ML25051A092

Comment (084) from Douglas E. True on Behalf of the Nuclear Energy Institute on Part 53 Rulemaking - Risk-Informed Technology-Inclusive Regulatory Framework for Advanced Reactors

ML25051A092
Person / Time
Site:	Nuclear Energy Institute
Issue date:	02/14/2025
From:	True D Nuclear Energy Institute
To:	NRC/SECY, NRC/EDO
References
NRC-2019-0062, 89FR86918, 90FR92609
Download: ML25051A092 (1)

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Doug True Sr. Vice President & Chief Nuclear Officer Phone: 925.998.8810 Email: det@nei.org February 14, 2025 Dr. Mirela Gavrilas Executive Director of Operations U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Submitted via Regulations.gov

Subject:

NEI Paper on NRCs Rulemaking on the Post-SRM Part 53, Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors (RIN-3150-AK31; NRC-2019-0062)

Project Number: 689

Dear Ms. Gavrilas:

The Nuclear Energy Institute (NEI)1 and our members appreciate the Nuclear Regulatory Commissions (NRCs) work to develop a more effective and efficient framework for new reactors, and we are pleased to submit comments on the proposed Part 53 rule. We have been engaging with the NRC over the past several years in developing the Part 53 framework that Congress directed in the Nuclear Energy Innovation Modernization Act (NEIMA). Since that time, Congress further directed, in the ADVANCE Act, the NRC to make more substantive changes to the regulatory framework, and NEI submitted a proposed approach for Rapid High-Volume Deployments (RHDRA) to implement major portions of the ADVANCE Act for advanced reactors. More recently, the President has issued Executive Orders that articulate the need for more aggressive regulatory changes to unleash Americas energy, and other voices affirm that more aggressive changes are needed to achieve a more effective and efficient regulatory framework. As directed by Congress, the NRC also updated their mission statement to emphasize regulatory efficiency and enabling the deployment of new reactors. Unfortunately, the NRCs current efforts to modernize the regulatory framework, including the proposed Part 53 and other efforts, does not adequately address these relevant and major developments and fails to meet the moment for ensuring that the NRC is enabling the safety and secure use of nuclear energy technologies through an efficient and effective regulatory framework.

The NRC would need to pursue significantly more aggressive regulatory changes to meet the direction from Congress and the President to achieve the most effective and efficient regulatory framework 1 The Nuclear Energy Institute (NEI) is responsible for establishing unified policy on behalf of its members relating to matters affecting the nuclear energy industry, including the regulatory aspects of generic operational and technical issues. NEIs members include entities licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, major architect and engineering firms, fuel cycle facilities, nuclear materials licensees, and other organizations involved in the nuclear energy industry.

Dr. Mirela Gavrilas February 14, 2025 Page 2 Nuclear Energy Institute possible, and to establish a foundation for future enhancements as they are identified. Fortunately, the Part 53 rule could be adapted and expanded to accomplish these goals through the following significant actions:

1. Eliminate undue regulatory burden by eliminating unnecessary and duplicative requirements (e.g.,

many requirements after 53.430, unless specifically needed to meet the Atomic Energy Act, such as Generally Licensed Operators), and moving details out of the rule language and into guidance.

A high-level rule is consistent with the Vote by Chair Wright, and the input from NEI and others at the beginning of the rulemaking process

2. Formulate the requirements consistent with the NEI RHDRA proposal that achieves a more effective and efficient framework for all advanced reactor technologies (including but not exclusive to micro-reactors) that present a very low level of hazard and potential impact to the public health and safety. This approach should consider establishing performance based requirements that are capable of being graded appropriately to reactors that achieve site boundary emergency planning zones. See Enclosure 3, Topic 16 for changes that can address the ADVANCE Act without delaying the Rulemaking schedule.

3. Eliminate consideration of a Part 56 (formerly Framework B of Part 53), which is diverting resources from pursuing more significant improvements in Part 53.

4. Perform a systematic and aggressive search for potential changes, in requirements, policy and guidance, to reduce unnecessary regulatory burden that have not yet been fully considered, such as Population Siting and ALARA, and requirements that have not been updated for decades, such as Part 100 and Part 50 Appendix B. If necessary, NRC should identify changes to the Atomic Energy Act that would enable a more efficient approach to licensing and regulating new reactors while protecting the public health and safety.

5. Perform more accurate cost/benefit analyses by assessing each change (as compared to the current Parts 50 and 52) individually to avoid inadvertently including burdensome requirements through cost/benefit analyses of the aggregated changes.

The NRC can achieve significantly more aggressive changes through expansion of Part 53 to meet the moment and achieve the most effective and efficient regulatory framework for advanced reactors that is being directed by Congress and the President. Protecting the public health and safety without imposing undue burden is important for the entire nation and is consistent with the recently updated NRC mission.

All of these can be implemented on the current NRC schedule for completing the Final Part 53 Rule by 2027, since 1) a substantial amount of work has been completed and the comments in this letter provide specific changes the NRC can make to significantly improve the existing rule language, 2) a majority of changes to the Part 53 rule is to move it from rule language to guidance and eliminating pursuit of Part 56 will make available more resources for Part 53, 3) NEI and other stakeholders have already provided the majority of the input needed to incorporate more aggressive approaches, 4) the NRC can use public

Dr. Mirela Gavrilas February 14, 2025 Page 3 Nuclear Energy Institute workshops to more quickly perform a systematic search for more aggressive changes, and 5) higher level requirements with details in guidance require fewer resources to establish and maintain, and enable greater flexibility for novel approaches in the future.

Over the course of the Part 53 rulemaking, we have promptly identified our concerns to the staff and have appreciated constructive dialog and evolution of Part 53 toward the framework that is needed to enable the timely, efficient, and cost-effective deployment of the next generation of reactors to meet our nations growing demand for the clean. The Rulemaking took a significant step forward with the Commission directive in SRM-SECY-23-0021 with many of the industrys largest concerns addressed. Notwithstanding the need to implement more significant changes in Part 53, unresolved issues remain that, if unaddressed, would prevent the widespread adoption of Part 53. This is a significant issue, since the purpose of developing Part 53 is to provide a more efficient licensing framework for advanced reactors. At this time, very few of our members are interested in using Part 53, in large part due to the regulatory uncertainty of the new Rule. Our hope is that Part 53, once finalized provides significant regulatory efficiency benefits for all reactors (Large LWRs, LWR SMRs, non-LWRs and microreactors). An efficient, technology-inclusive Part 53 should become the preferred licensing pathway for all new reactors and meet the goals of NEIMA, the ADVANCE Act and the Executive Orders of the current administration.

The areas of greatest concern include: the requirement for an all-hazards probabilistic risk assessment (PRA), the inability to transfer safety evaluations from Parts 50 or 52 to Part 53, increased requirements for Non-Safety Related, Safety-Significant (NSRSS) Structures Systems and Components (SSCs),

including in the area of Codes and Standards (C&S), and the lack of supporting regulatory guidance. The enclosures with this letter contain the comprehensive set of feedback on our major concerns and suggested changes to address them, as well as other comments intended to help improve the rule.

Requirement for an All-Hazards PRA Despite direction from the commission that the staff should allow PRA acceptability determinations for Part 53 applications to be appropriately flexible, considering how PRA insights are relied upon to support the licensing application, together with factors such as safety margin, simplicity of design, and treatment of uncertainties, the requirement in 53.450(a) for a PRA remains unchanged:

§ 53.450 Analysis requirements.

(a) Requirement to have a probabilistic risk assessment (PRA). A PRA of each commercial nuclear plant must be performed to identify potential failures, susceptibility to internal and external hazards, and other contributing factors to event sequences that might challenge the safety functions identified in § 53.230 and to support demonstrating that each commercial nuclear plant meets the safety criteria of § 53.220, or more restrictive alternative criteria adopted under § 53.470 While language in both the preamble and 53.450(e) suggest more flexibility should be provided, the 53.450(a) language clearly requires an all-hazard PRA. In order to enable a more performance-based and

Dr. Mirela Gavrilas February 14, 2025 Page 4 Nuclear Energy Institute risk-informed rule, the language should change to require a risk evaluation consistent with Commissioner Caputos markup of Part 53, which could be fulfilled by a PRA or qualitative risk assessment. This would be in line with the ADVANCE Act language that microreactors should develop strategies and guidance for risk analysis methods, including alternatives to probabilistic risk assessments. At a minimum, though it would not enable broad usability, the language needs to change to PRA in combination with other generally accepted approaches for systematically evaluating engineered systems so that 53.450(a) is consistent with 53.450(e), the preamble discussion and the Commission directive in the SRM. To help understand how alternatives to a PRA-led methodology can work with Part 53, NEI intends to attach providing a mapping for how existing or draft regulatory guidance for 4 methodologies can work with Part 53, including minor proposed changes to the Part 53 rule language to facilitate that methodological flexibility. NEIs detailed comments are in Enclosure 3 in the response to the NRC Request for public comment on PRA acceptability and Enclosure 2 under Subpart C.

Inability to Transition from Parts 50/52 to Part 53 / Regulatory Certainty The urgent need for advanced reactor deployments are leading early movers to pursue Part 50 and 52 applications in the near term, since the timeline for Part 53 availability does not support business plans with online operations until sometime in the early to mid-2030s. While part 53 does provide benefit in reduced exemptions, much of the benefit is in reduced programmatic burden at the operating stage. For this reason, it is essential that early stage approvals (e.g., early site permit, construction permit, standardized design approval and design certification) received under Parts 50 or 52 should be allowed by applicants to seek later stage approvals (e.g., operating license and combined licenses) under Part 53 while maintaining any finality approved under Parts 50 and 52. A number of early site permits (ESPs) already exist or are being pursued by utilities. ESPs would rely on the same technical and licensing bases whether under Part 52 or 53 and therefore should be transferable without issue. The Advanced Reactor Demonstration Project (ARDP) winners are both pursuing Licensing Modernization Project (LMP)

Construction Permits (CP) under Part 50. If Part 53 is available at the time that the ARDP winners pursue an Operating License (OL), they should have the option of pursuing the OL under Part 53 without reopening any of the safety issues resolved as part of the CP review or other engagement. Similarly, several microreactor vendors supported by the Department of Energy (DOE) are pursuing design certifications under Part 52. These should have the option to support Part 53 COLs once Part 53 is available. Finally, many Topical Reports, while written for Part 50 or 52 requirements remain equally valid for meeting Part 53 requirements and there should be an efficient pathway for transferring those safety evaluations. Currently this is precluded under Part 53 due to language in the proposed 10 CFR 53.1124, Relationship between sections.

(g) Operating license. (1) An application for an OL under this part may, but need not, reference an early site permit, standard design certification, or standard design approval issued under this part.

Dr. Mirela Gavrilas February 14, 2025 Page 5 Nuclear Energy Institute (h) Combined licenses. An application for a COL under this part may, but need not, reference an early site permit, standard design certification, standard design approval, or ML issued under this part.

NEI suggests removing issued under this part from this section and other sections that could preclude a transition from Parts 50 or 52 to Part 53. NEI also suggests a discussion in the Rule Package that such a transition should not reopen settled safety issues. NEI acknowledges that Part 53 does have slightly different language than Parts 50 and 52, but the technical and licensing basis for many of these requirements remain the same, particularly for Part 50 and 52 licensees following LMP. To help facilitate this understanding, NEI has provided, in Enclosure 2, a list or priority Regulatory Guidance documents that must be updated as a high priority effort to support implementation of Part 53 to provide the industry confidence that NRC-endorsed methods will remain valid for Part 53 licensees. Regulatory uncertainty was a hindrance for the first use of Part 52, and a robust Rule Package can reduce the regulatory uncertainty to encourage the use of Part 53. NEIs detailed comments on proposed language changes to facilitate transferability can be found in the detailed comments in Enclosure 2 Subparts E and H.

Without this change there needs to be some pathway for Part 50 and 52 applicants and licensees to gain access to the flexibility allowed under Part 53 for Generally Licensed Reactor Operators (GLROs), security requirements, Fitness for Duty, Access Authorization and more. In the long term, flexibility in Parts 50 and 52 could reduce the need for exemptions that would be unnecessary under Part 53.

Requirements for NSRSS SSCs, including C&S NEI is concerned that language in Part 53 suggests requirements traditionally limited to SR SSCs or a subset of important to safety SSCs under Parts 50 and 52 are being placed on all NSRSS SSCs under Part 53. 53.440(e) and 53.875(b) expands fire protection requirements to SR and NSRSS versus the traditional safe shutdown guidance endorsed in RG 1.189. The requirements in 53.440(c) and (d) expand environmental qualification requirements to all NSRSS SSCs where the current fleet is limited to SR and the scope of 10 CFR 50.49. Most problematic, 53.440(b) requires NRC-endorsed Codes and Standards (C&S) for all SR and NSRSS SSCs. Other examples are provided in the detailed comments of Enclosure

2. In each case the requirement is appropriate for SR SSCs and guidance can provide the scope of NSRSS SSCs for which there should be applicability to meet the design criteria.

NEI appreciates the staff not codifying Codes and Standards into Part 53 like 10 CFR 50.55a. Because of the breadth of designs under consideration, the traditional approach would have been problematic and the Rulemaking process too slow for designs built upon C&S that are being updated rapidly as C&S catch up with innovative technologies. NEI appreciates the coordination between the DOE and NRC on AR C&S development and applaud former Chairman Hanson and the staff for making AR C&S a priority. However, the language in 53.440(b) errs by expanding the scope of C&S requirements to NSRSS SSCs and implying the C&S itself must be endorsed by the NRC prior to use in an application. This would be an increase in regulation for non-safety related SSCs resulting in higher regulatory burden without an increase in safety.

Dr. Mirela Gavrilas February 14, 2025 Page 6 Nuclear Energy Institute NEI appreciated comments from the staff during the November 2024 public meeting on Part 53 that the intent was not to ratchet the requirements and that existing C&S endorsements were sufficient to meet the requirements of 53.440(b). The language wherever applicable might provide some flexibility, but the reference to 53.400 defines applicability to SR and NSRSS SSCs. Since 10 CFR 50.55a and the existing Regulatory Guidance focuses on C&S for SR SSCs, there is no economical pathway for meeting 53.440(b) for NSRSS SSCs. Many applicants would feel forced into applying SR C&S to NSRSS which goes against the spirit of LMP and Part 53. It also is contradictory with other parts of Part 53, such as 53.460(b). 53.460(b) applies Appendix B only to SR SSCs which is met in the current fleet by applying ASME NQA-1. Appendix B is elective for NSRSS SSCs under 53.460(b). Since NRC has endorsed no QA standard for NSRSS SSCs, to meet 53.440(b) applicants would be forced to apply NQA-1 to both SR and NSRSS SSCs.

NEI suggests limiting the scope of 53.440(b) to SR SSCs. NRC is welcome to question whether the treatment of NSRSS SSCs is sufficient to achieve their design criteria and other regulatory requirements on an application specific basis. In most cases this will include the use of industry consensus Codes and Standards. However, many applicants anticipate using commercial C&S for NSRSS SSCs that are not yet endorsed by the NRC. This application is in line with NEI 18-04 as endorsed by RG 1.233. Without a significant effort by NRC to endorses potentially dozens of C&S for NSRSS SSCs prior to publishing Part 53, the Rule, as written, becomes unworkable.

If applicants see Part 53 as requiring SR requirements for NSRSS SSCs there is little reason for choosing Part 53 over the important to safety requirements in Parts 50 and 52. NSRSS requirements should be what is necessary and sufficient to meet the design criteria, reliability and capability requirements derived from Subparts B & C. This should be performance-based and not prescribed by regulation. If commercial C&S with monitoring in line with 53.715 demonstrate adequate performance, none of the other requirements currently prescribed in Part 53 should be necessary for NSRSS SSCs. NEIs detailed comments are in Enclosure 2, primarily comments on Subpart C.

Other There are other requirements described in detail in the Enclosures. Among these are the following that, while not fatal flaws along the lines of the issued previously discussed, do impose significant unnecessary regulatory burden. The Integrity Assessment Program (IAP) requirement, 53.870, is duplicative with existing Part 53 requirements and should be removed. Comprehensive Risk Metrics (CRM) lack guidance, open litigation risks and are unnecessary as a stadnalone requirement given the current Commission Policy. ALARA requirements, while improved from the previous Rule text, could be read as increasing regulatory requirements beyond the existing, adequate Part 20 requirements and guidance. Siting requirements are carried over from Part 100 without being modified to be performance-based in line with NEIMA. NRC should work with stakeholders to establish acceptable means to comply with 53.250 defense in depth (DID) requirements for non-LMP applicants. All of these requirements should be addressed in a way that removes duplicative or otherwise unneeded requirements and provides clarity in regulatory guidance.

Dr. Mirela Gavrilas February 14, 2025 Page 7 Nuclear Energy Institute Conclusion We are vested in the NRC developing a successful Part 53 that will be used and useful. The industry has invested significant resources in participating in the public meetings the NRC has held, and in reviewing and commenting on the draft rule language, in the hope of having an inclusive and efficient Part 53. We believe the ability of the nation to meet the nations growing clean energy demands depends on a successful Part 53 rulemaking. We also believe it is clear that, with relatively straight forward changes to the NRC staffs Part 53 proposed rule language, the NRC can establish a Part 53 rule that allows the variety of risk-informed licensing approaches that industry plans to use for advanced reactors and this can be accomplished on the Commission directed schedule.

Our hope is that the comprehensive comments in the attached enclosures will enable the staff to pursue an inclusive and efficient rule. Note that Enclosure 1 remains under review and will be submitted separately. The industry stands ready to assist in the adoption of these alternatives in the rule, including the development of necessary guidance. We appreciate the NRCs consideration of these licensing approaches in the formation of Part 53. If you have questions concerning our input, please contact me, or Jon Facemire at 202-256-0190 jwf@nei.org.

Sincerely, Douglas E. True Sr. Vice President & Chief Nuclear Officer : Technology-inclusive, Risk-informed, Performance-based Approaches for Development of Licensing Bases Under Part 53 - forthcoming : Detailed Part 53 Comments : Part 53 NRC Request for Public Comment : Part 26 Updates for Part 53 NRC Request for Public Comment : Part 73 Updates for Part 53 C:

Mr. John Lubinski, NMSS, NRC Mr. John Tappert, NMSS, NRC Ms. Laura Dudes, NRR, NRC Mr. Raymund Furstenau, RES, NRC Mr. Robert H. Beall, NMSS/REFS/RRPB, NRC Mr. Greg Bowman, NRR, NRC Mr. Jeremy Bowen, NRR/DANU, NRC Mr. Anders Gilbertson, NRR/DANU/UARP, NRC Mr. Michael Wentzel, NRR/DANU/UARP, NRC

Comments on NRCs Proposed Part 53 Page 1 of 123 General Comments Regulatory Guidance Updates Much of Part 53 relies on regulatory guidance already endorsed for Parts 50 and 52. To increase regulatory certainty, it is important that existing Regulatory Guides, Standard Review Plans, and other NRC guidance documents get updated to reflect the methodology is acceptable under Part 53. In theory, all current RGs should be acceptable for use under Part 53, at least for certain reactor types (e.g., LWRs).

It may be more efficient in the near term for the NRC to identify RGs that would NOT be acceptable under Part 53 or which would only be acceptable with limitations. We are providing a list of the regulatory guides we consider most essential for update to support Part 53 applications. The lower priority guidance documents are discussed in specific comments in the comprehensive comment tables below.

Highest Priority Functional Design Criteria are core to a Part 53 safety case and we expect no applicant would be willing to pursue Part 53 without either RG 1.233 or RG 1.232 updated. Similarly, content of application guidance is essential for regulatory efficiency, so RG 1.253 and associated guidance should be updated for both Licensing Modernization Project (LMP) and non-LMP users applying under Part 53.

RG 1.233, with changes in scope to address LWRs. If needed, NEI is happy to work with NRC to develop an addendum to NEI 18-04 to address LWRs. We were concerned with the approach NRC implied during the November public meetings that a separate RG would be required for LWRs. NEI believes that the same or very similar guidance should be captured in one Reg Guide to facilitate regulatory clarity. NEI identified that NEI 18-04 refers repeatedly to 50.34 dose requirements. If necessary for an update to RG 1.233, NEI 18-04 can be updated to reference both 50.34 and 53.210 dose limits, though we believe this is unnecessary since the requirements are equivalent.

RG 1.253, with changes in scope to address LWRs. If needed, NEI is happy to work with NRC to address an addendum to NEI 21-07 to address LWRs. RG 1.253 only allows flexibility for Construction Permit (CP) applicants following LMP and presumes an all-hazards, all-modes, all-sources Probabilistic Risk Assessment (PRA) for Operating License and Combined License applicants. While it is understood that RG 1.253 was published prior to the commission directive, the update to the preamble in Part 53 is insufficient to meet the commission directive for flexibility for applicants without a revision to RG 1.253. The language The level of detail in a CP PRA should be established using the process provided in Section 3 of ASME/ANS RA-S-1.4-2021, Risk Assessment Application Process. from RG 1.253 should be extended to all LMP applicants.

RG 1.232, with changes in scope to address LWRs. Arguably the Advanced Reactor Design Criteria (ARDC) work for LWRs given the increased flexibility in design criteria introduced by Part

53. It is a subset of the ARDC that would meet the Function Design Criteria (FDC) requirements of Part 53. The X-Energy Principal Design Criteria Topical Report (ML24190A060) describes which ARDC are FDC and which are more appropriately handled as special treatments.

Comments on NRCs Proposed Part 53 Page 2 of 123 High Priority While some applicants may be willing to pursue Part 53 without these documents, most would be unwilling to take the regulatory risk. These are essential for broader adoption of Part 53.

Given the reliance on PRA in Part 53, it is very important to have RGs providing guidance on the acceptability of PRA for Part 53 RG 1.247 - This update is on schedule but would benefit from lessons learned from peer review experience.

