” and the NRC proposes to use that nomenclature in proposed part 57 as it has done in part 50. Proposed part 57 would include licensing options based on the CP and OL approaches in part 50, and proposed subpart C would contain several sections that would be similar to existing regulations in part 50. Proposed § 57.45, “License required; exceptions from licensing,” would address required licenses and identify certain exceptions from licensing. Proposed § 57.45(a) would describe activities requiring an NRC license and would be equivalent to § 50.10(b). Proposed § 57.45(b) would govern an exemption from the licensing requirements under proposed part 57. This proposed requirement would be equivalent to that in § 50.11(c). Proposed § 57.45(c) would require issuance of a construction permit, with the exception in proposed § 57.45(d), prior to starting construction of a utilization facility at a site and would be equivalent to § 50.10(c). Proposed § 57.45(d) would issue a general license for construction activities on a site that is specified in a joint application for a CP and associated OL(s) under proposed part 57 for a nuclear reactor or nuclear plant subject to certain conditions in proposed § 57.45(d)(1)-(7). The proposed general license would allow the general licensee to perform construction, as would be defined in proposed § 57.3, before NRC issuance of a construction permit for the nuclear reactor or nuclear plant. Proposed § 57.45(d)(1) would require that the general licensee has submitted, and the Commission docketed, a joint application for a CP and associated OL(s) under proposed part 57. This proposed requirement would include several additional conditions on the joint application. First, the joint application would be required to reference an ML issued by the Commission under 10 CFR chapter I. This condition would provide assurance that the general licensee would not complete construction of the nuclear reactor or nuclear plant before issuance of the CP because the manufactured reactor would be an essential part of the reactor or plant and proposed § 57.45(d)(5) would prohibit bringing it to the site under the general license. Second, the joint application would be required to reference a CP and OL issued pursuant to proposed part 57 that the Commission afforded generic finality under proposed § 57.142(e) and that referenced the same ML as the general licensee's joint application. This condition would ensure that the complete design had been reviewed and approved by the NRC and that a nuclear reactor or nuclear plant of the same design had been successfully constructed under NRC oversight and placed into operation. This would also ensure that the public had been afforded an opportunity for hearing on the design, including the postulated site parameters for the design, in accordance with §§ 57.142(e) and 57.60(c). Third, the joint application would be required to reference a design that met the criteria for a categorical exclusion under proposed subpart K of part 57. Taken together, the requirements proposed in § 57.45(d)(1)(i) and (ii) would provide assurance that the SSCs of the nuclear reactor or nuclear plant, which could be difficult to change after their construction, would not pose obstacles to eventual issuance of an OL under proposed part 57. Fourth, proposed § 57.45(d)(1)(iii) would require the joint application to include a plan for redress of any adverse environmental impact from conduct of activities under the general license should such redress be necessary. This proposed requirement would be similar to the requirements in § 50.10(d)(3)(iii), which requires a redress plan as part of an application for a limited work authorization, and § 50.11(b)(2), which requires the Commission to consider redress of adverse environmental impacts in determining whether to grant an exemption permitting the conduct of construction activities prior to the issuance of a construction permit. Proposed § 57.45(d)(2) would require that the general licensee has notified the NRC under proposed § 57.4 that all applicable permits, licenses, approvals, and other entitlements in connection with the proposed action that the general licensee was responsible for obtaining have been obtained. Proposed § 57.45(d)(3) would require that applicable Federal environmental consultations have been completed. This would ensure that construction activities would not begin unless the NRC has the information it would need to fulfill its obligations for environmental review under the AEA, NEPA, and other relevant laws. Proposed § 57.45(d)(4) would require that the general licensee not allow SNM or radioactive material that would be associated with the operation of the nuclear reactor or nuclear plant under an operating license issued pursuant to proposed part 57 to be brought to the site. This would ensure that activities under the general license would not create radiological hazards or irreversible radiological impacts at the site that would otherwise be controlled by a CP or OL under proposed part 57. This would also ensure that activities under the proposed general license would not involve radiological security concerns. In addition, proposed subpart P of part 26 