RG 1.200 - Technical work remains here, but as part of the comment package NEI has provided a means to combine Full Power Internal Events (FPIE) with bounding and screening approaches to provide confidence that the Quantitative Health Objectives (QHOs) would be achieved.

RG 1.174 - NEI intends to build off EPRIs work on a technology-inclusive advanced reactor risk metric and propose a technology-inclusive risk metric, along with performance objectives that provide confidence that the QHOs will be met. This report should be provided to the NRC for consideration in 2025 and endorsement is important for the ability to meet 10 CFR 53.220 requirements. There is also an opportunity to endorse Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) as a means of meeting the Part 53 requirement for a comprehensive risk metric.

While not directly required for a PRA, RG 1.203 is an essential input to the codes that the PRA will rely upon. It is also referenced in NEI 18-04 as the means of carrying out design basis accident analysis and therefore should be updated as acceptable for meeting the DBA requirements of Part 53.

o Similarly RG 1.183 and other guidance for source term calculations should be updated as potential means of meeting the Part 53 DBA requirements.

NEI appreciates the NRC effort to move Codes and Standards (C&S) from Code of Federal Regulation (CFR) 50.55a to Regulatory Guidance with the inclusion of 53.440(b) for C&S. This change, while appropriate, requires the C&S to be endorsed for Part 53 in Regulatory Guidance. Many of these changes will require little technical work as the technical basis should remain unchanged.

RG 1.26, RG 1.28 and RG 1.33 endorsing NQA-1 RG 1.84 endorsing ASME Section III Code Cases RG 1.87 endorsing ASME Section 3 Division 5 RG 1.97 endorsing IEEE 497 - The 2025 edition is proposed for endorsement as it includes flexible language aligned with LMP.

RG 1.192 endorsing the ASME OM Code, or preferably a future update that endorses OM-2.

RG 1.208 or RG 1.251 endorsing ASCE 43 and ASCE 4. DG-1410 should be finalized to support Part 53.

RG 1.246 endorsing ASME Section XI Division 2 - The applicability should be extended to LWRs.

Digital I&C RGs, including:

o RG 1.152 endorsing IEEE 7-4.3.2 for digital system requirements.

o RG 1.168 - RG 1.173 endorsing standards for software lifecycle. NOTE - NEI provided ADVANCE Act recommendations to evaluate modern system engineering approaches in lieu of isolated processes.

Comments on NRCs Proposed Part 53 Page 3 of 123 o RG 1.250 endorsing NEI 17-06 for use of IEC 61508 Safety Integrity Level (SIL) certified equipment.

NEI is concerned about the scope of 53.440 as discussed in the detailed comments below. We propose limiting the scope to Safety Related (SR) Structures, Systems, or Components (SSCs), but if the scope continues to include Non-Safety Related, Safety Significant (NSRSS) SSCs then NRC needs to endorse the following Codes and Standards as appropriate for NSRSS SSCs:

ISO-9001 - Note that 53.460 appropriately limits the requirement of Appendix B to SR SSCs with the option to apply Appendix B to NSRSS SSCs. The 53.460 language is appropriate and in conflict with the 53.440 requirement for using generally accepted consensus codes and standards that have been endorsed or otherwise found acceptable by the U.S. Nuclear Regulatory Commission. NQA-1 is the only C&S for Quality Assessment endorsed by NRC and was developed to meet the requirements of Appendix B. 53.440 effectively expands Appendix B requirements beyond SR SSCs to NSRSS SSCs.

o NEI 22-04 provides a pathway for ISO-9001 suppliers to meet Appendix B suppliers and should be considered for endorsement as a means of meeting 53.440(b) requirements.

ASME Section VIII ASME B31.1 and B31.3 ASCE 7 AISC 360 ACI 318 Any additional unendorsed C&S that may been used or proposed for use for U.S. reactors not mentioned above.

Note this list is incomplete and effort will be needed to identify a comprehensive set of C&S that applicants intend to use for NSRST SSCs under LMP. To be clear, NEI does not believe that NRC endorsement of C&S for NSRSS SSCs is appropriate and that the appropriate resolution is to limit the scope of 53.440 (b) to SR SSCs. If the requirement remains as is, then a significant amount of work will need to be completed by NRC to endorse commercial standards without providing significant benefit in terms of safety.

53.440(c) covers equipment qualification which has been traditionally completed in line with the following regulatory guidance documents: RG 1.40, RG 1.73, RG 1.87, RG 1.89, RG 1.97, RG 1.100, RG 1.142, RG 1.152, RG 1.153, RG 1.156, RG 1.158, RG 1.180, RG 1.209, RG 1.210, RG 1.211, RG 1.213 and RG 1.243. All of these should be updated as a means of meeting 53.440(c). While some of these documents will require minimal technical update, the Part 53 safety case changes the traditional scope of SR and therefore technical updates may be required to align these RGs with the more flexible requirements of Part 53.

Part 53 retains a requirement for aircraft impact assessments in 53.440(j), therefore RG 1.217 should be updated to indicate NEI 07-13 is acceptable guidance for meeting 53.440(j) and other associated Part 53 requirements. NEI appreciates that the Part 53 language is more technology inclusive, and NEI is working on an addendum to NEI 07-13 which intends to develop additional acceptance criteria that would meet the Part 53 requirements.

Comments on NRCs Proposed Part 53 Page 4 of 123 The following are essential to meeting the requirements of Subpart D. Per the discussion of Subpart D in the FRN, Existing approaches could be used to demonstrate compliance with this requirement. NEI expects these will require minimal technical work to update.

Regulatory Guide 1.23. Meteorological Monitoring Programs for Nuclear Power Plants o NEI is working on an alternative methodology for meteorological data collection to meet the more flexible EP requirements in 10 CFR 50.160. It may be worth delaying an update of RG 1.23 to endorse this alternative methodology expected to be delivered to the NRC in early 2025.

Regulatory Guide 1.27, Revision 3. Ultimate Heat Sink for Nuclear Power Plants Regulatory Guide 1.29, Revision 6. Seismic Design Classification o Some technical modification is required to allow flexibility in selecting SR SSCs.

Regulatory Guide 1.59, Revision 2. Design Basis Floods for Nuclear Power Plants o Appendix K in DG-1290 is valuable and should be incorporated into the revision of RG 1.59 that supports Part 53.

Regulatory Guide 1.76, Revision 1. Design-Basis Tornado and Tornado Missiles for Nuclear Power Plants Regulatory Guide 1.91, Revision 3. Evaluations of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plants Regulatory Guide 1.102, Revision 1. Flood Protection for Nuclear Power Plants o Some technical modification is required to allow flexibility in selecting SR SSCs.

Regulatory Guide 1.132, Revision 3. Site Investigations for Foundations of Nuclear Power Plants Regulatory Guide 1.198, Revision 0. Procedures and Criteria for Assessing Seismic Soil Liquefaction at Nuclear Power Plant Sites Regulatory Guide 1.208, Revision 0. A Performance-Based Approach to Define the Site-Specific Earthquake Ground Motion Regulatory Guide 1.221, Revision 0. Design-Basis Hurricane and Hurricane Missiles for Nuclear Power Plant Regulatory Guide 4.7, Revision 4. General Site Suitability Criteria for Nuclear Power Stations Regulatory Guide 4.26, Revision 1. Volcanic Hazards Assessment for Proposed Nuclear Power Reactor Sites To support 53.620, Regulatory Guides for Fuels and Materials Facilities (Division 3) should be updated to provide guidance on criticality prevention for fueled microreactors at a manufacturing facility.

Change Criteria in 53.1550 deviate from the past guidance for change evaluation and there is a pre-decisional draft guide endorsing NEI 22-05 for change evaluation under LMP for Part 50 / 52. This regulatory guide, once published, should be updated as an acceptable means of meeting 53.1550.

Several programs in Subpart F require updates to existing NRC guidance (RGs and/or NUREGs) and potentially new guidance for Part 53:

The fire protection requirements in 53.440(e), 53.450(g)(1) and 53.875 should be met with an update to RG 1.189 with minimal technical update.

Comments on NRCs Proposed Part 53 Page 5 of 123 The radiation protection requirements in 53.850 should be met with updates to note the applicability of appropriate Division 8 RGs to Part 53 licensees.

The Offsite Dose Calculation Manual requirements in 53.850 may warrant new guidance or updates to existing guidance since the existing Offsite Dose Calculation Manual (ODCM) guidance (e.g., NUREGs 1301/1302) focuses on PWR and BWR light water reactors licensed under Parts 50/52.

RG 1.242 should be updated for emergency planning. An understanding of the proposed guidance in this revision is necessary to develop a fully informed response to the NRC request for public comment. For this reason, NEI asks that this revision be made available for public comment prior to the Part 53 rule being finalized.

RG 1.160 would support the maintenance program under Part 53. This requires significant technical update and NEI proposes to submit a methodology for review to the NRC sometime in 2026.

RG 1.143, Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants, - needs revision to address Parts 52 and 53. Specifically, it will need to address other waste streams not typically generated at an LWR.

Safety Evaluations (SEs) for NEI 08-08 Contamination Control, NEI 07-08A ALARA, NEI 07-03A Radiation Protection Program, & NEI 07-10A Process Control Program - Option 1: NEI will need to revise these to include Part 53 and submit them to NRC for review. None of their SEs mention Part 53. NRC could approve the revised templates via an SE or Option 2: Given these SEs describe the full spectrum of traditional RP programs, a license applicant could choose to implement these industry guidance documents noting any deviations applicable to their Part 53 reactor SSCs.

Regulatory Guide 8.19, Rev. 3 (June 1979), Occupational Radiation Dose Assessment in Light-Water Reactor Power Plants -- Design Stage Man-Rem Estimates. This RG lists 10 CFR Part 50 and should be revised to list Parts 52 and 53.

Regulatory Guide 8.25, Rev.1 (June 1992), Air Sampling in the Workplace. This RG lists 10 CFR Part 50 and should be revised to list Parts 52 and 53.

Regulatory Guide 8.27, Rev. 0 (March 1981), Radiation Protection Training for Personnel at Light-Water-Cooled Nuclear Power Plants. This RG lists 10 CFR Part 50 and should be revised to list Parts 52 and 53 and include non-light water reactors.

The Advanced Reactor Content of Application Project (ARCAP) series of guidance (DANU-ISG-2022-01 through DANU-ISG-2022-09) should also be updated to provide clarity that the guidance is equally applicable to Parts 50, 52 and 53. More detail is provided in Section 3 on updates to align the ARCAP guidance for use under Part 53.

Licensing Basis Event (LBE) Analysis Requirements The regulation uses the term LBEs other than DBAs. While industry understands the intent of this phrase, it could add confusion to the public and LBEs is sufficient with clarification in supporting guidance. The phrase LBEs other than DBAs is used in the context of safety criteria, defense in depth, uncertainties and analysis requirements. In all cases, replacing LBEs other than DBAs with LBEs does

Comments on NRCs Proposed Part 53 Page 6 of 123 not change the intent of the regulation and supporting guidance can provide clarity in the LBEs requiring assessment.

53.220 - RG 1.233 discusses Defense in Depth (DID) holistically accounting for both DBAs and non-DBA LBEs.

53.220 (b) - While DBAs do not have a frequency component, the Design Basis External Hazard Level (DBEHL) concept ensures that risk of a set consequence remains below a certain frequency. As an example, RG 1.76 provides 1E-7 exceedance frequency for high winds, so deterministic analysis of the DBEHL High Winds event ensures that consequences greater than a bounding DBA would occur at <1E-7 frequency.

o Note that NEI 21-07 uses the term Design Basis Hazard Levels (DBHLs) instead of DBEHLs to account for internal fires, floods and other hazards. Part 53 may consider updating language to address DBHLs or be clear when requirements apply to DBHLs vs DBEHLs.

53.240 - Language here is already appropriately flexible but requires a modifier to add clarity that only appropriate failures are considered.

53.250 - RG 1.233 discussion of DID holistically accounting for both DBAs and non-DBA LBEs.

53.420 - For design criteria, the General Design Criteria (GDC) provide requirements for capability and reliability. Informed by the GDC, DBAs and other LBEs can establish the safety criteria required under 53.420.

53.450 (e) - (1) other generally accepted approaches for systematically evaluating engineered systems should allow for traditional DBA approaches to address (2) for internal and external hazards and (3) for establishing functional design criteria. (4) is arguably the most difficult criteria to meet, but SSCs credited in DBAs can be deterministically considered significant and relatively straightforward criteria could be set for systems credited for DID as described above for 53.250.

To summarize the above, the directive in NEIMA was a technology-inclusive framework and language in Part 53 should be flexible enough to be inclusive of the existing fleet of reactors licensed with a traditional licensing framework. The above changes in regulatory language allow that flexibility. Such flexibility would lean into decades of licensing experience without requiring exemptions for small LWRs which can readily meet the criteria for exemption to GDC 17 as an example.

The above approach would also provide more flexibility in licensing approaches to licensees as the industry transitions from Parts 50 and 52 to 53. For example, a licensee with a construction permit under Part 50 should be able to transition to an Operating license or Design Certification under Part 53.

Industry does recognize that requirements are different and that additional information would need to be provided by the applicant for such a transition, but such an approach would provide public confidence in the equivalence of all parts in addressing public health and safety.

Comments on NRCs Proposed Part 53 Page 7 of 123 Detailed Comments on IV. Part 53 Framework Section Section Desc.

Comment Suggested Resolution Subpart A General Provisions-Definitions (pg 23)

The discussion of Functional Design Criteria (FDC) states These are new terms that have not previously been defined or used in NRC regulation. While this is true, the concept is closely related to Principal Design Criteria (PDC) required by Parts 50 and 52 and the General Design Criteria.

Suggest a discussion of FDC in the context of PDC. Guidance should be updated to allow use of a subset of the GDC for LWRs and the ARDC in RG 1.232 to be used for FDC under Part 53.

Subpart A General Provisions-Definitions (pg 24)

The discussion of DBAs is too restrictive; a PRA should not be required to identify DBAs, particularly for LWRs with decades of regulatory precedent for DBA definition.

Discussion of DBAs should allow for definition based on past precedent for LWRs where sufficient operating experience exists. DBA definition should also be allowed based on bounding or conservative assessments with accepted methods endorsed in guidance.

Subpart A General Provisions (pg 26) 53.110 copying 50.13 should obviate the need for Aircraft Impact Assessment requirements.

50.150 and 53.440(j) should be removed as unnecessary given 50.13 and 53.110.

Subpart B Comprehensive Risk Metric (pg 29)

The language to approximate the total risk is somewhat problematic for more conservative, bounding approaches of risk assessment.

Suggest language that the comprehensive risk metrics ensure risk is controlled with sufficient margin to risk performance objectives. Getting an accurate approximation of risk is less important than ensuring that risk is controlled under acceptable levels.

Comments on NRCs Proposed Part 53 Page 8 of 123 Section Section Desc.

Comment Suggested Resolution Subpart B Defense In Depth (pg 34)

The DID discussion focuses on DBEs other than DBAs.

While this aligns with LMP, a bounding risk assessment can also inform a defense in depth assessment. For example, the systems credited in the DBA perform a mitigation function, systems that fail to create the initial conditions of the DBA perform a prevention function. DID can be assessed from that set of information to inform functional criteria.

Suggest removing other than DBAs from the discussion of DID.

Subpart C 53.450(e)

Part 53 is meant to be technology inclusive, so for a LWR, the traditional approach of identifying DBAs should be sufficient for licensing purposes. Guidance such as the SRP could describe how the traditional approaches meet the regulatory requirements of Part 53.

53.450(e) on LBEs should allow PRA or other generally accepted approaches and not PRA in combination with other generally accepted approaches. Alternatively, the phrase risk evaluation would be more flexible.

Subpart C 53.440(d)

The Integrity Assessment Program (IAP) requirements remain duplicative. See Enclosure 3 for our response to the NRC Request for public comment on this topic.

Remove the duplicative requirements.

Subpart C 53.440(j)

While the more flexible language in 53.440(j) is appreciated, the requirement is in conflict with 53.110 since a commercial aircraft impact would result from (a) Attacks and destructive acts, including sabotage, directed against the facility by an enemy of the United States, whether a foreign government or other person.

Remove 53.440(j).

Subpart C 53.450 The discussion in this section is good, but Existing processes for defining the scope and capability of a PRA Remove the language in 53.450(a) requiring internal and external hazards to be addressed

Comments on NRCs Proposed Part 53 Page 9 of 123 Section Section Desc.

Comment Suggested Resolution supporting an application offer flexibility in determining the degree to which the PRA needs to be developed and may be informed by other factors such as design complexity and the needed degree of realism and level of detail, consistent with the use of the PRA and substance of the application. Such processes are currently available for appropriately defining the scope of the PRA and determining applicability of supporting requirements in consensus PRA standards needed to satisfy the proposed regulatory requirements for the specific uses of analyses under § 53.450(b). is in contradiction with the requirement in 53.450(a) for the PRA to assess susceptibility to internal and external hazards.

The regulation needs to reflect the guidance that An NRC determination of the acceptability of a PRA includes but is not limited to assessing the initial and boundary conditions and key assumptions used in the analysis, treatment of uncertainties, and the use of screening tools and bounding or simplified methods for any mode or hazard, provided the use of those tools and methods is justified by an acceptable technical basis.

by the PRA. Also remove the requirement that a PRA in combination with other evaluations need to consider all plant operating states.

Assessment of lower modes of operation should be acceptable based on traditional methods. This is already accepted in RG 1.253 which does not require anything more than an Full Power Internal Events (FPIE) PRA for a CP application. To be consistent with the SRM, RG 1.253 should be updated to allow supplemental evaluations for all licensing applications if justified in line with the non-LWR PRA Standard. This is particularly important for LWRs where there is no consensus LPSD standard Subpart C 53.450 (e)

The DID discussion focuses on DBEs other than DBAs.

While this aligns with LMP, a bounding risk assessment can also inform a defense in depth assessment. For example, the systems credited in the DBA perform a mitigation function, systems that fail to create the initial conditions of the DBA perform a prevention function. DID Suggest removing other than DBAs from the discussion of DID.

Comments on NRCs Proposed Part 53 Page 10 of 123 Section Section Desc.

Comment Suggested Resolution can be assessed from that set of information to inform functional criteria.

Subpart H 53.1309 The PRA discussion Probabilistic risk assessments developed for commercial nuclear plants prior to construction would be based on the design and other information available at the time of the CP application.

PRAs performed in early design stages or prior to construction may be inherently less detailed and may include projected information that will be subsequently verified or revised when the plant is built is helpful, but should provide more information. For example, consistent with RG 1.253, guidance should be clear that only a FPIE PRA is required at the CP stage.

Suggest adding discussion similar to RG 1.253 for PRA acceptability at the CP stage.

Detailed Comments on Subpart A - General Provisions Comment Section Desc.

Comment Suggested Resolution A-1 53.000 Purpose The NRC requirement is similar to the scope requirements in 50.1 and 52.0. However, the NRC is restricting the applicability of Part 53 only to commercial nuclear plants licensed under AEA Section 103. It is recognized that the NRC made changes in the definition of commercial nuclear plant in the proposed 53.020 to enable applying part 53 to accelerator driven systems that use special nuclear material (SNM) but that do not involve self-Expand applicability to all production and utilization facilities licensed under AEA Section 103 or 104, for clarity and consistency with the scope of 50.1 and 52.0. Please modify 53.000 to read: This part provides for the issuance, amendment, renewal, and termination of licenses, permits, certifications, and approvals for production and utilization facilities licensed

Comments on NRCs Proposed Part 53 Page 11 of 123 Comment Section Desc.

Comment Suggested Resolution sustaining chain reactions. While this is a useful change, restricting the applicability of Part 53 to commercial nuclear plants licensed under AEA Section 103 is unnecessarily restrictive.

As discussed Enclosure 3, a class 104 framework is preferable for an operating license to enable factory testing of a fueled microreactor.

Furthermore, the creation of a new term, commercial nuclear plants, as the sole applicability of the rule, instead of using the terms of production and utilization facilities that define the applicability of Parts 50 and 52 lead to confusion on how Part 53 is an optional alternative to Parts 50 and 52. While the definition of a commercial nuclear plant in 53.020 states that for the purposes of requirements in this part that reference requirements in part 50 of this chapter, a commercial nuclear plant is equivalent to a nuclear power plant is helpful, it does not adequately resolve the concern that the purpose statement should address production and utilization facilities to be clear how Part 53 is an optional alternative to Parts 50 and 52. The inclusion of a definition for a utilization facility 53.020 tends to compound the problem rather than add clarity.

under Section 103 and 104 of the Atomic Energy Act of 1954, as amended (AEA) (68 Stat. 919) and Title II of the Energy Reorganization Act of 1974, as amended (ERA)

(88 Stat. 1242). A plant may be licensed and regulated under Part 53, instead of Part 50 or Part 52 at the election of the applicant or licensee of a production or utilization facility.

(Conforming changes throughout Part 53 may be necessary).

The definition of commercial nuclear plant in 53.020 also should be modified for consistency with this Purpose statement, and the definition of utilization facility should be deleted.

A-2 53.020 Definitions -

NRC has developed an incomplete definition of a consensus code or standard, and specifically has not included all of the important considerations in a Standards Development Organization being accredited by the Remove definition of Consensus code or standard.

Comments on NRCs Proposed Part 53 Page 12 of 123 Comment Section Desc.

Comment Suggested Resolution Consensus code or standard American National Standards Institute (ANSI) as an Accredited Standards Developer. The ANSI defined essential requirements for Due Process in developing a consensus standard are: openness, lack of dominance, balance, coordination and harmonization, notification of standards development, consideration of views and objections, consensus vote, appeals, written procedures, and compliance with normative American National Standards policies and administrative procedures.

ANSI Accredited Standards Development Organizations develop and publish consensus standards. Of course, NRC can review and adopt other standards, but in pointing to consensus standards NRC is making reference to a standard developed under a very specific set of requirements. NRC should not attempt to develop its own definition of consensus standards that is not consistent with the nationally recognized process and then imbed that definition in a federal regulation.

A-3 53.020 Definitions -

Probabilistic Risk Assessment The NRC definition probabilistic risk assessment (PRA) is not consistent with how this term is defined on the NRCs website. A change to this definition is needed to improve clarity and predictability. This is particularly important because there are different interpretations of how PRA are used, as discussed in NRC meetings on Part 53.

Change the definition to align with current NRC definition of this term as follows (from NRCs website):

Probabilistic risk assessment (PRA) is a systematic method for assessing three questions that the NRC uses to define "risk."

These questions consider (1) what can go wrong, (2) how likely it is, and (3) what its consequences might be. These questions allow

Comments on NRCs Proposed Part 53 Page 13 of 123 Comment Section Desc.

Comment Suggested Resolution the NRC to understand likely outcomes, sensitivities, areas of importance, system interactions, and areas of uncertainty, which the staff can use to identify risk-significant scenarios. The NRC uses PRA to determine a numeric estimate of risk to provide insights into the strengths and weaknesses of the design and operation of a nuclear power plant.

A-4 53.020 Definitions -

Programmatic controls It is important to define the purpose for the types of technical features (design features, human actions and programs) that are needed for nuclear facilities.

The NRC has introduced a new term programmatic controls, however, the definition and the application of the term in the Part 53 requirements is vague and subjective. Thus, there is no clarity or predictability in what would constitute acceptable programmatic controls.

This is concerning since there are over 20 Part 53 regulations that have open ended requirements for programmatic controls.

Furthermore, the concept of programmatic controls effectively duplicates the concept of programs which is not defined in NRC regulations but is reasonably well understood by virtue of its long history of use by the NRC and licensees. However, the NRC does not define this The term and definition for programmatic controls should be deleted and the use of the term in other instances in Part 53 should be deleted.

The term program should be defined so that the purpose of programs can be understood, as follows:

Programs are the administrative measures and controls that are relied upon by the NRC to provide reasonable assurance that plant design, construction, maintenance and operation meet the safety criteria in 53.210 and 53.220 for the lifetime of the plant.

Programs may apply to design features and/or credited human actions. Programs that require NRC approval are specified in the regulations for various technical areas (e.g., QA).

Comments on NRCs Proposed Part 53 Page 14 of 123 Comment Section Desc.

Comment Suggested Resolution term in Part 53, although it is used for over 24 required programs.

The concept and term of programmatic controls is not needed and is duplicative of the term program.

A-5 53.020 Definitions -

Small modular reactor

§ 53.020, Definitions, defines an SMR as follows:

Small modular reactor means a power reactor, which may be of modular design as defined in

§ 52.1 of this chapter, licensed under this part to produce heat energy up to 1,000 megawatts thermal per module.

This definition is essentially the same as the SMR definition found in §50.2, Definitions, including the reference to a reactor thermal power level that serves as the upper bound for applicants and licensees considering use of the performance-based, technology-inclusive, risk-informed, and consequence-oriented EP framework found in §§ 50.33, 50.34, and 50.160. The above definition aligns with the upper limit on electric power generation typically used to define SMRs, i.e., ~300 MWe per module. Both of these values, 300 MWe and 1,000 MWt, are derived from early definitions of an SMR, usually from the U.S.

Department of Energy or the International Atomic Energy Agency. While these power levels may be suitable for some regulatory purposes (e.g., setting of fees), a reactor power level by itself does not determine the safety, The NRC should reconsider and remove the power level criterion that prohibits the use of the alternative performance-based EP requirements by SMR facilities producing greater than 300 MWe or 1,000 MWt per module.

Comments on NRCs Proposed Part 53 Page 15 of 123 Comment Section Desc.

Comment Suggested Resolution security, and accident consequence characteristics of a given design.

A-6 53.024 Definitions -

Safety-related, Non-safety-related but safety significant (NSRSS), Special Treatment, and Non-safety-significant The NRC is applying special treatment requirements to SR SSCs and to certain NSRSS SSCs but does not define the differences in the special treatments between these types of SSCs. 53.460 provides the only clear difference in requirements, that SR SSCs require Appendix B QA, while NSRSS SSCs may optionally follow Appendix B. Other requirements, such as 53.440(b) imply equivalent C&S for SR and NSRSS SSCs. Thus, it is possible, or even likely, that NSRSS SSCs will receive an equivalent regulatory burden as SR SSCs. The application of the term special treatment to SR and NSRSS SSCs reduces regulatory predictability and in fact is not necessary.

While the NRC appears to be applying the term special treatment to describe how Part 53 requirements apply to SR and NSRSS SSCs, such an approach was not necessary in Parts 50 and 52. The NRCs Part 53 definition of special treatment effectively says that it is requirements that apply to certain SSCs, and thus provides no clarity or regulatory stability. The approach in Parts 50 and 52 is to state within specific requirements whether they apply to safety-related or risk-significant SSCs. The same can be done in Part 53, where the requirements define to which category of SSCs they apply, and in fact most Part 53 requirements already do this. Thus, the use of the term Provide a clear definition of risk-significant functions as it applies in the definition of special treatment and non-safety-related but safety significant.

This definition must lead to a clear distinction among special treatment requirements for SR and NSRSS SSCs.

Comments on NRCs Proposed Part 53 Page 16 of 123 Comment Section Desc.

Comment Suggested Resolution special treatment as used in the proposed Part 53 creates confusion and provides no regulatory benefit.

A-7 53.020 Definitions

- Defense in Depth The definition of defense in depth is not needed, since a specific requirement is not needed to achieve defense in depth. If the term and definition is retained, then the NRC proposed language should be revised to avoid creating unintended consequences through the prescriptive nature of the definition.

Delete the term and definition. If the term is not deleted, then the definition should be revised as follows:

Defense in depth is a design philosophy that provides reasonable assurance that the design meets the safety criteria in 53.210 over the life of plant by addressing uncertainties in the performance of safety functions through measures such as increased safety margin and multiple layers of protection means inclusion of multiple independent and redundant layers of defense in the design of a facility and its operating procedures to compensate for potential human and mechanical failures so that no single layer of defense, no matter how robust, is exclusively relied upon. Defense-indepth includes, but is not limited to, the use of access controls, physical barriers, redundant and diverse key safety functions, and emergency response measures.

A-8 53.020 Definitions -

Anticipated Event

Sequence, There is some confusion with these terms. Deviation from long-standing terms DBE and BDBE is unwarranted We recommend that the NRC replace the terms and definitions with the following:

Comments on NRCs Proposed Part 53 Page 17 of 123 Comment Section Desc.

Comment Suggested Resolution Licensing Basis Events, Unlikely Event Sequences, Very Unlikely Event Sequences, and Design Basis Accidents and is likely to lead to misunderstanding and misuse of the new terms.

Licensing basis events (LBEs) are unplanned events and include AOOs, DBAs, and BDBEs that are considered in the licensing of a production or utilization facility. LBEs may include one or more reactor modules.

Anticipated operational occurrences (AOOs) are a grouping of similar event sequences that are unplanned but may occur one or more times during the life of a nuclear facility. AOOs established through quantitative methods are event sequences with a mean frequency of 1x10-2/plant-year and greater. AOOs take into account the expected responses of all SSCs within the plant, regardless of safety classification. (if needed)

Design basis accidents (DBAs) may be derived from the DBEs and are used to establish the design of safety-related SSCs. DBAs take into account the expected responses of only those safety-related SSCs relied upon to mitigate or prevent event sequences.

A-9 53.020 Definitions - Site characteristics The definition of site characteristics was changed from the definition in the preliminary rule language. While the change is an improvement, including the licensing and permitting documents in which site characteristics appear does not contribute to a useful working definition.

Revise the definition as follows:

Site characteristics means the meteorological, geological, seismological, topographical, hydrological, and other characteristics of the site and surrounding

Comments on NRCs Proposed Part 53 Page 18 of 123 Comment Section Desc.

Comment Suggested Resolution area that may have a bearing on the consequences of a radionuclide active release material escaping from the nuclear plant as well as demographic features of a site. (§ 53.500).

A-10 53.020 Definitions -

Functional Design Criteria Functional design criteria in Part 53 serve the same underlying purpose as principal design criteria in Parts 50 and 52. Since PDC concept has a long regulatory precedent and is well understood in the design and licensing of nuclear facilities, guidance should provide clarity that the GDC and ARDC in RG 1.232 are valid means of identifying FDC under Part 53.

NEI appreciates 53.410, 53.420, 53.425 and 53.430 providing a clearer scope of the design criteria, thus allowing PDC development in line with RG 1.233, however applicants should have the option of developing design criteria in line with RG 1.232, RG 1.233 or the GDC.

Finally, while the definition of PDC in Part 50 is focused on being met solely by SSCs, the actual PDC are formed around the concept that they may be met by SSCs, human actions, or programs, or a combination thereof.

Change the term and definition as follows:

Functional design criteria means metrics for the performance of functions supported by SSCs, programs and operator actions. For SR SSCs functions, these criteria define performance metrics necessary to demonstrate compliance with the safety criteria in § 53.210. For NSRSS SSCs functions, these criteria define performance metrics necessary to demonstrate compliance with the safety criteria in § 53.220, 53.260 and 53.270.

A-11 53.020 Definitions -

Commercial Nuclear Reactor We believe that to meet the intent of the ADVANCE Act and address microreactor licensing issues around factory fueling and testing of a microreactor and transportation of a fueled microreactor, the NRC should change policy to Remove designed or from the 53.020 definition of a commercial nuclear reactor.

Comments on NRCs Proposed Part 53 Page 19 of 123 Comment Section Desc.

Comment Suggested Resolution

& Utilization facility define a utilization facility only after Part 70 requirements to preclude criticality are removed. This builds off of SECY-24-0008 and is described in more detail in Enclosure 3, Topic 7. To facilitate this policy change, we suggest removing designed or from the 53.020 definition of a commercial nuclear reactor (and the utilization facility definition which references the commercial nuclear reactor definition). This leaves Commercial nuclear reactor means an apparatus, other than an atomic weapon, used to sustain nuclear fission. For the purposes of requirements in this part that reference requirements in 10 CFR part 50, a commercial nuclear reactor is equivalent to a nuclear reactor as defined in 10 CFR 50.2.

A-12 53.020 Definitions -

Licensing-basis events Part 53 defines an LBE as a collection of event sequences considered in the design and licensing The word considered is vague as to distinguishing between those LBEs that were initially evaluated but did not need to be included in the licensing basis as compared to those that were evaluated and retained for the licensing basis. Only a subset of potential LBEs may be retained to support the licensing basis.

Revise 53.020 definition of licensing-basis events as follows:

Licensing-basis events means a collection of event sequences relied upon to support considered in the design and licensing of the commercial nuclear plant. Licensing-basis events are unplanned events and include anticipated event sequences, unlikely event sequences, very unlikely event sequences, and DBAs.

Comments on NRCs Proposed Part 53 Page 20 of 123 Detailed Comments on Subpart B - Technology Inclusive Safety Requirements Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution B-1 53.210 Safety Criteria for Design Basis Accidents This requirement uses language that is prescriptive, yet open-ended, rather than using performance-based language that is clear and measurable, thereby resulting in reduced regulatory predictability and flexibility.

Specifically, the phrase Design features and programmatic controls must be provided prescribes the features rather than defining the desired outcome. This phrase does not add clarity because other requirements, specifically 53.400 for design features, and numerous requirements related to programs already specify that these elements are needed for Part 53 applicants and licensees. Thus, duplication of this phrase here and in many other locations in Part 53 will add regulatory burden, in terms of demonstrating compliance, without any benefit to safety.

Revise the requirement to be more performance-based, by deleting the phrase Design features and programmatic controls must be provided for each commercial nuclear plant such that identification and analyses of must analyze design-basis accidents (DBAs) in accordance with § 53.240 to demonstrate the following B-2 53.220 Safety Criteria for Licensing Basis Events Other than Design Basis Accidents 53.220 provides safety criteria for licensing-basis events other than design -basis accidents. However, by referencing 53.240, 53.220(a) brings in 52.240(c) which explicitly requires analysis of DBAs and analysis of LBEs to meet the criteria of 53.210 (safety criteria for design-basis accidents). 53.240(a) also references 53.450 broadly which could imply 53.220 is meant to cover safety criteria for aircraft impact assessments and fire protection. It is unclear if this was the intent.

Modify 53.220 to read:

Design features and programmatic controls classified as NSRSS This change emphasizes the need to remove reference to SR safety criteria (53.240) from 53.220.

Suggest removing the reference in 53.220 and 53.220(a) to 53.240 so that safety criteria

Comments on NRCs Proposed Part 53 Page 21 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution 53.220 also makes no reference to the relative importance of 53.220 in comparison to 53.210. In line with the NEI 18-04 guidance endorsed in RG 1.233, suggest adding classified as NSRSS to clarify that design features and programmatic controls to meet 53.220 are not required to be Safety Related.

It is assumed that the intent, for those following LMP, is for 53.220 to cover the F-C chart and to address defense in depth in line with 53.250. Applicants seeking to use Part 53 outside of the LMP methodology could propose their own criteria for AOOs (potentially relying on Part 20 requirements), BDEs (potentially stricter safety criteria and associated design criteria more in line with traditional DBAs and associated requirements) and BDBEs (potentially following RG 1.226 and more traditional EP requirements).

for licensing basis events other than design basis accidents do not reference requirements for design basis accidents.

Provide clarity on whether 53.220 was intended to cover criteria from 53.450(e) or 53.450 broadly (with the exception of 53.450(f) which should clearly be covered by 53.210).

One option is to consider modifying 53.220(a) as follows:

(a) Plant SSCs, personnel, and programs provide the necessary capabilities and maintain the necessary reliability to address LBEs other than DBAs in accordance with §§ 53.240 and 53.450(e), and provide measures for defense in depth in accordance with § 53.250; and If references remain, they should be clear in scope, probably limited to 53.450(e) and 53.250.

B-3 53.220 Safety Criteria for Licensing Basis Events Other Consistent with NEIMA and NEI 18-04 as endorsed by RG 1.233, the comprehensive risk metrics should be a piece of an integrated safety assessment. As discussed in the NRC documents that describe the DID philosophy, layers and DID attributes play a significant role in the approach Replace 53.220 as follows:

53.220 Safety Criteria for an Integrated Safety Assessment

Comments on NRCs Proposed Part 53 Page 22 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution than Design Basis Accidents to DID capability. However, there do not exist any well-defined regulatory acceptance criteria for deciding the sufficiency of the DID for nuclear power plant licensing or operation.

We suggest 53.220 to be modified to not have separate requirements to assess risk but assess risk in an integrated manner. This is more in line with the RG 1.233 guidance and the language in the preamble: It is worth noting that the evaluation of plant risks, as represented by a comparison of analysis results to acceptable risk performance objectives for comprehensive risk metrics, would be one of several performance standards used in subpart B. The proposed use of multiple performance standards, including deterministic criteria and defense-in-depth measures, reflects an integrated decision-making process The NRCs approval of using a comprehensive risk metric or set of metrics with associated risk performance objectives is not, by itself, an indicator of adequate protection. Rather, the comparison of comprehensive risk metrics to associated risk performance objectives that are acceptable to the NRC is part of a suite of regulatory requirements that, when considered holistically, form the basis for the NRCs decision-making. This is analogous to the approach used for plants licensed under part 50 and part 52, where no Design features and programmatic controls for NSRSS SSCs must be provided for each commercial nuclear plant to assure adequate protection of the public health and safety.

This is achieved through an integrated safety assessment which must consider the necessary capabilities and reliability of design features and programmatic controls to address LBEs in accordance with 53.450(e),

provide measures for defense in depth in accordance with § 53.250; and evaluate residual risk.

Comments on NRCs Proposed Part 53 Page 23 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution single regulatory requirement governs whether a plant is safe enough.

This meets the intent of the SRM by providing a risk-informed approach where the NRCs approval of the metric or set of metrics is not, by itself, an indicator of adequate protection.

B-4 53.230 Safety Functions It is unclear what purpose is served by defining primary and alternative safety functions, other than to justify including the comprehensive risk metrics in the rule language in 53.220. Does the NRC intend that some safety functions (primary) are only needed to meet the standard of reasonable assurance of adequate protection of public health and safety and other safety functions (additional) are needed to meet the standard to protect health or to minimize danger to life or property? While such distinction could make Part 53 more efficient, it is not clear that this is the NRCs intent since these standards were replaced in 53.210. Safety functions should relate to the DBAs and requirements for safety-related SSCs. Any other functions the plant needs to provide may be safety significant but would not be safety-related.

The statement that the safety functions must be maintained for all licensing basis events implies that there is an equivalence in the design standards for SR SSCs needed for the DBAs to meet 53.210, and the Revise the requirement for greater clarity, as follows:

(a) The primary safety function is limiting the release of radioactive materials from the facility and must be maintained during normal operation and for licensing basis events over the life of the plant.

(b) Additional safety functions necessary to meet 53.210--must be identified for each commercial nuclear plant needed to support the retention of radioactive materials during licensing basis eventssuch as controlling reactivity, heat generation, heat removal, and chemical interactions are examples of possible safety functions depending on the specific technology and design.

(c b) The primary and additional SR SSCs must be capable of performing their intended

Comments on NRCs Proposed Part 53 Page 24 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution NSRSS and NSS SSCs that are relied upon for AOOs and BDBEs. This could result in unintended consequences in that the NSRSS and NSS SSCs are elevated to need similar confidence in performance as SR SSCs. Such an outcome would increase regulatory burden without an increase in safety.

This is particularly important because many Codes and Standards rely on the term safety function as traditionally understood to mean Safety Related functions. Expanding the definition of safety function to NSRSS will cause confusion and overprescription of requirements.

Examples in regulation may be interpreted as minimum requirements. Suggest language that makes clear that safety functions are design specific.

safety functions. NSRSS SSCs may be relied upon to accomplish the safety functions for other LBEs, are required to satisfy the safety criteria defined in §§ 53.210 and 53.220, or more restrictive alternative criteria adopted under § 53.470, and must be fulfilled by the design features human actions and programmatic controls specified throughout this part.

B-5 53.240 Licensing Basis Events This requirement provides valuable clarity in terms of the types of events that must be considered in the design and licensing of a nuclear facility. The requirement also provides clarity in referencing other requirements that interface with the selection of licensing basis events. However, the requirement does include language that duplicates requirements in other parts of the rule, and such duplication should be avoided as it could lead to unintended consequences and Revise the requirement as follows:

(a) Licensing-basis events must be identified for each commercial nuclear plant and analyzed under § 53.450 to demonstrate that the safety requirements in this subpart have been satisfied.

(b) The identified LBEs, ranging from anticipated event sequences to very unlikely event sequences, must collectively address

Comments on NRCs Proposed Part 53 Page 25 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution increased regulatory complexity and burden, without an increase in safety.

It's also important to ensure that only relevant or appropriate combinations of hazards are considered. As an example, it is longstanding PRA practice to only consider single initiators: must collectively address combinations of malfunctions of plant SSCs, human errors, facility hazards, and the effects of external hazards. without relevant or appropriate could imply a requirement to consider an internal events initiating event coincident with an independent external hazards initiator.

appropriate combinations of malfunctions of plant SSCs, human errors, facility hazards, and the effects of external hazards.

(c) The analysis of LBEs must (1) Include analysis of one or more DBAs under § 53.450(f);

(2) Confirm the adequacy of design features and programmatic controls needed to satisfy the safety criteria defined in §§ 53.210 and 53.220, or more restrictive alternative criteria adopted under § 53.470, and (32) Establish related functional requirements for plant SSCs, personnel, and programs.

B-6 53.250 Defense in Depth A specific requirement in Part 53 for defense in depth (DID) is not necessary in order to achieve defense in depth. As an example, Parts 50 and 52 do not contain a requirement for defense in depth but do achieve the desired outcome by applying a DID philosophy.

The NRCs proposed DID requirement is prescriptive and is not performance-based or risk-informed. For example, the NRC prescribes that no single feature no matter how robust should be exclusively relied upon. However, Delete the requirement. If the requirement is not deleted, then it should be revised as follows:

(a) Measures must be taken for each commercial nuclear plant to ensure appropriate Defense in depth is must be provided to compensate for uncertainties in the analysis of the safety criteria such that there is reasonable assurance that the safety

Comments on NRCs Proposed Part 53 Page 26 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution what if the consequences of a particular design that did not protect against a single failure were less than 1 rem?

In this case the single failure protection is not needed to meet the safety criteria in Part 53.

Furthermore, the requirement is written in a way that it is applicable to NSRSS and NS SSCs, because it applies to 53.220. However, is not the single failure more effective when it is applied to SR SSCs relied upon for DBAs in meeting 53.210? Thus, the requirement is adding regulatory complexity and burden, without an increase in safety.

Thus, this prescriptive application of a single failure protection is inconsistent with the basis for the Commission direction in SRM-SECY 19-0036 to not apply deterministic criteria when they are not needed based on risk insights.

This requirement uses language that is prescriptive, yet open-ended, rather than using performance-based language that is clear and measurable. Consequently, this requirement reduces regulatory predictability and flexibility. Specifically, the phrase Measures must be taken prescribes the feature rather than defining the desired outcome.

criteria in this subpart are met over the life of the plant.

(b) The uncertainties that must be addressed under paragraph (a) of this section include those related to the state of knowledge and modeling capabilities, the ability of barriers to limit the release of radioactive materials from the facility during LBEs other than DBAs, the reliability and performance of plant SSCs and personnel, and the effectiveness of programmatic controls.

(c) Defense in depth measures may include increased safety margin and redundant layers of protection. The safety analysis may not rely upon a single engineered design feature, human action, or programmatic control, no matter how robust, to address the range of LBEs other than DBAs.