would require implementation of an appropriate FFD program during construction. Proposed § 57.45(d)(6) would require that the general licensee allow for any NRC inspections that the Commission would deem necessary related to activities that would be performed under the general license. This would ensure that the NRC could apply experience gained from inspection of the construction of the same nuclear reactor or nuclear plant design if needed during construction activities that would be conducted under the proposed general license. Proposed § 57.45(d)(7) would clarify that any activities undertaken by the general licensee or on its behalf under the general license would be entirely at the risk of the general licensee and would have no bearing on the issuance of a construction permit under proposed part 57 with respect to the requirements of the AEA, and rules, regulations, or orders issued under the AEA. However, the general licensee would be able to mitigate this additional regulatory risk through careful site selection to ensure that site characteristics are within the bounds of the postulated site parameters and by performing construction activities following appropriate QA and FFD programs. Based on the proposed requirements in § 57.45(d)(1)-(7), the Commission has determined that such general licensing would be for only parts of utilization facilities, not constitute an unreasonable risk to the common defense and security, and, therefore, be consistent with the authority provided to the Commission by section 109(a) of the AEA. Proposed § 57.55, “Content of applications; general information,” would provide general information requirements for the content of joint applications under proposed part 57 and would be equivalent to § 50.33, “Content of applications; general information,” with the exception that no emergency planning zones would be defined for facilities licensed under proposed part 57. Proposed § 57.60, “Contents of applications; technical information,” would provide technical information for the content of joint applications and would be equivalent to § 50.34, “Contents of applications; technical information,” but would not include a preliminary safety analysis report. Proposed § 57.60(a) would provide the technical requirements for an FSAR submitted as part of a joint application under proposed part 57. Proposed § 57.60(a)(1)(i) would address the intended use of the reactor to include maximum power and inventory of radioactive material. Proposed § 57.60(a)(1)(ii) would provide requirements for an FSAR to describe and assess safety features and barriers designed into the facility to prevent or mitigate the consequences of an accident similar to § 50.34(a)(ii)(D) without the requirement to comply with part 100 or the radiation dose criterion for an individual in § 50.34(a)(1)(ii)(D). Proposed § 57.60(a)(1)(iii) would require the applicant to demonstrate, through an evaluation, that the dose- based entry criterion specified in proposed § 57.25(a) is satisfied. Proposed § 57.60(a)(1)(iv) through (vi) would require the applicant to describe the design features associated with any remote or autonomous operation or remote monitoring capabilities. Proposed § 57.60(a)(1)(vii) would require the applicant to provide the analysis, appropriate test programs, prototype testing, operating experience, or a combination thereof that would demonstrate that each of the design criteria attributes described by proposed § 57.30 would be met. Proposed § 57.60(a)(2) would require the applicant to include design basis and principal design criteria information in the application including the relation of the design bases to the design criteria, and the relation of the principal design criteria to the design criteria attributes described in proposed § 57.30. The principal design criteria establish the necessary design, fabrication, construction, testing, and performance requirements for safety-related SSCs that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public. The reference to principal design criteria in proposed § 57.60(a)(2) would not require the applicant to meet the General Design Criteria in appendix A of part 50. However, the General Design Criteria in appendix A could be generally applicable to other types of nuclear plants and used as guidance in establishing the principal design criteria for a facility using part 57. This proposed rule would not impose QA requirements under existing appendix B to part 50. Proposed § 57.60(a)(3) would require the applicant to describe its QA program to be applied to the design, fabrication, manufacturing, construction, and testing of safety-related SSCs and would be equivalent to § 50.34(a)(7). Qualified suppliers of nuclear-grade SSCs have decreased over the last several decades. This shrinking base of suppliers, increasing demand for advanced reactors, existing SSC upgrades and maintenance needs for the operating fleet, restart of shutdown plants, and policies to buy U.S. products, are creating a need for new suppliers to enter the market. At the same time, the evolution of quality system requirements has led to the development of several QA standards with shared elements. The NRC's proposal to enable