Comments on NRCs Proposed Part 53 Page 27 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution B-7 53.260 Normal Operations A requirement for ALARA is not necessary since 53.260 is redundant with the dose standard for normal operations (0.1 rem) in Part 20. ALARA is also already achieved by operational considerations through Part 20, which is applicable to all Part 53 licensees even if it is not explicitly stated in Part 53.

Delete the requirement 53.260 for a requirement for ALARA.

B-8 53.270 Protection of Plant Workers The requirement for ALARA in 53.270 is not necessary since 53.270 is redundant with the dose standard for occupational doses in Part 20. ALARA is also already achieved by operational considerations through Part 20, which is applicable to all Part 53 licensees even if it is not explicitly stated in Part 53.

Delete the requirement 53.270 for a requirement for ALARA.

Comments on NRCs Proposed Part 53 Page 28 of 123 Detailed Comments on Subpart C - Design and Analysis Requirements Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution C-1 53.425 Design features and functional design criteria for normal operations The proposed definition of Commercial nuclear plant in Section 53.020 may result in the requirements outlined in Section 53.425 being more restrictive and burdensome than the existing requirements for the design and control of normal operational effluents under 10 CFR Parts 50 and

52. To ensure regulatory consistency and avoid unnecessary burdens, further clarification is recommended regarding how these requirements align with those under the current licensing frameworks.

In the proposed Part 53 rule, Commercial nuclear plant is defined as:

A facility consisting of one or more commercial nuclear reactors and associated co-located support facilities, including the collection of buildings, radionuclide sources, and SSCs for which a license, certification, or approval is being sought under this part, that is or will be used for producing power for commercial electric power or other commercial purposes. For the purposes of requirements in this part that reference requirements in Part 50 of this chapter, a commercial nuclear plant is equivalent to a nuclear power plant.

Under this definition, the requirements in Section 53.425, for the design criteria for normal operations, adopt a site-wide approach rather than the per-reactor basis utilized in the current licensing framework under Parts 50 and 52.

Revise the requirements in Section 53.425 to apply on a per-reactor and/or per-license basis rather than a per-site basis.

Comments on NRCs Proposed Part 53 Page 29 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution Specifically, 10 CFR Part 20 establishes requirements on a per-licensee basis, while 10 CFR Part 50, Appendix I, applies design criteria on a per-reactor basis. Additionally, it is important to note that, 10 CFR 20.1301(e) (which references 40 CFR 190) already establishes limits that consider the impacts of effluents from the entire fuel cycle, rather than on a per-reactor or per-licensee basis. Given that Parts 50 and 52 remain as licensing options for applicants, it appears unnecessary for Section 53.425 to adopt a site-wide approach, which could impose overly restrictive constraints from a design perspective for effluents during normal operations.

This shift to a Commercial nuclear plant approach could impose additional burdens on designers and operators, especially when multiple reactors are planned for a single site. Designers may be required to implement additional design changes for subsequent reactors sited after the initial reactor, potentially complicating compliance efforts and increasing costs.

This comment is not intended to imply the site-wide approach being proposed for accident analyses is inappropriate. This comment is limited to the Part 20 requirements referenced in 53.425.

C-2 53.410 Functional Design Criteria for It is noted that the Part 53 functional design criteria serve a similar purpose as principal design criteria in Parts 50 and 52. NEI appreciates the effort to limit Revise the 53.410 requirement as follows:

Comments on NRCs Proposed Part 53 Page 30 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution Design Basis Accidents.

53.420 Functional Design Criteria for Licensing Basis Events other than Design Basis Accidents.

53.425 Design Features and Functional Design Criteria for Normal Operations 53.430 Design Features and Functional Design Criteria for protection of plant workers.

functional design criteria appropriately and relegate special treatments to other requirements. However, the NRC should make clear in guidance that the ARDC (RG 1.232) and the GDC may be used to identify functional Design Criteria. The X-Energy PDC Topical provides precedent on which PDC are functional and should be FDC vs special treatment and would not be an FDC under Part

53.

The NRC establishes four requirements related to functional design requirements (53.410, 53.420, 53.425 and 53.430). The language in each of these is nearly identical, and thus the repetition and duplication reduces regulatory clarity and predictability, and increases regulatory complexity and burden without an increase in safety. This is exacerbated by other requirements not making distinction in the special treatments for FDC required to support SR SSCs vs NSRSS SSCs (see comments C-5 and C-11).

53.410(b) is not necessary as the requirements are appropriately addressed in Subpart F making 53.410(b) duplicative.

(a)Functional design criteria must be defined for each design feature required by § 53.400 relied upon to demonstrate compliance with the safety criteria defined in § 53.210, 53.220, 53.260 and 53.270. (b)Corresponding programmatic controls and interfaces must be established in accordance with this and other subparts to achieve and maintain the reliability and capability of SSCs relied upon to meet the established functional design criteria and the safety criteria required in § 53.210 and to maintain consistency with analyses required by

§ 53.450.

Change the name of the requirement, to align with the hierarchical flow of design features and the functional design criteria, as follows:

53. 410 Functional Design Criteria for Design Basis Accidents Delete 53.420, 53.425 and 53.430, as the requirements were incorporated into the proposed revision.

Update RG 1.232 with guidance on how the ARDC may be used to define FDC.

Comments on NRCs Proposed Part 53 Page 31 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution If the above approach is not taken, the 53.420 requirement should be modified as follows:

Functional design criteria must be defined for each design feature classified as NSRSS, Similar changes should be made to 53.425 and 53.430.

RG 1.233 and RG 1.253 make a distinction between safety criteria and associated design criteria to meet the DBA requirements and safety criteria and associated design criteria to meet the non-DBA LBE requirements. This clarification in the rule language helps applicants understand and apply the appropriate level of rigor to different classifications of components and their associated requirements.

C-3 53.425 (c) Design features and functional design criteria for normal operations The footnote in this section indirectly establishes a design objective of 10 mrem for maintaining public doses as ALARA. However, there is currently no guidance available on how this design objective can be achieved.

For clarity and flexibility, the NRC staff should delete this footnote and instead develop and provide guidance that documents what could be considered as a reasonable If 53.425 is kept, delete the footnote and develop new guidance or update existing guidance to document how a reasonable design objective can be developed and achieved.

Comments on NRCs Proposed Part 53 Page 32 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution design objective, including flexible methods or criteria for meeting it.

C-4 53.440, 53.450 and more Chairman Wright in his voting record proposed a higher-level rule with details provided in regulatory guidance. We see merit in that approach as well as efficiency gains in the rulemaking process. 53.210 through 53.430 provide high level objectives and considerations. Design features and functional design criteria provide criteria for SSCs. Much of the following rule language provides details on how analysis can show that the safety criteria can be met or how design requirements can show the design criteria are met. We suggest removing much of this detail from the rule and placing the detail in guidance consistent with the voting record of Chairman Wright.

Delete 53.440 and develop sufficient regulatory guidance to show that the pertinent design requirements are an acceptable means of meeting the design criteria in 53.400 to 53.430. Conforming changes would be required to much of the rule language.

Delete 53.450 and develop sufficient regulatory guidance to show that the pertinent analysis requirements are an acceptable means of meeting the safety criteria in 53.210 to 53.270. Conforming changes would be required to much of the rule language.

Look for opportunities throughout the rule to reduce rule detail and move that information to guidance.

C-5 53.440 Design Requirements.

The proposed 53.440 includes multiple design requirements that duplicate requirements elsewhere in the regulations. Specifically:

The requirement (c), to qualify SSCs for their service conditions over the design life of the SSC, contributes to eliminating the need for an Integrity Assessment Program Revise the requirement as follows:

Combine proposed (c) and (d) to read as (c)

The materials used for safety related SSCs must be qualified for their service conditions over the plant lifetime. Qualification must consider possible degradation mechanisms related to

Comments on NRCs Proposed Part 53 Page 33 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution in 53.870. The requirement (d), to evaluate possible degradation mechanisms over the plant lifetime can be reassigned to provide further clarification for the need to qualify materials in (c). These requirements also increase the traditional scope of qualification beyond SR SSCs.

While qualification might be an appropriate method for ensuring the design criteria for NSRSS SSCs are met, monitoring may be adequate. Therefore, qualification should only be a requirement for SR SSCs and an option for NSRSS SSCs. Adequacy of NSRSS design can be determined in the individual review of applications.

The design requirements for fire protection are prescriptive, rather than performance-based, in that they mandate specific features of the design to address fire protection. However, fire hazards are already considered in the LBEs in 53.240, and thus the design would already need to have considered these in (a) of this requirement.

These requirements replicate the 10 CFR 50 Appendix A General Design Criteria 3 for Fire Protection.

The requirement (f), to consider safety/security interface, duplicates 73.58, which is significantly more detailed than the requirement here.

The requirement in (g) to achieve and maintain a subcritical condition is a fundamental aspect of evaluating service time, fatigue, chemical interactions, operating temperatures, effects of irradiation, and other environmental factors that may affect their performance.

Delete proposed 53.440(e) related to design features related to fire protection.

Delete 53.440(f) related to safety and security Remove 53.440(i) requirements related to fuel and radionuclides outside the reactor as it is covered implicitly in Subpart B and is explicitly required for analysis under 53.450(b)(4) and (5).

Remove 53.440(j) as it postulates an aircraft impact that results from an attack or destructive action. Per 53.110, licensees are not required to provide for design features to protect against such attacks.

Delete 53.440(k) but ensure chemical hazards are addressed in the EP plan consistent with the language in 70.22.

Delete 53.440(g), (h), (l), (m), and (n).

Comments on NRCs Proposed Part 53 Page 34 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution LBEs in accordance with 53.240. Duplicating this as a design requirement is unnecessary.

The requirement in (h) to provide long-term cooling of the reactor and waste stores is fundamental to meeting the safety criteria in 53.210 and 53.220. Duplicating this as a design requirement is unnecessary.

The requirements in 53.440(i) for considering fuel and waste appears to be in the wrong place in Part 53. These relate to understanding all of the sources of radionuclides that could be released to the environment, and are first order considerations in the safety paradigm, not a downstream consideration during the design requirements stage. They should be addressed in Subpart B, and not as a design requirement.

53.440(j) should be removed since it contradicts 53.110 that licensees are not required to provide for design features to protect against such attacks.

The requirement in 53.440(k) addressing chemical hazards goes beyond requirements in Part 70 and is not warranted.

Consistent with 70.22(i)(1)(ii), the chemical hazards should be addressed in the emergency plan responding to radiological hazards and should not be included as a design requirement.

Comments on NRCs Proposed Part 53 Page 35 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution The requirement in 53.440(l) duplicates the requirements in 20.1406 and should not be included as a design requirement in 53.440.

The requirements in 53.440(m) on criticality monitoring duplicate requirements in 70.24 and should not be included as a design requirement in 53.440.

The requirement in 53.440(m)(3) address operational requirements but it is noted that the Certificate of Compliance for the package or cask are the applicable requirements for fuel in that package or cask. Thus, the text in (m)(3) addressing operations in accordance with Parts 71 and 72, should be addressed in Subpart F, if at all.

Overall, the requirements in 53.440(m) duplicate other requirements and should not be included as design requirements in 53.440.

The requirements in 53.440(n) addressing human factors duplicate requirements in Subpart F and should not be included as design requirements in 53.440. Additionally, including individual elements of a human factors program within regulation (e.g., functional requirements analysis and function allocation) may prevent applicants from maintaining the state-of-the art as required. It is recommended that program elements be relocated to guidance and not within regulation.

Comments on NRCs Proposed Part 53 Page 36 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution Duplication of requirements reduces clarity and predictability and increases regulatory complexity and burden without an increase in safety.

C-6 53.440(b) Design Requirements -

Codes and Standards.

NEI appreciates the approach to have a high-level requirement for Codes and Standards (C&S) in Part 53 instead of incorporating by reference C&S into the regulation as is done in 10 CFR 50.55a. This approach is appropriate given the many types of reactors that will be licensed under Part 53 and the many different standards that will apply to specific types of designs.

However, the requirement to use NRC endorsed C&S is not performance-based and presents an implementation challenge. The requirement reads as requiring endorsement of a standard prior to its use in an application. Given the long timeframes for NRC endorsement of standards, this may be prohibitive for innovation. DG-1410 for example is pre-decisional, but being used as guidance in LMP submittals under Part 50.

This would arguably require an exemption under Part 53 until ASCE 43-19 is endorsed in RG 1.251. Part 53 effectively increases regulatory burden here and increases exemptions.

The scope of 53.440(b) is also problematic in a way that would make Part 53 unworkable in the near-term.

53.440(b) references 53.400 which in turn references 53.210, 53.220, 53.470 and 53.230. This effectively Revise the requirement as follows:

The design features required by § 53.400 must, wherever applicable, be designed using generally accepted consensus codes and standards that are sufficient to meet the design criteria defined under 53.410, 53.420, 53.425, and 53.430 have been endorsed or otherwise found acceptable by the U.S.

Nuclear Regulatory Commission (NRC).

NRC will still be able to endorse C&S to define what is acceptable in meeting this requirement, but those RGs will not unnecessarily slow down the licensing process.

At a minimum:

The Safety Related design features required by § 53.400 must, wherever applicable, be designed using generally accepted consensus codes and standards that have been endorsed or otherwise found acceptable by the U.S.

Nuclear Regulatory Commission (NRC).

Comments on NRCs Proposed Part 53 Page 37 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution increases the scope to SR and NSRSS SSCs whereas 50.55a was primarily SR SSCs. This is problematic since for many NSRSS SSCs there are no NRC-endorsed C&S other than the SR C&S. This strongly goes against the spirit of the Licensing Modernization Project (LMP) upon which Part 53 was developed. NEI suggests that 53.440(b) be limited to solely SR SSCs. NEI would be able to assess NSRSS Codes and standards in the context of 53.220, 53.400 and the rest of Part 53 without requiring a deviation every time an NSRSS system is designed to a standard not previously endorsed by NRC.

It would benefit both the NRC and industry to build a list of C&S approved for certain NSRSS functions; NEI cheers the effort NRC is undertaking under the ADVANCE Act to endorse commercial C&S. However this is a monumental effort and will never be complete as new Codes and Standards are introduced. Putting the barrier to the use of commercial C&S for NSRSS SSCs during their development and NRC review goes against LMP, NEIMA and the ADVANCE Act goals for a more flexible and efficient licensing process.

The language in 53.440(b) also lacks clarity. The design features required by § 53.400 must, wherever applicable, be designed using generally accepted consensus codes and standards is unclear on what the scope of design is. As an example, NQA-1 is arguably the only generally accepted consensus codes and standards used for quality Add language in the statements of consideration to describe the scope of the requirements in 53.440(b).

Comments on NRCs Proposed Part 53 Page 38 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution assurance during the design phase (see RG 1.28). 53.440(b) may effectively require NQA-1 for SR and NSRSS SSCs which directly contradicts 53.460(b) which states: The special treatments for NSRSS SSCs and special treatments for SR SSCs beyond those required under (b)(1) of this section may include meeting selected quality assurance requirements from appendix B of part 50 of this chapter when such treatment is needed to address performance requirements, equipment reliability, or uncertainties.

NEI believes that 53.460(b) more accurately captures the intent of the NRC, and this example demonstrates the problem with the current language in 53.440(b).

Clarification in the statements of consideration to describe the scope of 53.440(b) would be valuable and reduce regulatory uncertainty.

C-7 53.440(e) Fire Protection If 53.440(e) is kept (see comment C-5), limit the scope to SR SSCs which would be consistent with the guidance in RG 1.189 for safe shutdown equipment.

Requirements for NSRSS SSCs may flow from the risk assessment and design process, but should not be a blanket requirement for all NSRSS SSCs.

Revise 53.440(e) as follows:

(e)(1) Safety-related and NSRSS SSCs must be designed and located to minimize, consistent with other safety requirements in this part, the probability and effect of fires and explosions.

(2) Noncombustible and fire-resistant materials must be used wherever practical throughout the facility, particularly in locations with SR and NSRSS SSCs.

Comments on NRCs Proposed Part 53 Page 39 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution (3) Fire detection and fire suppression systems of appropriate capacity and capability must be provided and designed to minimize the adverse effects of fires on SR and NSRSS SSCs.

(4) Fire suppression systems must be designed to ensure that their rupture or inadvertent operation does not significantly impair the ability of SR and NSRSS SSCs to perform their safety functions to satisfy § 53.230.

C-8 53.450 Analysis Requirements.

(a)-(b)

NEI appreciates the increased flexibility described in the statements of consideration and the language of 53.450(e).

However, the language in 53.450(a) remains problematic and must change to align the 53.450(e), the statements of consideration and the SRM.

§ 53.450 Analysis requirements.

(a) Requirement to have a probabilistic risk assessment (PRA). A PRA of each commercial nuclear plant must be performed to identify potential failures, susceptibility to internal and external hazards, and other contributing factors to event sequences that might challenge the safety functions identified in § 53.230 and to support demonstrating that each commercial nuclear plant meets the safety criteria of § 53.220, or more restrictive alternative criteria adopted under § 53.470 Revise the PRA requirement as follows:

§ 53.450 Analysis requirements.

(a) Requirement to have a probabilistic risk evaluation assessment (PRA). A risk evaluation of each commercial nuclear plant must be performed to identify potential failures, susceptibility to internal and external hazards, and other contributing factors to event sequences that might challenge the safety functions identified in § 53.230 and to support demonstrating that each commercial nuclear plant meets the safety criteria of § 53.220, or more restrictive alternative criteria adopted under § 53.470.

Comments on NRCs Proposed Part 53 Page 40 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution While language in both the preamble and 53.450(e) suggest more flexibility should be provided, in line with the SRM directive, the 53.450(a) language clearly requires an all-hazard PRA. At a minimum, the language needs to change to PRA in combination with other generally accepted approaches for systematically evaluating engineered systems so that 53.450(a) is consistent with 53.450(e), the preamble discussion and the Commission directive in the SRM. NEI would prefer the language to require a risk evaluation consistent with Commissioner Caputos markup of Part 53. This would be in line with the ADVANCE Act language that microreactors should develop strategies and guidance for risk analysis methods, including alternatives to probabilistic risk assessments.

To help understand how alternatives to a PRA-led methodology can work with Part 53, NEI has attached providing a mapping for how existing or draft regulatory guidance for 4 methodologies can work with Part 53, including minor proposed changes to the Part 53 rule language to facilitate that methodological flexibility.

Unnecessary detail in the requirements for the PRA reduces flexibility, without any increase in clarity or predictability. The details in the PRA requirements also increase regulatory complexity and burden, since it will likely increase the amount of information from the PRA that must be included in the licensing basis, with no increase in safety.

(b) Specific uses of analyses. The PRA risk evaluation in combination with other generally accepted approaches for systematically evaluating engineered systems must be used (1) In informing the selection of the LBEs, as described in § 53.240, which must be considered in the design to determine compliance with the safety criteria in subpart B of this part.

(2) For informing the classification of SSCs according to their safety significance in accordance with § 53.460 and for identifying the environmental conditions under which the SSCs and operating staff must perform their safety functions.

(3) In evaluating the adequacy of defense-in-depth measures required in accordance with § 53.250.

(4) To identify and assess all plant operating states where there is the potential for the uncontrolled release of radioactive material to the environment.

Comments on NRCs Proposed Part 53 Page 41 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution The requirements for fire protection and aircraft impacts should be conditioned on the applicability of these types of events to the nuclear facility. Some designs, such as microreactors, should not need to consider aircraft impacts because the consequences do not pose an undue risk to the public. Similarly, for some plants a fire hazard may not pose an undue risk to the public.

The requirement 53.460(c) related to confidence that human actions will be performed as assumed in the analysis are out of place there and would be more appropriate here in 53.450.

Flexibility for bounding analyses, informed by DG-1414 should also be included here.

(5) To identify and assess events that challenge plant control and safety systems whose failure could lead to the uncontrolled release of radioactive material to the environment. These include internal events, such as human errors and equipment failures, and external events identified in accordance with subpart D of this part.

C-9 53.450(c)

See C-8 above for the change from PRA to risk evaluation.

Consistent with the part 50/52 lessons learned rulemaking, the upgrade requirement should be triggered when a new standard is endorsed and not based on an arbitrary deadline. Even this requirement may be inappropriate for some microreactors. Some microreactor designs may be designed for longer refueling cycles and may remain largely unchanged for decades. A 5-year update requirement is proscriptive and not performance-based or worth the cost for some reactor designs.

(c) Maintenance and upgrade of analyses. The PRA risk evaluation must be maintained at least every time the standard is updated 5 years or refueling outage, whichever is later until the permanent cessation of operations under § 53.1070 and upgraded in conformance with generally accepted methods, standards, and practices that have been endorsed or otherwise found acceptable by the NRC.

Comments on NRCs Proposed Part 53 Page 42 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution C-10 53.450(d)

This language could be understood to put unmeetable requirements on PRA software which are used in modeling plant behavior in analyses of licensing-basis events but cannot be qualified in the traditional sense of RG 1.203 compliance.

The PRA acceptability and qualification of codes such as CAFTA, SAPHIRE, FTREX, and PRAQUANT should be maintained in line with current industry practice.

(d) Qualification of analytical codes. The analytical codes used in modeling plant behavior in physics-based analyses of licensing-basis events (including but not limited to thermodynamics, reactor physics, fuel performance, and mechanistic source term codes) must be qualified for the range of conditions for which they are to be used.

C-11 53.460 Safety Categorization and Special Treatment.

While the NRC does not need to prescribe the specific safety categories that all nuclear facilities must use, the set of safety categories established in this requirement is reasonably flexible. However, modifications are needed to some of the details related to special treatment, which should not be required in the manner that 53.460 establishes.

Without defining the differences in the special treatments between these types of SSCs it is possible, or even likely, that NSRSS SSCs will receive an equivalent regulatory burden as SR SSCs. The application of the term special treatment without clear definitions of key terms to SR and NSRSS SSCs reduces regulatory predictability and is not necessary. This is exacerbated by requirements in 53.440 which overprescribe requirements for NSRSS SSCs as described in comments C-5 and sub-comments.

(1) 53.460(c) should be relocated to 53.450.

(2) Delete 53.450(b) as special treatments are already required per 53.440, and numerous requirements in Subpart F (53.865 for QA, which should be updated to include the guidance that was in 53.460 for Appendix B applicability only required for SR SSCs).

Making the changes proposed in the second option would require conforming changes throughout Part 53.

Comments on NRCs Proposed Part 53 Page 43 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution While the NRC appears to be applying the term special treatment to describe how Part 53 requirements apply to SR and NSRSS SSCs, such an approach was not necessary in Parts 50 and 52. The NRCs Part 53 definition of special treatment effectively says that it is requirements that apply to certain SSCs. Thus, the definition provides no clarity or regulatory stability. The approach in Parts 50 and 52 is to state within specific requirements whether they apply to safety-related or risk-significant SSCs. The same can be done in Part 53, where the requirements specify to which category of SSCs they apply, and in fact most Part 53 requirements already do this. Thus, the use of the term special treatment creates confusion and provides no regulatory benefit.

The requirement in 53.460(c) related to the confidence that human actions will be performed as assumed in the analysis is out of place and would be more appropriately included in 53.450.

C-12 53.470 Maintaining Analytical Safety Margins Used to Justify Operational Flexibilities.

This requirement is confusing in that it is not clear what the alternative criteria are supposed to be. Nor is it clear what benefit in operational flexibility would result from using the alternative criteria. The requirement is also essentially unused throughout the rest of Part 53.

Thus, the creation of this confusing and unnecessary requirement in 53.470 reduces regulatory clarity and predictability because applicants do not know what Delete 53.470.

If 53.470 is kept, it needs to be supported by regulatory guidance providing information on what potential benefit is gained by this optional requirement. One option would be clarity in the rule language that exemptions would not be needed if the staff determines

Comments on NRCs Proposed Part 53 Page 44 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution purpose it serves or operational benefits it offers. It also could be used inappropriately to force more strict criteria on designs to achieve the same operational flexibility that is provided in Parts 50 and 52 without an equivalent requirement in those parts.

that margins justify deviation from the regulatory requirements.

C-13 53.480 Earthquake Engineering 53.480 increases the scope of earthquake engineering to SR and NSRSS SSCs without providing sufficient clarification on the design requirements for NSRSS SSCs. This is particularly problematic combined with the requirement 53.440(b) which could imply both SR and NSRSS structures be designed to ASCE 43-05 (and eventually ASCE 43-19) as the only NRC endorsed standards available.

NEI suggests limiting the scope of 53.480 to SR SSCs. If the scope remains SR and NSRSS SSCs, the finalization of DG-1410 as RG 1.251 is essential to providing guidance for NSRSS SSCs. Part 53 is not economically implementable as written without NRC-endorsed guidance for NSRSS SSCs.

§ 53.480 Earthquake engineering.

(a) Effects of earthquakes. Structures, systems, and components classified as SR or NSRSS must be Publish DG-1410 as RG 1.251 to support implementation of the more flexible language in Part 53.

Consider publishing guidance, building on the graded approach to site characterization discussed in ML24213A337 to meet the requirements of 53.480 and Subpart D.

C-14 53.480(c) (vi) Soil Structure interaction The requirement in 53.480(c) (vi) on soil structure interaction is overly prescriptive and may not be appropriate for some sites and designs. This is recognized in NUREG-0800, 3.7.2: for sites where SSI effects are considered insignificant and fixed base analyses of structures are performed, bases and justification for not performing SSI analyses are reviewed by the NRC on a Revise 53.480(c)(vi) as follows:

(vi) The evaluation of SSCs required by this section to show they are able to function during and after earthquake ground motion should consider must take into account soil-structure interaction effects and the expected duration of vibratory motion. It is permissible

Comments on NRCs Proposed Part 53 Page 45 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution case-by-case basis. If the SSI analysis is not required, the input motion at the base of the structures will be the design motion reviewed in SRP Section 3.7.1" Since Part 53 is intended to be performance-based and minimize exemptions, the NRC should consider language such as should consider, if appliable soil-structure instead of must take into account Also consider changing strain limits in excess of yield strain to inelastic behavior to better characterize the properties of concrete.

to design for strain limits in excess of yield strain inelastic behavior in some of these SSCs during the DBGMs and under the postulated concurrent loads, provided the necessary safety functions are maintained.

Comments on NRCs Proposed Part 53 Page 46 of 123 Detailed Comments on Subpart D - Siting Requirements Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution D-1 Subpart D While we are willing to support the NRC approach to include siting requirements in Part 53, rather than continue to point to Part 100, the NRC preliminary rule language is essentially identical to the current requirements. Thus, the NRC does little more than relocate the siting requirements from one part to another and does not endeavor to establish a more modern technology-inclusive, risk-informed and performance-based approach to siting that is more appropriate for Part 53 We believe an incremental approach to Part 53 is a missed opportunity to achieve transformational changes that result in a more efficient regulatory framework to protect the public health and safety.

Part 53 should reevaluate the approach to siting by recognizing that it is largely the same as it was originally conceived in 1960s/1970s, and that Part 53 is being built upon more a more modern and flexible regulatory framework.

A more modern technology-inclusive, risk-informed and performance-based approach to siting was proposed by NEI in the comment letter submitted February 11, 2021, that better aligns with the Part 53 framework. The essence of that proposal was to integrate siting with safety, security and EP to The NRC should pursue a more modern technology-inclusive, risk-informed and performance-based approach to siting.

Such approach should consider how safety, security, EP and siting could be better integrated to establish a more efficient Part 53 framework.

Specifically, the NRC should consider the NEI proposed approach to modernize siting requirements, submitted in the February 11, 2021 comments, which would require that the characteristics of the site that have a significant impact on the ability of the facility to meet the public protection criteria in the NEI proposed language and the location of the site boundary be part of the facility characteristics. It would also require the establishment of design features and human actions to protect against manmade hazards related to the site such that the public protection criteria are met.

The NEI proposal would achieve a more efficient, less burdensome regulatory framework, that is enabled by better

Comments on NRCs Proposed Part 53 Page 47 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution achieve a more holistic and efficient regulatory framework.

integrating safety, security, EP and siting, resulting in an effectively higher level of safety than is currently achieved in Parts 50 and 52. This is accomplished by establishing the site boundary, in lieu of the low population zone and exclusion area boundary as the key boundary, in alignment with the SMR Emergency Preparedness rulemaking, streamlining the licensing basis of the facility. This would avoid the need to establish a requirement for distance from a population center, since it is not necessary, nor does it include specific requirements for a population density distance or the seismic and geologic criteria, since it is not needed for a technology-inclusive, performance-based and risk-informed approach.

D-2 53.500 General Siting and siting assessment It is unclear what the NRC is attempting to accomplish with this requirement, since the requirement essentially duplicates all of the requirements in Subpart D at a high level. If the NRCs intent is to establish the purpose of Subpart D, then the requirement should be written to state that this is a purpose statement, which clarifies the collective nature of Subpart D requirements and the outcome achieved by meeting the collective Revise the requirement as follows:

The purpose of Subpart D is to ensure that (a) The siting of each commercial nuclear plant must be is supported by assessments of proposed sites such that the design, including design features and programmatic controls corresponding to the site characteristics, satisfies the safety

Comments on NRCs Proposed Part 53 Page 48 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution requirements, and that this is not a requirement that an applicant and licensee must explicitly meet.

Establishing a high-level requirement that duplicates all the other requirements in Subpart D, without stating that this requirement does not need to be met but is met implicitly by meeting the other requirements in Subpart D reduces regulatory clarity and predictability and increases regulatory burden without an increase in safety.

criteria defined in §§ 53.210 and 53.220 or more restrictive alternative criteria adopted under § 53.470. The siting assessment must ensure that site characteristics that might contribute to the initiation, progression, or consequences of licensing-basis events (LBEs) analyzed under §§ 53.450 and 53.480 are identified and mitigated by design features or programmatic controls.

The siting assessment must take into consideration the potential adverse impacts that a commercial nuclear plant may have on nearby populations as a result of normal operations or LBEs.

(b) Activities performed to identify site characteristics or otherwise needed to determine site-specific contributors to functional design criteria or analysis assumptions under subpart C of this part must satisfy the applicable special treatment requirements of § 53.460, including, where applicable, the quality assurance requirements from appendix B of part 50 of this chapter.

This is accomplished by meeting the set of requirements in Subpart D

Comments on NRCs Proposed Part 53 Page 49 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution D-3 53.510 External Hazards 53.520 Site Characteristics 53.530 Population-related considerations 53.540 Siting interfaces 53.550 Environmental Considerations Per comment D-1, these requirements should be completely reconsidered and established in a more modern technology-inclusive, risk-informed and performance-based approach to siting. Such an approach should consider how safety, security, EP and siting could be better integrated to establish a more efficient Part 53 framework.

See proposed resolution to comment D-1.

Additional specific comments are not provided as Subpart D should be significantly revised.

D-4 53.530 Population-related considerations Reference to 53.020 Definitions The population center distance requirement in Part 53 copies the requirement from Part 100 which may not be appropriate for smaller, low-consequence reactors. NRC seems to acknowledge this requirement may be too restrictive in SECY-24-0008:

Some micro-reactor license applicants may seek to site reactors at locations that would not conform to the current Commission policy and the regulations in 10 CFR 100.21(b). Such deployment scenarios are being considered for several reasons including replacing existing coal plants or providing process heat for heating or industrial applications, or to provide power to remote communities or smaller Ideally remove 53.530(c) as this information can be captured in guidance.

If retained, revise 53.530 follows:

Every site must have an exclusion area, a low-population zone, and a population center distance as defined in §53.020. For sites that establish that no plume exposure pathway emergency planning zone (EPZ) is required or that the plume exposure pathway EPZ does not extend beyond site boundary in accordance with the requirements of § 53.1109(g)(2) of this chapter, the low-population zone and

Comments on NRCs Proposed Part 53 Page 50 of 123 grids with relatively small but concentrated populations that would be close to a reactor site.

In that SECY, NRC identifies as a next step, The NRC staff will inform the Commission if it becomes aware of any license applicants who intend to seek exemption from 10 CFR 100.21(b) and will raise associated policy issues to the Commission accordingly.

Since Part 53 intends to be performance-based and should consider microreactor deployment models as directed by the ADVANCE Act, the population center distance criteria should not be prescriptive and exclusionary.

population center distance would coincide with the site boundary.

(c) Reactor sites should be located away from very densely populated centers. Areas of low-population density are, generally, preferred for reactor sites. However, in determining the acceptability of a particular site located away from a very densely populated center but not in an area of low-population density, consideration will be given to safety, environmental, economic, or other factors, which may result in the site being found acceptable even if it is not located in an area of low-population density, including a site that is located close to or within a densely populated center containing more than about 25,000 residents.

Performance-based guidance on potentially colocating the low population zone and population center distance with a site boundary EPZ should be developed for the population density criterion to potentially allow siting of low-consequence reactors close to or within densely populated centers containing more than about 25,000 residents. NEI intends to further justify this position in an upcoming White Paper to the NRC.

Comments on NRCs Proposed Part 53 Page 51 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution D-5 Subpart D - ADVANCE Act One of the concepts raised in the Rapid High Volume Deployment of microreactor (RHDRA) report (ML24213A337) is a pathway to a general license for microreactors. Such a pathway should be captured under Subpart D of Part 53, but will require significant engagement between NRC and industry.

Add a pathway to a general license for microreactors and simplified siting requirements to the extent allowable by statute if bounding consequence requirements or defined site parameters are met.

Comments on NRCs Proposed Part 53 Page 52 of 123 Detailed Comments on Subpart E - Construction and Manufacturing Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution E-1 53.605 Reporting of defects and noncompliance It is not clear why this requirement has been included.

It simply duplicates Part 21. This duplication is essentially noted in multiple paragraphs and subparagraphs in 53.605.

NRC and industry considered removing 50.55(e) in 2012 (ML13163A401) as unnecessary given Part 21.

This argument remains valid, and consistent with the NRC position at the time The underlying purpose of 10 CFR 50.55(e) [and therefore 53.605] is fully achieved through the implementation of 10 CFR Part 21 and other regulatory processes. Therefore, 10 CFR 50.55(e) should be deleted and can be deleted without any reduction to the health and safety of the public.

Delete 53.605 and, if warranted, simply make reference to the applicability of Part 21 in 53.600.

E-2 53.610 Construction The requirements addressing Construction, while generally clear, do have some internal overlap and in some cases duplicate (or nearly so) QA requirements in Appendix B. For example, 53.610(a)(1-3) essentially duplicates requirements in Appendix B. 53.610(a)(6),

which requires a QA program meeting the requirements of Appendix B. Consequently, a reference to the Appendix B requirements in lieu of specific requirements in 53.610(a) would be clearer and avoid inconsistent interpretation of these requirements.

53.610 should be revised to eliminate duplication and overlap of requirements, such as 53.610(a)(1-3), 53.610(a)(6).

53.610(a) should be revised to group all of the required programs under a single paragraph, with subparagraphs for each program, and separating this from the other requirements on organization and procedures in 53.610(a).

Comments on NRCs Proposed Part 53 Page 53 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution A separate paragraph identifying the programs relevant to Construction, with subparagraphs listing the specifically required programs. This is nearly done in 53.610(a)(1) but the listing of programs is mixed with other requirements on organization and various procedures. A reorganization of 53.610(a) would improve clarity.

53.610(b)(1)(i)(C) requires a plant staff training program associated with the receipt of radioactive material. This requirement should be relocated to Subpart F and included with other training programs.

53.610(b)(1)(i)(C) should be relocated to Subpart F.

E-2a 53.610(b)

Construction Activities The language under this Part would preclude a Part 50 construction permit (CP) holder or a Part 52 Early Site Permit holder from transitioning to a Part 53 Operating License or Combined License. This restriction is unnecessary and precludes legitimate licensing pathways with no technical basis, particularly for Part 50 and 52 permit holders following LMP.

Change 53.610(b) to begin:

No person may begin the construction of a commercial nuclear plant on a site on which the facility is to be operated under this part until that person has been issued either a CP or COL, an early site permit authorizing activities under § 53.1130, or an LWA under this part.

E-3 53.620 Manufacturing As with 53.610, the requirements addressing Manufacturing in 53.620, while generally clear, do have some internal overlap and in some cases duplicate (or nearly so) QA requirements in Appendix B. For example, 53.620(a)(1-3) essentially duplicate 53.620 should be revised to eliminate duplication and overlap of requirements, such 53.620(a)(1-3) and 53.620(a)(6).

Comments on NRCs Proposed Part 53 Page 54 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution requirements in Appendix B. 53.620(a)(6) requires a QA program meeting the requirements of Appendix B.

Consequently, a reference to the Appendix B requirements in lieu of specific requirements in 53.620(a) would be clearer and avoid inconsistent interpretation of these requirements.

A separate paragraph identifying the programs relevant to Construction, with subparagraphs listing the specifically required programs. This is nearly done in 53.620(a)(1) but the listing of programs is mixed with other requirements on organization and various procedures. A reorganization of 53.620(a) would improve clarity.

There are additional requirements for programs in 53.620(c). A fire protection program is required under 53.620(c)(2), and an emergency plan is required under 53.620(c)(3). Similarly, there are programs required under 53.6220(d) fuel handling. 53.620(d)(2)(iv) requires a physical security program for the storage of fresh fuel in accordance with 73.67 and 53.620(d)(v) requires an MC&A program in accordance with Part 74.

53.620(c)(4) requires a plant staff training program associated with the receipt of radioactive material. This requirement should be relocated to Subpart F and included with other training programs.

53.620(a) should be revised to group all of the required programs under a single paragraph, with subparagraphs for each program, and separating this from the other requirements on organization and procedures in 53.620(a), and consideration should be given to grouping all programs required under 53.620 in a single paragraph.

53.610(c)(4) should be relocated to Subpart F.

Comments on NRCs Proposed Part 53 Page 55 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution There are tradeoffs in having the requirements for programs associated with key paragraphs versus grouping all of the program requirements with descriptions of when they are applicable.

The overarching issue is clarity in the requirements and whichever approach is deemed to provide the most clarity is the approach that should be used in Part 53.

E-4 53.620 Manufacturing 53.620(d)(1)(i) allows fuel loading of a manufactured reactor only if at least two independent physical mechanisms are in place, each of which is sufficient to prevent criticality assuming optimum moderation and neutron reflection conditions. The requirement also assumes the ML owns a license to possess special nuclear material pursuant to 10 CFR Part 70.

The requirements of Subpart H of 10 CFR Part 70 already provide reasonable assurance that subcriticality will be maintained throughout all processes. For example, 10 CFR 70.61(d) states that "the risk of nuclear criticality accidents must be limited by assuring that under normal and credible abnormal conditions, all nuclear processes are subcritical, including use of an approved margin of subcriticality for safety. Therefore, the additional requirement of two independent physical mechanisms to maintain subcriticality is unnecessary and burdensome.

53.620(d)(1)(i) should be revised to remove...only if the manufactured reactor is configured during its loading, storage, and transport with at least two independent physical mechanisms in place, each of which is sufficient to prevent criticality assuming optimum neutron moderation and neutron reflection conditions.

53.620(d)(1)(iii) should be revised to: The Commission has determined that any such fueled manufactured reactor possessed under a 10 CFR part 70 license is not in operation.

52.620(d)(1)(iv) should be revised to:

Upon installation of the fueled manufactured reactor in its place of operation and a Commission finding that

Comments on NRCs Proposed Part 53 Page 56 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution Subsequent requirements 53.620(d)(1)(iii & iv) also contain references to the independent physical mechanisms described above.

the acceptance criteria in the COL that authorized reactor construction are met under § 53.1452(g), the licensee may remove manufacturing controls in place to ensure subcriticality. Upon removing those controls, a manufactured reactor has commenced operation.

Footnote 1 in 53.1452 should be revised to state...of its scheduled date for initiating the removal of manufacturing subcriticality controls required under 53.620(d)(1) through the ML no later than...

50.160(c)(2) should be updated to Removal of manufacturing subcriticality controls at the operating site instead of removal of any one of the independent physical mechanisms to prevent criticality All other instances in the proposed rule of physical mechanisms should be rephrased to manufacturing subcriticality controls.

E-5 53.620(e)(1) 53.620(e)(1) requires that the recipient of a reactor manufactured under this section have a COL issued The NRC should revise the rule in

§ 53.620(e)(1) to allow transport to

Comments on NRCs Proposed Part 53 Page 57 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution under Part 52. This regulation, if enacted would exclude international transport for a microreactor manufactured in the U.S. under a Part 53 ML.

Microreactor paradigms include manufacture in the U.S. and transportation to international deployment sites. The Atomic Energy Act authorizes export of utilization facilities, therefore so should Part 53 for microreactors manufactured under an ML, therefore the NRC rule should not exclude it or require exemption to facilitate export.

international destinations, with the requirements that the holder of an ML obtains an NRC export license under 10 CFR part 110, otherwise the ML holder would have to request an exemption to export each reactor.

E-6 53.620(e)(2) & (4)

The requirement in § 53.620(e)(2) appears to be duplicative with § 53.620(e)(4). The regulations in

§ 53.620(e)(2) state that any contract to transport the reactor must contain language requiring the person transporting the manufactured reactor comply with all shipping requirements in applicable NRC regulations, certificates of compliance, and NRC-issued licenses.

The requirements in § 53.620(e)(4) require compliance with 10 CFR Part 71. By including certificates of compliance in § 53.620(e)(2), this part of the proposed rule appears to indicate that the manufactured device contains fuel, and therefore, the shipper (the licensee possessing the material) must include in a contract that the carrier must comply with all shipping requirements in applicable NRC regulations, certificates of compliance, and NRC-issued licenses. The licensee is responsible for ensuring that the shipment, regardless Remove 53.620(e)(2) as duplicative.

Comments on NRCs Proposed Part 53 Page 58 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution of which carrier transports the radioactive material, complies with all NRC requirements in 10 CFR Part71, including 10 CFR 71.5, which requires compliance with the requirements of the Department of Transportation, which include package, marking and labeling, placarding, and road, rail, air, and vessel modal requirements.

Detailed Comments on Subpart F - Requirements for Operation Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution F-1 53.700 Operational Objectives Are these operational objectives for all advanced reactors (either self-reliant-mitigation or interaction-dependent-mitigation facilities)? The reason this is asked is because 53.800 is specifically labeled self-reliant-mitigation facilities.

Because 53.800 references that subsections of 53.800 are used in lieu of specific subsections of 53.700, recommend a similar statement be added here. All sections of 53.700 apply unless otherwise noted in 53.800 F-2 53.700 (a)(1)

Each holder of an OL or COL under this part must maintain the capabilities, availability, and reliability of plant SSCs to ensure that the safety functions identified in § 53.230 will be performed if called upon during licensing-basis events (LBEs).

Recommend that more clarity be added to how much defense in depth is required or cite a reference that is based in regulation to ensure a consistent interpretation.

Comments on NRCs Proposed Part 53 Page 59 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution 53.220, part (a) states (a) Plant SSCs, personnel, and programs provide the necessary capabilities and maintain the necessary reliability to address LBEs other than DBAs in accordance with §§ 53.240 and 53.450(e),

and provide measures for defense in depth in accordance with § 53.250 (emphasis added); and Defense in depth, used in this context, is not specifically defined, so can leave a wide variation between end users as to how in depth, or not, this approach is applied - from both the plant designers and the future NRC regulators.