applicants to select QA programs could broaden the supplier base and increase flexibility in procurement. This approach may encourage participation from qualified commercial suppliers, thereby expanding the pool of vendors available to support nuclear projects. This could mitigate risks of shortages, backlogs, and higher costs of deployment of microreactors and reactors with comparable risk profiles. Proposed § 57.60(a)(4) would specify requirements related to sites at which multiple nuclear reactors may be built or installed. Proposed § 57.60(a)(4)(i) and (ii) would require the applicant to analyze and specify limits on the number and configuration of reactors at the site and evaluate potential hazards to safety-related SSCs of any operating reactors that could arise from activities associated with construction, operation, and decommissioning of other reactors at the site. These requirements would be similar to existing requirements in § 50.34(a)(11). Proposed § 57.60(a)(4)(iii) would require the joint application to include a description of the portions of the nuclear plant that a nuclear reactor would share with one or more other reactors over the lifetime of the plant and to specify the functional requirements and measures to meet the requirements for any shared safety-related SSCs. Proposed § 57.60(a)(4)(iv) would require the joint application to include technical specifications, as appropriate, for shared portions of the nuclear plant. Proposed § 57.60(a)(5) would require the applicant to include current and projected population distributions and site evaluation factors for seismic, meteorological, hydrologic, and geologic characteristics with appropriate consideration of natural phenomena. The reason for establishing siting requirements would remain the same as it has been historically, which is to ensure that licensees and applicants assess what impact the site environs may have on a nuclear plant ( e.g., external hazards) and, conversely, what potential adverse health and safety impacts a nuclear plant may have on nearby populations in view of the site characteristics. Natural phenomena's and site characteristics' impacts are key inputs into the design of safety-related SSCs to ensure they can perform their intended safety functions. The information required by proposed § 57.60(a)(5) would inform site selection demonstrating that the site characteristics would be bounded by site parameters postulated for a given design. Proposed § 57.60(a)(6) would require the applicant to provide an analysis and evaluation of safety-related SSCs related to performance requirements and information that show that safety functions will be accomplished and would be equivalent to § 50.34(b)(2). Proposed § 57.60(a)(7) would require the applicant to provide information on the kinds and quantities of radioactive materials expected to be produced by operation and the means for controlling and limiting radioactive effluents and radiation exposures within the limits set forth in 10 CFR part 20 and would be equivalent to § 50.34(b)(3). The application would have to include an estimate of the quantity of each of the principal radionuclides expected to be released annually to unrestricted areas in liquid effluents produced during normal reactor operations, an estimate of the quantity of each of the principal radionuclides of the gases, halides, and particulates expected to be released annually to unrestricted areas in gaseous effluents produced during normal reactor operations, and a description of the equipment and procedures for the control of gaseous and liquid effluents and for the maintenance and use of equipment installed in radioactive waste systems. Proposed § 57.60(a)(8) would require the applicant to provide information related to operational programs concerning facility operations. These programs could be developed specifically for an individual reactor or generically for a particular design to be administered at a corporate or institutional level to support fleet operations. Proposed § 57.60(a)(8)(i)-(iii) would require the applicant to include information related to the organizational structure, training and qualification, conduct of operations, plans for preoperational testing and initial operations, and plans for normal operations, and would be equivalent to § 50.34(b)(6)(i)-(iv). Proposed § 57.60(a)(8)(iv) would require emergency plans for responding to an accidental release or loss of control of radioactive material. Proposed § 57.60(a)(8)(iv) would also require the applicant to coordinate response needs with local emergency planning and offsite response organizations. This proposed provision would ensure adequate communication, coordination, and cooperation among applicants, licensees, and offsite response organizations to establish agreements and arrangements for offsite support and to ensure protective measures can and will be taken as conditions warrant. An emergency planning zone (EPZ) would not be defined for facilities licensed under proposed part 57. An EPZ is most useful as a planning tool for implementing precautionary actions through predetermined, prompt protective measures to respond to events that involve a wide-scale area involving