An update to RG 1.233 endorsing its use under Part 53 provides one avenue for providing clarity, but other regulatory guidance addressing DID should also provide an acceptable path to addressing this requirement.

F-3 53.710 The term controls is used throughout this set of requirements. Controls is later defined in 53.725(c) as apparatus and mechanisms, the manipulation of which directly affects the reactivity or power level of the reactor. The use of controls within 53.710 is not consistent with the defined term.

Replace controls with measures.

Review Part 53 for other uses of controls to ensure consistent use.

F-4 53.720 Response to seismic events.

The section has a statement that operating basis earthquake Ground Motion or significant plant damage due to vibratory ground motion occurs The second requirement in this sentence needs definition or clarification, i.e.,

significant plant damage is arbitrary, therefore, should be removed or defined.

LCOs are already defined for SR SSCs in 53.710, so the intent of significant plant damage is likely duplicative.

Comments on NRCs Proposed Part 53 Page 60 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution F-5 53.730 Many of the elements in this section are elements of a "state-of-the-art human factors engineering" program (refer to NUREG-0711). If specific elements are required to meet regulation, the industry will not be able to maintain the state-of-the-art without future rulemaking.

This part should mirror the Part 50 approach which maintains a requirement for a Human Factors Engineering program, but does not prescriptively require in regulation certain elements. This provides flexibility to the end-users to keep up with the "state-of-the-art" without future rulemaking.

F-6 53.730 (f)(1) 53.730 (f)(1) states... staffing plan must include a description of how engineering expertise will be available to the on-shift operating personnel during all plant conditions...

Guidance should provide clarity on the meaning of be available to the on-shift personnel. Is this a 24/7 position or just available to support when needed?

This requirement should be further described and bounded. If the design supports that this position is not needed, provide this option in the rule without having to go through the exemption process. Since the NRC agreed with NuScales approach for eliminating the STA position the regulation should include a pathway for not requiring engineering expertise to be available to the crew.

Recommend that there is a clarifying statement added that allows/requires flexibility for what time frame engineering expertise is required, and that time frame is based on analyzed accident progression timelines specific to the design. Provide guidance to support meeting the requirement.

Comments on NRCs Proposed Part 53 Page 61 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution F-7 53.730(f)(4)

The section states in part... Applicants for or holders of OLs or COLs under this part must include within their staffing plans a description of how the numbers, positions, and responsibilities of personnel contained within those plans will adequately support all necessary functions within areas such as plant operations, equipment surveillance and maintenance, radiological protection, chemistry control, fire brigades, engineering, security, and emergency response.

Having a requirement for staffing plans to include specific minimum numbers for how many persons are required to perform equipment surveillance and maintenance seems to be excessive, and over-burdensome. Taken too literally, this could include a requirement to pre-define how many maintenance personnel are required to perform refueling outage maintenance. Is it the intent of the NRC to mandate minimum maintenance contingent.

Recommend that numbers be removed from part 53 and changed to read...how the positions and responsibilities.

F-8 53.730 (g) 53.730 (g) states: Training, examination, and proficiency programs. Develop, implement, and maintain programs that comply with the following requirements. These programs must be approved by the NRC as part of its approval of the OL or COL for the plant.

Provide clarification if facility licensee training programs accredited by the National Nuclear Accrediting Board (NNAB) would be considered a Commission-approved training program. This could be provided in the applicable definition section defining

Comments on NRCs Proposed Part 53 Page 62 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution NUREG 1021, Operator Licensing Examination Standards for Power Reactors, establishes the policies, procedures, and practices for examining licensee and applicants for reactor operator and senior reactor operator licenses at the nuclear power reactor facilities under Title 10 of the Code of Federal Regulations (10 CFR) Part 55, Operators licenses.

2.2 Applications, Medical Requirements, and Waiver and Excusal of Examination and Test Requirements of NUREG 1021, the commission recognized facility licensee training programs accredited by the National Nuclear Accrediting Board (NNAB) would be considered a Commission-approved training program. Specifically, As stated in the Statement of Considerations for the 1987 final rule amending 10 CFR 55 (Volume 52 of the Federal Register (FR), page 9456);

March 1987), subject to continued Commission endorsement of the industrys accreditation process under the Final Policy Statement on Training and Qualification of Nuclear Power Plant Personnel (50 FR 11147; March 20, m1985), facility licensees training program would be considered a Commission-approved training program if it is accredited by the National Nu8clear Accrediting Board (NNAB)..

DRO-ISG-2023-01, Operator Licensing Programs Draft Interim Staff Guidance 10 CFR 53, October 2024, provides interim staff guidance to facilitate staff commission approved training programs and including similar wording contained in NUREG 1021, Operator Licensing Examination Standards for Power Reactors. Provide clarification guidance in DRO-ISG-2023-01, Operator Licensing Programs Draft Interim Staff Guidance, 10 CFR 53, October 2024. However, this guidance is applicable to operator licenses to include reactor operator, senior reactor operator, and general licensed operator and would not include the requirements of 53.830 for general personnel training requirements. (See 53.780 for similar comment)

Comments on NRCs Proposed Part 53 Page 63 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution reviews of applications for an operating license or combined license under Title 10 of the Code of Federal Regulations (10 CFR) Part 53. The interim guidance does not provide similar guidance in acknowledging operator licensing or general licensing training programs accredited by the National Nuclear Accrediting Board (NNAB) would be considered a Commission-approved training program.

F-9 53.735 (a) & (b)

General Exemptions 53.735 states: The regulations in 53.725 through 53.830 do not require a license for an individual who -

(a) Under the direction and in the presence of an operator, senior operator or generally licensed reactor operator, as appropriate as a part of the individuals training in a facility licensees training program (b) Under the direction and in the presence of a senior operator or generally licensed reactor operator, as appropriate to load or unload the fuel into, out of, or within the reactor vessel while the reactor is not operating For part (b) moving fuel it is unclear why oversight of a reactor operator is insufficient. Why would the interaction-dependent-mitigation facility require SRO.

Recommend that part (b) uses the same language as part (a).

Under the direction and in the presence of an operator, senior operator or generally licensed reactor operator, as appropriate

Comments on NRCs Proposed Part 53 Page 64 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution F-10 53.775 Applications for operators and senior operators.

Form NRC 398, which appears in multiple sections.

Should be corrected to NRC Form 398.

F-11 53.775 (a) (i)

A draft NRC Form 398 was not updated as part of the Part 53 Rule Package.

Update NRC Form 398 to reflect Part 53 allowances (e.g., Commission approved training programs).

F-12 53.775(iii) 53.775(iii) includes the language: Provide evidence that the applicant, as a trainee, has successfully demonstrated competence in manipulating the controls of either the facility for which a license is sought or a simulation facility that demonstrates compliance with the requirements of §53.780(e). For operators applying for a senior operator license, certification that the operator has successfully operated the controls of the facility as an operator will be accepted; and...

In contrast to the existing requirements of 10 CFR 55.31, the Part 53 language does not specify the minimum number of control manipulations to be performed. Can the NRC clarify if it intends to also revise NRC Form 398 to remove the requirement to perform at least 5 significant control manipulations?

The number of required control manipulations needed to be performed should be determined through the SAT-based methodology and based on the design and Recommend keeping the language as-is.

NRC clarify whether they plan to revise NRC Form 398. Recommend removing the 5-control manipulation requirement from NRC Form 398.

Comments on NRCs Proposed Part 53 Page 65 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution operating characteristics of the reference plant. For designs that can demonstrate reduced reliance on humans to achieve safety, operators would only need to demonstrate successful completion of 1 or 2 control manipulations. For designs that have increased reliance on humans to achieve safety, operators would need to demonstrate more.

F-13 53.780 (a)(1) and (c)(1)(ii)

Training, examination, and proficiency program.

53.780 (a) (1) states: The program must be approved by the commission prior to its use for training applicants, as described under 53.730 (g).

53.780 (c) (1) (ii) states for operator requalification program The program must be approved by the Commission prior to its use for continuing training as described under 53.730 (g). NUREG 1021, Operator Licensing Examination Standards for Power Reactors, establishes the policies, procedures, and practices for examining licensee and applicants for reactor operator and senior reactor operator licenses at the nuclear power reactor facilities under Title 10 of the Code of Federal Regulations (10 CFR) Part 55, Operators licenses. 2.2 Applications, Medical Requirements, and Waiver and Excusal of Examination and Test Requirements of NUREG 1021, the commission recognized facility licensee training programs accredited by the National Nuclear Accrediting Board (NNAB) would be considered a Commission-approved training program. Specifically, As stated in the Provide clarification if facility licensee training programs accredited by the National Nuclear Accrediting Board (NNAB) would be considered a Commission-approved training program. This could be provided in the applicable definition section defining commission approved training programs and including similar wording contained in NUREG 1021, Operator Licensing Examination Standards for Power Reactors.

Provide clarification guidance in DRO-ISG-2023-01, Operator Licensing Programs Draft Interim Staff Guidance 10 CFR 53, October 2024. However, this guidance is applicable to operator licenses to include reactor operator, senior reactor operator, and general licensed operator and would

Comments on NRCs Proposed Part 53 Page 66 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution Statement of Considerations for the 1987 final rule amending 10 CFR 55 (Volume 52 of the Federal Register (FR), page 9456); March 1987), subject to continued Commission endorsement of the industrys accreditation process under the Final Policy Statement on Training and Qualification of Nuclear Power Plant Personnel (50 FR 11147, March 20, 1985), a facility licensees training program would be considered a Commission-approved training program if it is accredited by the National Nuclear Accrediting Board (NNAB).. DRO-ISG-2023-01, Operator Licensing Programs Draft Interim Staff Guidance 10 CFR 53, October 2024, provides interim staff guidance to facilitate staff reviews of applications for an operating license or combined license under 10 CFR Part 53. The interim guidance does not provide similar guidance in acknowledging operator licensing or general licensing training programs accredited by the National Nuclear Accrediting Board (NNAB) would be considered a Commission-approved training program.

not include the requirements of 53.830 for general personnel training requirements.

F-14 53.780(b)(2) 53.780(b)(2) states: The facility licensee must submit prepared examinations to the Commission for review and approval in advance of their administration The regulations of 53.780(b)(2) do not clearly define the timelines in which facility licensees are expected to Recommend clarification to delete "in advance of" and replace with "prior to."

Provide a time frame of how far prior to in guidance documents.

Comments on NRCs Proposed Part 53 Page 67 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution meet regarding certain aspects of the operator licensing initial examination program.

F-15 53.780(b)(3)

The section contains the following... The facility licensee must ensure that sufficient advance notification is provided to the Commission to either administer the examination or allow for a representative of the Commission to be afforded the opportunity to be present when the facility licensee administers the examination.

The regulations of 53.780(b)(2)-(4) do not clearly define the timelines in which facility licensees are expected to meet regarding certain aspects of the operator licensing initial examination program.

The word sufficient seems to be vague and should be defined in guidance documents.

Recommended rephrase "The facility licensee must ensure that the Commission is notified of the expected date the examination is to be administered to provide the Commission with the opportunity to be present when the facility licensee administers the examination."

F-16 53.780(b)(4) 53.780(b)(4) states: (4) Graded examination documentation for each applicant must be promptly provided to the Commission for review in making operator licensing decisions."

It is unclear what is prompt.

Recommend the following (4) Delete the word "promptly." NRC can provide expectation as to "how early" in guidance documents.

F-17 53.780(b)(5) 53.780(b)(5) states: The facility licensee must maintain operator licensing initial examination program records documenting the participation of each operator and senior operator applicant in the initial examination. The records must contain copies of Delete the requirement to maintain the answers given by the applicant or clarify the basis behind maintaining the answers

Comments on NRCs Proposed Part 53 Page 68 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution examinations administered, the answers given by the applicant, documentation of the grading of examinations, and documentation of any additional training administered in areas in which an applicant exhibited deficiencies. The facility licensee must retain these records during the period in which the associated operators or senior operators remain licensed at the facility.

Operating experience from the U.S. Navy - exam answers are not maintained as part of the qualification record, only the exam coversheet which documents the results (pass/fail grade), date of the exam, the grader of the exam, and a signature from the candidate confirming they have reviewed the results. Maintaining the answers given by the candidate increases the number of records that need to be maintained with no added value.

given by the applicant on the initial examination.

F-18 53.780(c)(2)(ii)(B) 53.780(c)(2)(ii)(B) states: The facility licensee must ensure that a representative of the Commission is afforded the opportunity to be present during requalification examination administration.

Similar to the requirements of 53.780(b)(2)-(4), the NRC must define more clearly what the expectation is regarding being "afforded the opportunity to be present."

Delete requirement or update language to be clearer on what is expected (i.e.,

notification of expected administer date)

Recommend this be in a guidance document.

Comments on NRCs Proposed Part 53 Page 69 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution F-19 53.780(c)(2)(ii)(C) 53.780(c)(2)(ii)(C) states: The facility licensee must ensure that each operator and senior operator is administered a complete requalification examination on a periodicity not to exceed 24 months. Additionally, the facility licensee must ensure that any licensed operator or senior licensed operator who either demonstrates unsatisfactory performance on, or fails to complete, the biennial requalification examination is removed from the performance of licensed operator and senior licensed operator duties until such time that any necessary remedial training has been completed and a retake examination has been passed.

Is there a difference between the "complete requalification examination" and the "biennial requalification examination"? If not, recommend using more consistent language.

Clarify if the NRC intends to require a complete examination "retake" if a candidate fails the requalification exam or are there cases where the candidate may only have to retake certain areas if failures are only in particular areas.

Recommend rephrasing the last sentence.

The facility licensee must ensure that each operator and senior operator is administered a biennial requalification examination on a periodicity not to exceed 24 months. Additionally, the facility licensee must ensure that any licensed operator or senior licensed operator who either demonstrates unsatisfactory performance on, or fails to complete, the biennial requalification examination is removed from the performance of licensed operator and senior licensed operator duties until such time that any necessary remedial training has been completed and the candidate has successfully passed the necessary reexamination.

F-20 53.780(c)(2)(ii)(D) 53.780(c)(2)(ii)(D) states: The facility licensee must promptly provide a summary of examination results NRC to clarify or delete the term "promptly." NRC to also clarify the entity

Comments on NRCs Proposed Part 53 Page 70 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution for each operator and senior operator following the completion of the requalification examination.

The definition of prompt is unclear the summary of examination results is being provided to.

Recommended language: The facility licensee must provide the Commission with a summary of the examination results for each operator or senior operator following the completion of the requalification examination.

F-21 53.780(c)(3) 53.780(c)(3) states: The facility licensee must maintain operator licensing requalification program records documenting the participation of each operator and senior operator in the requalification program. The records must contain copies of examinations administered, the answers given by the operator or senior operator, documentation of the grading of examinations, and documentation of any additional training administered in areas in which an operator or senior operator exhibited deficiencies. The facility licensee must retain these records until the operator's or senior operator's license is renewed.

Operating experience from the U.S. Navy - exam answers are not maintained as part of the qualification record, only the exam coversheet which documents the results (pass/fail grade), date of the exam, the grader of the exam, and a signature from the candidate confirming they have reviewed the results. Maintaining Delete the requirement to maintain the answers given by the applicant or clarify the basis behind maintaining the answers given by the applicant on the requalification examination.

Comments on NRCs Proposed Part 53 Page 71 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution the answers given by the candidate increases the number of records that need to be maintained with no value added.

F-22 53.780(e)(iii) 53.780(e)(iii) states: "Promptly correct modeling and hardware discrepancies and discrepancies identified from scenario validation and from performance testing or provide justification as to why the presence of such discrepancies will not adversely affect simulator performance with respect to the criteria of paragraph (e)(2) of this section."

This requirement is similar to 10 CFR Part 55.46(d)(2).

However, the Commission includes the term "promptly."

Recommend deleting the word "promptly." The fix timeline may be impacted by a variety of factors such as supply chain, part availability. The timeline of the repair is the responsibility of the simulation facility owner and not the NRC.

F-23 53.785 (b) Conditions of operator and senior operator licenses 53.785 (b) contains the language: The license is limited to the facility for which it is issued...

The definition in 53.020 would indicate that the facility is one or more reactors in the same location and does not allow for being licensed to an identical design at another location. For microreactors that could be located at various locations, and potentially be operated remotely from a central location, the limitation of 53.020 definition will not make part 53 desirable for micro reactors, or other concept of Provide a means for operation of multiple remote facilities, from a central control facility and licensing to a specific reactor design in addition to a location as currently proposed in the rule.

Comments on NRCs Proposed Part 53 Page 72 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution operations that involve remote centralized control facilities.

F-24 53.785(c) 53.785(c) states: The license is limited to those controls of the facility or facilities specified in the license.

The language adds in the phrase "or facilities" (compared to 10 CFR 55.53).

Can the NRC clarify why it would not do the same with 53.785(b)? Or delete the phrase because the definition of term facility already includes words to the same effect.

Delete the phrase "or facilities" in (c) or add the phrase "or facility" to 53.785(b) -

The license is limited to the facility, or facilities, for which it is issued.

F-25 53.800 (a) Facility licensees for self-reliant-mitigation facilities 53.800 (a) states: A commercial nuclear plant is a self-reliant-mitigation facility if the NRC determined as part of its approval of the OL or COL for that plant that its design demonstrates compliance with criteria (a)(1) though (a)(5) of this section...

(a)(4) is the only paragraph that talks about how compliance may be achieved. That should be moved to a guidance document.

In general, there appears to be a lot of give and take to show all paragraphs will be met.

Guidance documents should ensure that both staff and developers can clearly agree that criteria are met. Defense in depth posing the biggest challenge (next comment)

Comments on NRCs Proposed Part 53 Page 73 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution F-26 53.800(a)(5) 53.800(a)(5) states: The plant design must provide for a layered defense-in-depth approach that is not dependent upon any single barrier or credited human action The criteria for self-reliant mitigation facilities need to be better defined. Its hard to say whether a design would meet the criteria, especially when it gets to DID actions. This requirement is unbounded.

As such, a design that required supplemental water for cooling after 100 days would not meet the definition due to the need for human action at that point. A time (i.e., less than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) should be included in guidance, or few plants will meet an unbounded criteria. The basis would be that the plant is stable and in a safe shutdown condition, however some DID after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> may be needed to ensure it remains that way. It is assumed that plant operators or the ERO response are fully engaged after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to provide additional oversight, technical support and staffing.

F-27 53.805(a)(1)(i)-(vii)

Facility licensee requirements related to generally licensed reactor operators.

53.805(a)(1) begins: (1) Ensure that, in addition to being qualified to perform those items identified by the facility-specific systems approach to training conducted under 53.815, generally licensed operators are qualified to safely and competently...

Delete requirement.

Comments on NRCs Proposed Part 53 Page 74 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution This is a prescriptive requirement. Generally licensed operators may be assigned roles and responsibilities different than the ones listed under (a)(1) depending on the facility design and the resulting concept of operations.

GLRO training programs are already required to be developed using a systems approach to training in accordance with 53.815 and 53.805(3).

F-28 53.805(a)(2) 53.805(a)(2) states: Develop, implement, and maintain facility technical specifications that provide the necessary administrative controls to ensure the implementation of these requirements.

Clarify what the phrase "these requirements" means?

Is it referring to just (a)(1) or all of 53.805?

Update to more specific implementation of 53.800 requirements or words to that effect.

F-29 53.805(a)(5) 53.805(a)(5) states: Report annually to the NRC the identity of all generally licensed reactor operators at the commercial plant, including all additions and deletions since the previous report.

Clarify what the purpose, objective, and intended use of the report is. Additionally, the report may not be reflective of the status of GLROs at other points in time between reporting periods.

The requirement should be deleted or rephrased to be more flexible and less burdensome (both for NRC and licensees).

One option is to have this information available for inspection as opposed to an annual report.

Comments on NRCs Proposed Part 53 Page 75 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution F-30 53.805(a)(6) 53.805(a)(6) states: Ensure the facility design continues to meet the criteria of 53.800 This is ensured by 53.1550. Having the requirement here is duplicative.

Delete requirement.

F-31 53.810 (general)

Generally licensed reactor operators.

Request for clarification: Its unclear when a "general license" is issued. Is it issued on the date of initial qualification or when they are employed into a position? Define how this is handled.

Provide clarification of when a general license is "issued."

F-32 53.810 (c) 53.810 (c) states: The general license is limited to the facility or facilities at which the operator is employed.

This is a benefit if facilities at which the operator is employed are defined to mean the same design of SMRs or Micro reactors that could be located at different locations and operated remotely.

Provide clarity on the intent of or facilities.

Arguably, facility is already defined to include "facilities" meaning multiple reactors under one operating license for a plant.

Or Rephrase to "is qualified." Language is not inclusive of certain fleet-based operating concepts as noted above.

Delete at which the operator is employed.

Comments on NRCs Proposed Part 53 Page 76 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution Reword: "The general license is limited to the facility for which it was issued" Or Rephrase: The general license is limited to the facility or facilities of which the operator is qualified.

F-33 53.810 (d) 53.810 (d) states: The Commission will suspend the general license on an individual basis for violations Does this mean suspending the general license for the individual facility or the general license for an individual at that facility?

The wording in (e) is very clear that it is talking about the GL with respect to the individual operator.

Rephrase: The Commission will suspend the general license of a generally licensed reactor operator on an individual basis for violations F-34 53.815 (b)(3)(vi)

Generally licensed reactor operator training, examination, and proficiency programs.

53.815 (b)(3)(vi) states: The training program must be approved by the Commission prior to its use examination program must provide for valid and reliable examinations and must be approved by the Commission prior to their use NEI appreciates the improvement that the initial examination is different and positive in that the NRC approves the process not the exam before the fact.

Keep requirement as-is.

Comments on NRCs Proposed Part 53 Page 77 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution This should be less of a burden than the traditional license route.

F-35 53.815(c) 53.815(c) states: The facility licensee must maintain records documenting the participation of each generally licensed reactor operator in the training and examination programs. The records must contain copies of examinations administered, the answers given by the generally licensed reactor operator, documentation of the grading of examinations, and documentation of any additional training administered in areas in which a generally licensed reactor operator exhibited deficiencies. The facility licensee must retain these records while the associated generally licensed reactor operators remain employed at the facility.

Operating experience from the U.S. Navy - exam answers are not maintained as part of the qualification record, only the exam coversheet which documents the results (pass/fail grade), date of the exam, the grader of the exam, and a signature from the candidate confirming they have reviewed the results. Maintaining the answers given by the candidate increases the number of records that need to be maintained.

Delete the requirement to maintain the answers given by the applicant or clarify the basis behind maintaining the answers given by the applicant on the initial examination.

F-36 53.815(e)

This section appears to be a copy and paste of 53.780(e).

Might be an opportunity to consolidate requirements applicable to both

Comments on NRCs Proposed Part 53 Page 78 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution interaction dependent and self-reliant facilities into a separate section.

F-37 53.815(e)(2)(iii) 53.815(e)(2)(iii) states: "Promptly correct modeling and hardware discrepancies and discrepancies identified from scenario validation and from performance testing or provide justification as to why the presence of such discrepancies will not adversely affect simulator performance with respect to the criteria of paragraph (e)(2) of this section."

This requirement is similar to 10 CFR Part 55.46(d)(2).

However, the Commission includes the term "promptly." Recommend deleting the word "promptly." The corrective action timeline may be impacted by a variety of factors such as supply chain and part availability. The timeline of the repair is the responsibility of the simulation facility owner and not the NRC.

Delete the term "promptly." This is in line with the existing language of 10 CFR Part 46(d)(2).

F-38 53.830(b) 53.830(b) states: Prior to initial fuel load (or, for a fueled manufactured reactor, prior to initiating the physical removal of any one of the independent physical mechanisms to prevent criticality required under 53.620(d)((1)), each holder of an operating or COL under this part must, with sufficient time to provide training and qualified personnel to operate the facility, establish, implement, and maintain a training Provide guidance on whether the training programs must be approved by the commission before initial use.

Additionally, provide policy statement on provisions to consider licensee training programs accredited by the National

Comments on NRCs Proposed Part 53 Page 79 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution program that demonstrates compliance with the requirements of paragraphs (c) and (d) of this section.

There is no specific guidance or direction in this section relating to if the training programs must be commission-approved before implementation. It is clear in other sections relating to licensed and generally licensed operators that the program must be approved before implementation. Additionally, will the provisions relating to the training programs accredited by the National Nuclear Accrediting Board (NNAB) be acceptable?

Nuclear Accrediting Board (NNAB) as commission approved training programs.

F-39 53.830 (c) 53.830 (c) states the following: The training program must be derived from a systems approach to training as defined in this part and must provide, at a minimum, for the training and qualification of the following categories for commercial nuclear plant personnel: (1) Supervisors (e.g., shift supervisors); (2)

Technicians (e.g., maintenance, chemistry, and radiological); and (3) Other appropriate operating personnel (e.g., auxiliary operators, certified fuel handlers, and individuals who provide engineering expertise to on-shift operation personnel).

It appears the training programs required are similar to those program areas required under 10 CFR 50 Part 120. The only difference is this rule combines the roles into specific categories (Supervisors, Technicians, and Clarify the intent and classifications of training programs required per this subpart. Specifically:

1. Intent of the category of Supervisors.

Does it include all supervisors to include technical program areas such as maintenance or does it only include operational shift supervisors like 10 CFR 50 Part 120.

2. Align definitions relating to non-licensed operator and auxiliary operator for consistency. If the intent to specify the

Comments on NRCs Proposed Part 53 Page 80 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution other operating personnel). The subpart requires clarification. 1. Does the category of Supervisors include supervisors from the technical disciplines or only supervisors overseeing operational aspects. For example, Shift Supervisor in 10 CFR 50 Part 120 has been interpreted as what is referred to as Shift Manager overseeing operational activities and does not include supervisors of other technical disciplines. 2.

Operating personnel includes engineering expertise to on-shift operation personnel. 10 CFR 50 Part 120 specifically lists Engineering Support Personnel. Is the intent in this subpart similar to what is included in 10 CFR 50 Part 120? 3. 10 CFR 50 Part 120 does not specifically call out certified fuel handlers. Is this the intent of this subpart? 4. This section references auxiliary operators: however, earlier in 53.020 Definitions, the term non-licensed operator is used. Is the intent for these two references indicating a difference or are they the same?

positions as different, specify that in 53.020 Definitions.

F-40 53.850 Radiation Protection The radiation protection requirements in 53.850 should be met with updates to note the applicability of appropriate Division 8 RGs to Part 53 licensees.

The Offsite Dose Calculation Manual requirements in 53.850 may warrant new guidance or updates to existing guidance since the existing ODCM guidance Provide as part of the Part 53 Rule package sufficient guidance to meet the requirements of 53.850. NEI believes this would be ODCM guidance for AR types (HTGR, MSR, SFR) and updates to numerous Division 8 RGs.

Comments on NRCs Proposed Part 53 Page 81 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution (e.g., NUREGs 1301/1302) focuses on PWR and BWR light water reactors licensed under Parts 50/52.

F-41 53.860, Security Programs The proposed framework for physical protection programs uses a binary approach whereby the results of a consequence analysis (per § 53.860(a)(2))

determine whether a facility is required to protect against the DBT of radiological sabotage or not. If protection against the DBT is necessary, then the applicant or licensee will need to meet the physical security requirements in either § 73.55 or § 73.100. For some facility designs, the requirements and guidance for conducting the consequence analysis are too conservative/restrictive.

With respect to the consequence analysis, the NRC should modify the requirements and guidance to allow an applicant or licensee to consider (credit) security-related design features in the consequence analysis required by

§ 53.860(a)(2) when a sufficient technical basis for such consideration can be provided. For example, an applicant or licensee could install features providing reasonable assurance of a reactor shutdown if a given event is detected (e.g., an intrusion into certain areas), thus lowering the potential consequences from a security event to levels below that specified in §53.210(b).

F-42 53.860, Security Programs When finalizing the security requirements contained in this rule, the NRC should also consider the topics and recommendations contained in:

NEI Proposal Paper, Regulation of Rapid High-Volume Deployable Reactors in Remote This comment also applies to the related security requirements in Part 73 and the associated guidance.

Comments on NRCs Proposed Part 53 Page 82 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution Applications (RHDRA) and Other Advanced Reactors, dated July 2024.

NRC Staff White Paper, Nth-of-a-Kind Micro-Reactor Licensing and Deployment Considerations, including the enclosures.

F-43 53.865 Quality Assurance If comment C-11 is accepted, the requirement for Appendix B QA for SR SSCs and optional for NSRSS SSCs should be moved here.

If comment C-11 is accepted, the requirement for Appendix B QA for SR SSCs and optional for NSRSS SSCs should be moved here.

F-44 53.870 Integrity Assessment Program The Integrity Assessment Program (IAP) remains duplicative as noted in past NEI comments on Part 53.

The purpose of the IAP is more than adequately covered by requirements in 53.220, 53.250, 53.440, 53.450, 53.715, 53.865 and 53.880. The NEI response to NRC request for public comment on the IAP provides more detail on why 53.870 is duplicative.

Remove 53.870 F-45 53.875 Fire Protection NEI appreciates the language in 53.875(b) that Each holder of an OL or COL under this part must develop a performance-based or deterministic fire protection program.

RG 1.189 and RG 1.205 should be updated to facilitate implementation of this regulation. This will require coordination with NFPA as NFPA 805 (endorsed in RG Provide updated guidance in RG 1.189 and RG 1.205 as part of the Part 53 Rule Package for meeting Part 53.

Comments on NRCs Proposed Part 53 Page 83 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution 1.205) is not technology-inclusive and relies on risk metrics of CDF and LERF which may not be applicable for non-LWRs.

F-46 53.875(a)(2)

This section provides examples of specific features to be included within a fire protection plan. Providing examples within regulation implies minimum criteria.

Remove all examples of specific features listed.

F-47 53.875(b)(2)

This section replicates the 10 CFR 50 Appendix A General Design Criteria 3 for Fire Protection.

The scope of 53.875(b)(2) implies all SR and NSRSS SSCs must meet the requirements which is inappropriate and inconsistent with RG 1.189 which limits this requirement to safe shutdown equipment.

Remove 53.875(b)(2).

If NRC decides not to remove, delete and NSRSS to be consistent.

F-48 53.880 Inservice inspection and inservice testing.

53.880 is far too detailed and prescriptive and much of the information should be moved to guidance or deleted in full. 53.440(b) already requires C&S which would specify in-service inspection and testing.

Justification for removal of specific parts is provided in the comments below.

Delete 53.880.

If kept, suggest revising to:

§ 53.880 Inservice inspection and inservice testing.

(a) Each holder of an OL or COL under this part must develop, implement, and maintain a program for inservice inspection (ISI) and inservice testing (IST) prior to receiving an OL or COL. The ISI/IST

Comments on NRCs Proposed Part 53 Page 84 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution programs must, wherever applicable, be in accordance with generally accepted consensus codes and standards that have been endorsed or otherwise found acceptable by the NRC. The ISI/IST program must include all inspections and tests required by the codes and standards used in the design and be supplemented by risk insights that identify the most important SSCs to plant safety. The types of testing and inspections and their frequency should be informed by risk insights to maintain the reliability and performance of SSCs consistent with the associated design and analyses activities involving those SSCs. Risk insights must also be used to determine when to conduct the inspections and tests (e.g.,

full power, shutdown, refueling) to minimize risk to the plant workers and the public. The ISI/IST program must be documented in a written manual and managed by qualified personnel reporting to the Plant Manager.

Comments on NRCs Proposed Part 53 Page 85 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution F-49 53.880 Inservice inspection and inservice testing.

53.880(a) (b) specifies a responsibility of the Plant Manager which may not be the job title at every plant. Suggest the more generic director, responsible officer, or designated person.

Remove Plant Manager from 53.880(a) and (b) or replace with the more generic director, responsible officer, or designated person F-50 53.880(a)

NEI understands the intent behind the discussion of consensus codes and codes but believes this is more appropriately handled through guidance. Some codes and standards are prescriptive in establishing surveillance frequencies and to use risk insights to inform inspection frequency would require deviation from the code. These intricacies are well understood, and appropriately dispositioned in various guidance documents. The 53.880 requirement should remain high level and implementation of 53.880 would be through regulatory guidance such as RG 1.246 endorsing ASME Section XI Division 2.

Delete:

The ISI/IST programs must, wherever applicable, be in accordance with generally accepted consensus codes and standards that have been endorsed or otherwise found acceptable by the NRC.

The ISI/IST program must include all inspections and tests required by the codes and standards used in the design and be supplemented by risk insights that identify the most important SSCs to plant safety. The types of testing and inspections and their frequency should be informed by risk insights to maintain the reliability and performance of SSCs consistent with the associate design and analyses activities involving those SSCs.

Update regulatory guidance including RG 1.246, and those related to the surveillance frequency control program.

Comments on NRCs Proposed Part 53 Page 86 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution F-51 53.880(a)

The line within 53.880(a): Risk insights must also be used to determine when to conduct the inspections and tests (e.g., full power, shutdown, refueling) to minimize risk to the plant workers and the public is overly prescriptive and unnecessary. The Maintenance Rule Requirements in 53.715 are sufficient for governing maintenance and inspections. The language minimize risk is ambiguous and impossible to implement. Minimal risk to the public may require inspection of equipment in a high rad area immediately, thus threatening plant workers. Minimal risk to plant workers would require all work to be completed with the plant shut down with sufficient time for radionuclide decay.

Remove: Risk insights must also be used to determine when to conduct the inspections and tests (e.g., full power, shutdown, refueling) to minimize risk to the plant workers and the public. From 53.880(a). Rely on the requirements in 53.715.

F-52 53.880(b)

While is it accurate that baseline information/data may be appropriate for comparison with subsequent operational phase inspections and testing, to imply that actual tests or inspections will need to be carried out to satisfy Part 53 requirements may be overly prescriptive.

For instance, under the provisions of ASME XI Division 2, an applicant for a CP, OL, etc. will be obligated to submit their Reliability Integrity Management program to the NRC for review and acceptance. That process should establish and determine what, if any, baseline Revise the requirement for baseline inspections to be less prescriptive and more performance-based.

Comments on NRCs Proposed Part 53 Page 87 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution examinations or tests are appropriate for a given reactor technology.

Detailed Comments on Subpart G - Decommissioning Requirements None Detailed Comments on Subpart H - Licenses, Certifications, and Approvals Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution H-1 General Subpart H is very complex and appears to be substantially longer than equivalent requirements in Parts 50 and 52 combined. The length and complexity in Subpart H likely stemming from the way the paragraphs for each license, permit, and approval type (LWA, ESP, CP, OL DC, SDA, and COL) have been structured. The approach that has been adopted of referencing the detailed technical requirements for the various application types to those in the technical requirements for a Design Certification has been a major improvement.

However, the length and complexity lead to a Streamline the requirements in Subpart H to further reduce the duplication of language and eliminate unnecessary details.

Comments on NRCs Proposed Part 53 Page 88 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution reduction in regulatory clarity and introduces the potential for similar requirements having unintended differences.

H-2 53.1124 Relationship between sections.

NEI appreciates the ability to reference existing OLs or COLs to support a standard design certification or approval. Similar reference should be appropriate for Part 50 or 52 licenses using LMP.

While it seems unlikely that an applicant would choose to transition from Part 53 to Part 52 or 50, there is no technical reason some of those approvals could not be transferred.

Consider removing under this part throughout the rule.

Part 50/52 to 53 At a minimum, remove under this Part from 53.1124(a) through (h), 53.1221, 53.1312, 53.1330(b), 53.1384(b), 53.1425, 53.1443(d), 53.1470, 53.1525, 53.1530.

Alternatively add or Part 50 or or Part 52 following under this Part as appropriate. 53.1434 requires additional references to Part 50 and 52 LWA provisions. Conforming changes and updated references will be required.

Guidance could provide additional clarity on limitations (must be an LMP application under Parts 50 or 52 using PDC that would be equivalent to the Part 53 FDC) and any additional requirements as appropriate.

Comments on NRCs Proposed Part 53 Page 89 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution The associated preamble language and some definitions in 53.020 would require conforming changes.

Part 53 to 50/52 Consider removing under this part from 53.1161, 53.1218, 53.1221, 53.1251, 52.1279, 53.1288(a)(3) to allow a Part 53 permit or approval to transition to a Part 52 or 50 application. 53.1300 should allow a transition from Part 53 to 50.

H-3 53.1188 Finality of early site permit determinations 53.1188 does not include the provision on information requests from 52.39(f), which other than for requests seeking to clarify compliance with the current licensing basis of the ESP, requires that information requests to the holder be evaluated before issuance to ensure that the burden to be imposed is justified in view of the potential safety significance of the issue. The evaluations are to be in accordance with 50.54(f).

A limit on information requests similar to 52.39(f) should be included in 53.1188, to ensure similar rigor in evaluating information requests before they are issued.

H-4 53.1203 Filing of applications 53.1203, essentially identical to 52.135(a). However, the requirements in 52.135(b) on submitting the application, and 52.135(c) on review fees being set forth in Part 170, are not included. While these are administrative matters, being specific in the Move all of the filing requirements into 53.1100, deleting them from the other Sections in Subpart H. When unique requirements for a specific application type must be addressed, they can be identified

Comments on NRCs Proposed Part 53 Page 90 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution regulation is important to ensure the application is submitted properly and to specify the review fee regulation.

This issue goes beyond 53.1203. 53.1100(a)(1) requires filing each application for a standard design approval, standard design certification, or license under part 53, or any amendments to the applications, must be submitted to the NRC under 53.040, as applicable. This captures the requirement under 52.135(b).

52.1100(e) Filing fees. Requires that [E]ach application for a standard design approval, standard design certification, or commercial nuclear plant license under this part, including, whenever appropriate, a construction permit, combined license, operating license, manufacturing license, or early site permit, other than a license exempted from 10 CFR part 170, must be accompanied by the fee prescribed in 10 CFR part 170. No fee will be required to accompany an application for renewal, amendment, or termination of a construction permit, operating license, combined license, or manufacturing license, except as provided in

§ 170.21 of this chapter. This captures the requirement under 52.135(c). However, it is not clear in 53.1100. An alternative is to keep the current Section structure but include the Section on filing of applications for every application type and make the language consistent. Where additional specific requirements must be addressed, they can be handled by paragraphs under the filing of applications Sections.

The text on filing applications should be consistent in addressing 53.1118 for the different application types.

Comments on NRCs Proposed Part 53 Page 91 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution why these specific requirements are separated from 53.1203.

There is further inconsistency in Subpart H in that paragraphs addressing filing of applications appear in 53.1143 for ESPs, 53.1203 for SDAs, 53.1233 for DCs, and 53.1273for MLs, but does not appear for Construction Permits, Operating Licenses, or Combined Licenses. There is no obvious logic behind when the provision is included and when it is not.

The specific language in the paragraphs that address filing of applications is inconsistent. In the case of 53.1273 on MLs, the language states any person may apply for an ML except those excluded from applying by 53.1118. However, since 53.1118 addresses persons ineligible for applying for or obtaining a license under Part 53, it is not clear why reference to 53.1118 is not included for all license types (e.g.,

LWA, ESP, CP, OL or COL, in addition to MLs).

These inconsistencies contribute to lack of clarity on filing requirements and have the potential for misunderstandings and increased complexity and burden in filing applications.

Comments on NRCs Proposed Part 53 Page 92 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution H-5 53.1209 Contents of applications; technical information 53.1209 (a) includes general expectations on submitting major portions of a design for a Standard Design Approval. The text notes that the scope of the application for which approval is sought must include all functional design criteria as can be identified at that stage of design. Such applicants must identify conditions related to interfaces with systems outside the scope of the major portion of the standard design for which NRC staff approval is ought and the remainder of the standard design.

While these requirements are, in principle, understandable, their implementation given the PRA-based requirements in the balance of 53.1209 warrant further explanation and guidance.

We recommend limiting the requirement to design criteria and removing requirements related to a PRA.

PRAs evaluate overall plant performance and a PRA of a major portion of the design provides little value, particularly when the major portion does not contain any radiological sources.

Develop regulatory guidance and examples, potentially including tabletop exercises, to clarify implementation of the expectations for approving major portions of an SDA.

Significantly reduce the scope of SDA application requirements informed by those tabletops.

H-6 53.1209 Contents of applications; technical information 53.1209 and 52.1239 provide the Contents of applications; technical information for Standard Design Approvals and Standard Design Certifications.

The general structure of these requirements is consistent with 52.137 for SDAs and 52.47 for DCs.

However, the specific requirements in 53.1209 and 53.1209 and 53.1239 should be revised to permit use of other risk-informed approaches.

Comments on NRCs Proposed Part 53 Page 93 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution 53.1239 Contents of applications; technical information 53.1239 derive from the PRA-based requirements in Subparts B and C. These requirements are based on the PRA-based approach addressed in the Licensing Modernization Project and do not permit use of other risk-informed approaches that may be more appropriate for very simple designs.

53.1239(a)(18) by referencing 53.450(a) requires an all-hazards PRA. As discussed in comment C-8, 43.450(a) should be updated to allow flexibility in the use of PRA. This change is even more important for SDAs, MLs, DCs and CPs where it is unrealistic to have a hazards PRA for an unbuilt plant, particularly if a site is not selected. RG 1.253 currently only requires full-power internal events PRA CPs. This guidance should be updated to allow for PRA scope to be determined by use of Section 3 of the PRA standard for all Part 53 applications, and certainly for CP, ML, DC and SDA applications.

Design certifications have traditionally been for a single unit, we suggest that multi-unit considerations be left to CPs, OLs and COLs.

53.450(a) should be modified as discussed in comment C-5.

RG 1.253 should be updated to allow flexibility in PRA scope selection for all applications.

Remove reference to 53.1239(16)

H-7 53.1221 Finality of standard design approvals; information requests.

53.1221 is essentially identical to 52.145. However, 52.145 does not include a provision akin to 53.1221(d) which states in part The Commission will require, before granting a construction permit, combined license, operating license, or NRC should revise Subpart H to include the noted language in a consistent manner.

Additionally, the requirement is inconsistently used in Subpart H, sometimes being in the paragraphs on

Comments on NRCs Proposed Part 53 Page 94 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution manufacturing license which references a standard design approval, that information supporting required design and analysis application content be completed and available for audit if the information is necessary for the Commission to make its safety determination The language appears in other Part 53 paragraphs:

52.1239 - Design Certification Technical Information 53.1221 - SDA Finality 53.1254(a) - Design Certification Renewal 53.1263(c) - Design Certification Finality 53.1416(a) - COL FSAR Technical Information 53.1263(c) states [T]he Commission will require, before granting a construction permit, combined license, operating license, or manufacturing license that references a design certification rule, that engineering documents, such as analyses, drawings, procurement specifications, or construction and installation specifications, be completed and available for audit if the more detailed information is necessary for the Commission to verify the information in the application and make its safety determination Language in 53.1221(d) and 53.1263(c) makes clear that for CPs, Ols, COLs, or MLs referencing an SDA or Finality and other times being addressed in the required application technical information. This also should be resolved so that, if incorporated, the language appears in the same paragraph for each application type.

Comments on NRCs Proposed Part 53 Page 95 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution DC, the Commission will require this information.

The language is included for the other application types, whether or not they reference an SDA or DC.

These inconsistencies contribute to lack of clarity on filing requirements and have the potential for misunderstandings during the reviews and increased complexity and burden in filing applications.