multiple jurisdictions and rapidly progressing incidents that could result in acute doses or early health effects. The characteristics of facilities that would be licensed under proposed part 57 provide assurance that planning for such precautionary actions is unnecessary. Consistent with other NRC-licensed facilities that do not have defined EPZs, the proposed rule would ensure that applicants and licensees develop and maintain capabilities to protect emergency workers and the public. Proposed § 57.60(a)(8)(v) would require the applicant to describe its physical security program, cybersecurity program, information security program, and access authorization program and is equivalent to § 50.34(c). The physical security program would need to meet the security requirements in part 70. For radiological sabotage, because these events could disrupt the performance of the design of reactors licensed under proposed part 57, the applicant would need to perform an assessment against the threat of radiological sabotage. The purpose of this assessment would be to evaluate the design against security events derived from the design basis threat (DBT) of radiological sabotage defined in § 73.1, “Purpose and scope,” to determine if an operational program for physical security is needed. The criterion for the assessment in proposed § 57.60(a)(8)(v)(A)( 3 ) would require an applicant to show that potential consequences resulting from an event initiated by the DBT would result in offsite doses below the values in § 50.34(a)(1)(ii)(D) even if mitigation and recovery actions, including any operator action, were unavailable or ineffective. For those proposed part 57 applicants not able to meet the criterion in proposed § 57.60(a)(8)(v)(A)( 3 ), proposed subpart J would provide performance-based requirements for licensees. Proposed § 57.60(a)(8)(v)(B) would require licensees to establish, implement, and maintain a cybersecurity program in accordance with either § 73.54, “Protection of digital computer and communication systems and networks,” or proposed § 73.110, “Cybersecurity program.” Proposed § 57.60(a)(8)(v)(C) would require licensees to establish, implement, and maintain an information protection system that complies with the requirements of §§ 73.21, “Protection of Safeguards Information: Performance requirements,” 73.22, “Protection of Safeguards Information: Specific requirements,” and 73.23, “Protection of Safeguards Information—Modified Handling: Specific requirements,” as applicable. Proposed 57.60(a)(8)(v)(D) would require licensees to establish, implement, and maintain an access authorization program in accordance with § 73.56, “Personnel access authorization requirements for nuclear power plants.” Proposed § 57.60(a)(8)(vi) would require the applicant to provide proposed technical specifications prepared in accordance with the requirements of § 50.36, “Technical specifications,” and would be equivalent to § 50.34(b)(6)(vi). Proposed § 57.60(a)(8)(vii) would require the applicant to submit procedures to be used to provide assurance that limiting conditions for any operating reactors will not be exceeded as a result of activities associated with the construction of any additional reactors at the same site and would be equivalent to § 50.34(b)(6)(vii). Proposed § 57.60(a)(8)(viii) would require the applicant to provide a radiation protection program as part of its application and would be similar to § 20.1101, “Radiation protection programs.” Proposed § 57.60(a)(8)(ix) would require the applicant to provide a fire protection program and would be similar to § 50.48(a). Proposed § 57.60(a)(8)(ix)(A)-(C) would require the applicant to describe the fire protection program for the facility, any specific features necessary to implement the program, and an analysis to demonstrate that a fire or explosion in any area of the plant would not prevent a safety-related SSC from performing its safety function. Proposed § 57.60(a)(8)(ix)(D)-(H) would establish specific requirements for the fire protection program. Proposed § 57.60(a)(8)(x) would require the applicant to describe how the human factors engineering requirements of proposed § 57.395 would be addressed. Proposed § 57.60(a)(8)(x) would also require the applicant to describe the training, examination, and proficiency programs necessary to meet the requirements of proposed subpart P. Proposed § 57.60(a)(8)(xi) would require the applicant to submit its description and plan for implementation of a remote operation or monitoring program, if applicable. Remote operation and remote monitoring are defined in proposed § 57.3 as control of the reactor and observation of plant data, respectively, from a location outside of the site boundary. Stakeholders have expressed interest in the incorporation of remote operation and monitoring into their plant designs. Proposed § 57.60(a)(8)(xii) would require the applicant to submit its program to ensure that systems and components meet the requirements in the codes and standards identified in the ap [Text truncated at 120,000 characters. The full text is on the page linked above.] --- Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A2026-08550. Public record. Not legal advice.