H-8 53.1251(a) & 53.1260 53.1251(a) states that a DC should be issued to be valid for 15 years. In SRM-COMDAW-24-00001, the Commission approved extending the duration of a Part 53 DC to 40 years.

53.1260 states that a DC can be renewed for no more than 15 years. Consistent with the previous comment, this should be extended to 40 years.

Part 53 should be revised to be consistent with this direction, and a DC should be able to be issued to be valid for 40 years.

H-9 53.1279 Contents of applications for manufacturing licenses; technical information 53.1279(a)(2), Design Information, points to 53.1239, Contents of applications; technical information, for Standard Design Certifications. As noted in comment H-6, the specific requirements for 53.1239 derive from the PRA-based requirements of Subparts B and C. These requirements are based on the PRA-based approach addressed in the Licensing Modernization Project and do not permit use of other risk-informed approaches or more traditional approaches that may be more appropriate for very 53.1279 should be revised to permit use of other risk-informed and more traditional approaches.

53.450(a) should be modified as discussed in comment C-8.

RG 1.253 should be updated to allow flexibility in PRA scope selection for all applications.

Comments on NRCs Proposed Part 53 Page 96 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution simple design, especially those designs where a manufacturing license might be sought.

53.1239(a)(18) by referencing 53.450(a) requires an all-hazards PRA. As discussed in comment C-8, 43.450(a) should be updated to allow flexibility in the use of PRA. This change is even more important for MLs where it is unrealistic to have a hazards PRA for an unbuilt plant, particularly if a site is not selected.

RG 1.253 currently only requires full-power internal events PRA CPs. This guidance should be updated to allow for PRA scope to be determined by use of Section 3 of the PRA standard for all Part 53 applications, and certainly for ML applications.

A manufacturing license will likely manufacture single units, multi-unit considerations should be left to CPs, Ols and COLs.

Remove reference to 53.1239(16)

H-10 53.1291 Duration of manufacturing licenses.

53.1291 states, An ML issued under this part is valid for not less than 5, nor more than 15 years from the date of issuance. In item 32 in SRM-SECY-22-0052, Proposed Rule: Alignment of Licensing Processes and Lessons Learned from New Reactor Licensing (RIN 3150 AI66), the Commission approved the proposed rule to extend the duration of manufacturing licenses from 15 to 40 years.

Part 53 should also be consistent with the 40-year duration in the Commission-approved proposed rule for a Part 52 ML.

Comments on NRCs Proposed Part 53 Page 97 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution H-11 53.1270 Manufacturing licenses The proposed Part 53 does not appear to give holders of an ML authority to make changes to the ML similar to the requirements in § 52.171(b)(1). Not authorizing changes to an ML would not allow for technological innovation after feedback from reactor operations.

Add provisions for updating MLs under Part 53 similar to § 52.171(b)(1).

H-12 53.1306 Contents of applications for construction permits:

general information This requirement supplements the general information required under 53.1109 and is specific to the information to be submitted to demonstrate financial qualifications. However, Sections 53.1660 through 53.1700 set out the requirements and procedures related to financial qualifications and related reporting requirements. For clarity and completeness, 53.1306 should include a clear reference to these Sections.

Revise 53.1306 to include clear reference to the requirements and procedures related to financial qualifications and related reporting requirements in 53.1660 through 53.1700.

H-13 53.1309 Contents of applications for construction permits:

technical 53.1309(a)(2), Design information, requires design information equivalent to that required for a standard design certification as defined in 53.1239(a)(2)-(27). This level of detail is not consistent with the notion of a Preliminary Safety Analysis Report. While the topics addressed are appropriate, the level of detail that would support a Design Certification is not. The language in 53.1309(a)(2)(ii) notes that the application may include some aspects of the design that are not fully 53.1309 should generally be revised to clearly introduce the preliminary nature of the design adequate for a CP review, consistent with 50.34. Issuance of a CP as described in 53.1333 should be consistent with 50.35.

53.1309 should be revised to permit use of other risk-informed and more traditional approaches.

Comments on NRCs Proposed Part 53 Page 98 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution developed, and the information is therefore preliminary. However, it is not clear what aspects of the design the staff would accept as preliminary. This creates a situation where inconsistent interpretation of the language creates uncertainty in the staffs expectations for the level of detail in the CP application.

Additionally, as noted in comment H-6, the specific requirements for 53.1239 derive from the PRA-based requirements of Subparts B and C. These requirements are based on the PRA-based approach addressed in the Licensing Modernization Project and do not permit use of other risk-informed approaches or more traditional approaches that may be more appropriate for very simple designs, especially those designs where a manufacturing license might be sought.

53.1239(a)(18) by referencing 53.450(a) requires an all-hazards PRA. As discussed in comment C-8, 43.450(a) should be updated to allow flexibility in the use of PRA. This change is even more important for CPs where it is unrealistic to have a hazards PRA for an unbuilt plant. RG 1.253 currently only requires full-power internal events PRA CPs. 53.1309 should be updated to be consistent with RG 1.253.

The revisions to 53.1309 also should address conforming changes to 53.1309 on Planned Research, and Programmatic controls.

53.450(a) should be modified as discussed in comment C-8.

Remove reference to 53.1279(24).

Comments on NRCs Proposed Part 53 Page 99 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution 53.1239(24) for interface requirements makes sense for SDAs, DCs and MLs, but should be removed for a CP.

H-14 53.1369 Contents of applications for operating licenses; technical information 53.1369(b), Design information, generally requires design information equivalent to that required for a standard design certification as defined in 53.1239(a)(2)-(27) excluding (8) and (10). However, as noted in comment H-6, the specific requirements for 53.1239 derive from the PRA-based requirements of Subparts B and C. These requirements are based on the PRA-based approach addressed in the Licensing Modernization Project and do not permit use of other risk-informed approaches or more traditional approaches that may be more appropriate for very simple designs.

53.1369(d), Integrity assessment program. Industry has previously commented that the integrity assessment program required in 53.870 should be removed since it duplicates other requirements and applies requirements that are not applicable until license renewal.

53.1369(g) seems duplicative of 53.1239(27),

.1369(c) is duplicative of.1239(25),.1369(d) is duplicative of.1239(13),.1369(e) & (n) is duplicative 53.1369(b) should be revised to permit use of other risk-informed and more traditional approaches.

53.1369(d) should be removed.

Remove duplicative references.

Comments on NRCs Proposed Part 53 Page 100 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution of.1239(14).1369(l) is duplicative of

.1239(22).1369(aa) is duplicative of.1239(26)

H-15 53.1416 Contents of applications for combined licenses; technical information.

53.1416(a)(2), Design information, requires design information equivalent to that required for a standard design certification as defined in 53.1239(a)(2)-(27) excluding (8) and (10). However, as noted in comment H-6, the specific requirements for 53.1239 derive from the PRA-based requirements of Subparts B and C. These requirements are based on the PRA-based approach addressed in the Licensing Modernization Project and do not permit use of other risk-informed approaches or more traditional approaches that may be more appropriate for very simple designs.

53.1416(a)(4), Integrity assessment program.

Industry has previously commented that the integrity assessment program required in 53.870 should be removed since it duplicates other requirements and applies requirements that are not applicable until license renewal.

53.1416(a)(3) is duplicative of 53.1239(25),

53.1416(a)(7) is duplicative of 53.1239(27),

53.1416(a)(14) is duplicative of 53.1239(14),

53.1416(a)(12) is duplicative of 53.1239(22),

53.1416(a)(25) is duplicative of 53.1239(26),

53.1416(a)(2) should be revised to permit use of other risk-informed and more traditional approaches.

53.1416(a)(4) should be removed.

Remove duplicative references.

Comments on NRCs Proposed Part 53 Page 101 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution H-16 53.1449 Inspection during construction 53.1452 Operation under a combined license The timelines for specific actions provided in 53.1449(a) and (c)(3) and in 53.1452(a) are reasonable for large plants or plants that have long construction periods owing to design-specific considerations. However, for small, simple plants, the overall construction schedules may be significantly shorter than what has been experienced for large plants. Some of the timelines could create administrative burdens (reporting uncompleted ITAAC 225 days before the scheduled date for initial fuel load) if the pace of construction compresses the time between specific actions and the scheduled initial fuel load date. While it is reasonable to specify timelines for the actions listed in these regulations, the specific times may not be practical. A reporting schedule based on the ITAAC completion schedule for each plant submitted under 53.1449(a) could provide a schedule that supports NRCs interests but does not impose unrealistic burdens for the licensee.

Consider revisions to reporting timelines in 53.1449 and 53.1452 that would be linked to the licensees construction and ITAAC completion schedules. This should be informed by SECY-24-0008 and NEIs proposal paper on Rapid, High-Volume Deployment of microreactors.

H-17 53.1470 Relationship between sections.

NEI appreciates the ability to reference existing CP, OLs or COLs to support licenses for identical designs at multiple sites. Similar reference should be appropriate for Part 50 or 52 licenses using LMP.

Remove under this Part from 53.1470(a).

Alternatively, add or Parts 50 or 52following under this Part. Guidance could provide additional clarity on limitations (must be an LMP application under Parts 50 or 52 using PDC that would

Comments on NRCs Proposed Part 53 Page 102 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution be equivalent to the Part 53 FDC) and any additional requirements as appropriate.

H-18 53.1118 Ineligibility of certain applicants.

The ADVANCE Act section 301 directs changes removing certain limitations on foreign ownership of some types of licensed facilities 53.1118 should be updated, consistent with the ADVANCE Act Section 301

Comments on NRCs Proposed Part 53 Page 103 of 123 Detailed Comments on Subpart I - Maintaining and Revising Licensing Basis Information Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution I-1 53.1510 Application for amendment of license The second sentence in 53.1510 is significantly more prescriptive than equivalent language in 50.90. While the analyses to support an amendment under 50.90 would address the equivalent topics using the approaches in the original Part 50 or 52 application, it is not clear why the NRC is including this level of specificity in 53.1510.

The level of detail in the second sentence of 53.1510 should be made consistent with a performance-based regulation.

I-2 53.1545 Updating final safety analysis reports The specific language in these requirements presumes the licensee has made use of the PRA-based approach in Subparts B and C. The requirements do not reflect other risk-informed approaches or more traditional approaches.

In addition, NEI 22-05 was submitted to the NRC providing industry suggestions for proposed alternates to the criteria in 50.59. This document should align with the proposed 53.1545 language.

The language in 53.1545(a)(2)-(4) should be modified to reflect FSAR updates where other risk-informed approaches or more traditional approaches have been used in supporting licensing of the plant. NEI 22-05 provides language suggestion for those following LMP.

This should inform a more methodology-inclusive set of criteria in 53.1545.

I-3 53.1545(4) 53.1545 (4) is problematic because it references 53.220 which includes the cumulative risk metric in (b). There is a contradiction where 53.450(c) and 53.1545(3) would require a PRA update every 5 years, but to assess 53.220(b) in accordance with 53.1545(4) youd need to update the PRA every 2 years.

Remove 53.1545(4). 53.1545 (3) is adequate to update PRA information in the SAR and 53.1545(1) is adequate to address changes to facility and procedures. Clarification that 53.1545(1) includes addressing cumulative changes can be provided in guidance.

I-4 53.1550 Evaluating This requirement is only applicable to one type of risk-informed approach in which the safety case is based NEI provided what it considers to be appropriate criteria for a 50.59-like evaluation

Comments on NRCs Proposed Part 53 Page 104 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution changes to facility as described in final safety analysis reports almost entirely on the PRA, and therefore cannot be applied to other risk-informed approaches. It is further noted that it appears that the outcome of the change control criteria here would be identical to the use of the criteria in 50.59, so that there is no regulatory advantage other than the ability to solely use the PRA to evaluate changes. The experience from implementing 50.59 is that the guidance is crucial to understanding how the change control criteria will be applied, and the guidance for 50.59 took significant resources and years to develop. No such guidance has been provided for 53.1550, and so the PRA-based criteria could be found to be undesirable once the details are developed in guidance.

Even if some licensees desire to solely use the PRA to evaluate whether changes need prior NRC approval, a requirement must be more inclusive and flexible.

The 10 percent change criterion, in 53.1550(a)(2)(ix) and 25 percent change in 53.1550(a)(2)(vii), is likely to be overly burdensome given the frequency and cumulative risk values for many event sequences. The more than minimal reduction in margin criteria for comprehensive risk metrics is unclear, adding regulatory uncertainty and likely overly burdensome and duplicative of other criteria.

A more realistic target to achieve the desired change threshold should be defined.

for licensees following LMP in NEI 22-05:

Technology-Inclusive Risk-Informed Change Evaluation (TIRICE). NRC released a pre-decisional Draft Guide (ML24354A075) which appears to indicate that NRC finds these criteria acceptable.

NRC should use the criteria in NEI 22-05 to inform an update to the 53.1550(a)(2) criteria.

Language should not replicate the NEI 22-05 (which is informed by LMP) but the criteria should be the base upon which more methodology-inclusive criteria are derived.

RG 1.187 with adjustments to account for additional criteria should be updated as an additional means of meeting 53.1550.

Workshops between NRC and industry can resolve gaps between the existing guidance and the Part 53 language.

The 10 percent change criterion, in 53.1550(a)(2)(ix) and 25 percent change in 53.1550(a)(2)(vii) is too prescriptive for Rule language and should be included in guidance, if at all. Consistent with NEI 22-05, the comprehensive risk metric criteria should be

Comments on NRCs Proposed Part 53 Page 105 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution exceeding the performance target, not reducing margin.

The more than minimal decrease in defense in depth is undefined, lacks guidance and should point back to exceeding the 53.250 criteria.

The method of evaluation criteria should reference 53.450(f), not 53.450 broadly. PRA Upgrades should not require a LAR.

I-5 53.1535 Amendments during construction.

The language to add flexibility for COL holders during construction is greatly appreciated and appropriately addresses the lessons learned from Vogtle 3 and 4. The Rule would benefit from clarification that 53.1535 only applies to CP holders that have requested finality on select design features or specifications and that this section would be not applicable to CP holders that have not requested this finality. This is implied by the reference to 50.35(b), but the preamble would benefit from providing clarity in the rule package itself.

Add a discussion to the preamble for the implications of 50.35(b). 53.1535 only applies to CP holders that have requested finality on select design features or specifications and this section would be not applicable to CP holders that have not requested this finality.

I-6 Subpart I,

§ 53.1565(d)(3),

Emergency preparedness program The proposed change process for emergency preparedness programs under Part 53 is essentially the same as that contained in 10 CFR 50.54(q), the process used by currently operating large LWRs. NEI believes there is an opportunity to risk-inform and improve this process.

NEI recommends the NRC adopt a risk-informed approach to these change evaluation requirements. As currently proposed, a facility may make changes to its emergency plan without NRC approval only if the changes do not reduce the effectiveness of the plan AND

Comments on NRCs Proposed Part 53 Page 106 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution the plan, as changed, continues to meet the requirements applicable to the chosen EP licensing path, either:

Appendix E to part 50 and the planning standards of § 50.47(b), or

§ 50.160.

These requirements could be risk-informed by making the reduction in effectiveness assessment applicable only to changes affecting the functions needed to implement risk-significant planning standards (RSPSs). For changes affecting other (non-risk-significant) functions, the licensee would simply need to demonstrate that the plan, as changed, continues to meet the applicable regulatory requirements. The definition of an RSPS is contained in Inspection Manual Chapter 609, Appendix B, Emergency Preparedness Significance Determination Process, and a modified version of it (e.g., to address the alternative performance-based EP framework in § 50.160) could be added to the definitions proposed in § 53.1565(d)(3)(i). This approach would be consistent with the logic used by the NRC to revise the EP Significance Determination Process (SDP), which is described in SECY-22-0089, Recommendation

Comments on NRCs Proposed Part 53 Page 107 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution for Enhancing the Emergency Preparedness Significance Determination Process for the Reactor Oversight Process.

In addition, the NRC could take this rulemaking opportunity to similarly risk-inform the requirements in § 50.54(q) by making the same changes to that section.

Comments on NRCs Proposed Part 53 Page 108 of 123 Detailed Comments on Subpart J - Reporting and Other Administrative Requirements Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution J-1 53.1630 Immediate Notification Requirements for Operating Commercial Nuclear Plants The discussion of this requirement states that it was taken from 50.72. However, it does not specify the 1-hour activation time for the data links (formerly ERDS) in 50.72(a)(4). Rather, details on activation of the data links are left to the emergency plans. Removing the activation time requirement would appear to be a relaxation so long as more restrictive requirements are not included elsewhere in Part 53.

It is unclear whether this requirement is consistent with 50.72(a)(4) in the inclusion of the criteria for declaring an Emergency Class for events of actual or potential substantial degradation of plant safety or security, probable risk to site personnel life, or site equipment damage caused by hostile action. It is not clear why these criteria have been included in 53.1630(a)(4) when they are not in 50.72(a)(4). It also is not clear that this does not duplicate, or is in conflict, with other requirements or guidance on Emergency plans.

Ensure that provisions in Part 53, or associated guidance, do not impose an activation time for the data links more restrictive than the 1-hour specified in 50.72(a)(4).

Ensure that the criteria for declaring an Emergency Class in 53.1630(a)(4) are not duplicated elsewhere in Part 53 and are not in conflict with other requirements related to Emergency Class declarations.

J-2 Subpart J

§ 53.1630, Immediate notification requirements for operating This section would establish requirements for immediate notifications. In the Statements of Consideration (SOC),

the NRC notes that A separate rulemaking activity, Reporting Requirements for Nonemergency Events at Nuclear Power Plants, has been initiated to consider possible changes to the requirements in § 50.72. At a future date, the NRC may consider reconciling future The Commission is requested to consider the potential agency and industry resource savings available if a decision was made on the staff recommended changes in SECY 0049, Proposed Rule: Reporting Requirements for Nonemergency Events at Nuclear Power Plants, before or at

Comments on NRCs Proposed Part 53 Page 109 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution commercial nuclear plants changes to § 50.72 with the requirements proposed in part 53, which have been taken or derived from the current reporting requirements.

approximately the same time as the Part 53 rule is finalized. Doing so would allow for a coordinated and parallel reconciliation of requirements and obviate the need to make separate changes at a later date.

J-3 Subpart J § 53.1630 (b)(2) & (3),

Immediate notification requirements for operating commercial nuclear plants In line with SECY-24-0049 the following requirements should be removed:

(iii) Any event or condition that results in an unplanned actuation of a safety-related (SR) standby cooling system or the unplanned sole reliance on an SR standby cooling system for those systems that are in constant operation.

This is similar to the SECY-24-0049 recommendation to remove 10 CFR 50.72(b)(2)(iv)(A).

(v) Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactively contaminated materials.

This is similar to the SECY-24-0049 recommendation to remove 10 CFR 50.72(b)(2)(xi).

Remove 53.1630(b)(2)(iii), 53.1630(b)(2)(v),

53.1630(b)(3)(ii), 53.1630(b)(3)(iii),

53.1630(b)(3)(iv) and 53.1630(b)(3)(v) and make conforming changes.

Comments on NRCs Proposed Part 53 Page 110 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution (ii) Any event or condition that results in valid actuation of an SR system, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.

This is similar to the SECY-24-0049 recommendation to remove 10 CFR 50.72(b)(3)(iv)(A).

(iii) Any event or condition that at the time of discovery could have prevented the fulfilment of the safety functions identified under § 53.230. Events covered may include one or more procedural errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, and/or procedural inadequacies. However, individual component failures need not be reported pursuant to this paragraph if other equipment was operable and available to perform the required safety function.

This is similar to the SECY-24-0049 recommendation to remove 10 CFR 50.72(b)(3)(v).

(iv) Any event requiring the transport of a radioactively contaminated person to an offsite medical facility for treatment.

This is similar to the SECY-24-0049 recommendation to remove 10 CFR 50.72(b)(3)(xii).

Comments on NRCs Proposed Part 53 Page 111 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution (v) Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, ENS, or offsite notification system).

This is similar to the SECY-24-0049 recommendation to remove 10 CFR 50.72(b)(3)(xiii).

J-4 53.1645 Reports of radiation exposure to members of the public Note that per the general comment at the beginning of the document, 53.1645 is duplicative with Part 20 requirements and should be removed. If 53.1645 is retained, this comment and the following 4 comments should be addressed Section 53.1645(a) states that licensees must submit radiological reports as required by 10 CFR Part 20. This reference is redundant and unnecessary, as Part 53 licensees are already required to comply with 10 CFR Part 20, including the reporting requirements specified in Subpart M.

Delete the phrase radiological reports as required by 10 CFR Part 20, as well as.

J-5 53.1645 Reports of radiation exposure to members of the public NRC Regulatory Guide (RG) 1.21, Revision 3, distinguishes between the terms release and discharge. According to RG 1.21, an effluent discharge refers to an effluent release that reaches an unrestricted area, while an Revise the section to use the term discharge where appropriate.

Comments on NRCs Proposed Part 53 Page 112 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution effluent release refers to the emission of effluent into the onsite environs.

For the reports required by Section 53.1645, the term discharge should be used instead of release, as a discharge specifically refers to effluent that reaches an unrestricted area. Without radioactive material reaching the unrestricted area, there is no potential exposure to a member of the public offsite (i.e., in an unrestricted area).

Using the term discharge would provide clarity and consistency with the definitions already established by the NRC. It would also ensure that these reports focus on radioactive material discharged to offsite environs that has the potential to expose members of the public.

J-6 53.1645 Reports of radiation exposure to members of the public The title of Section 53.1645 should be revised to better reflect the content of the annual reports required under this section.

Revise the section title to Reports of Discharges of Radioactive Material in Liquid and Gaseous Effluents.

J-7 53.1645 Reports of radiation exposure to members of the public Section 53.1645(a) states: If the total effective dose equivalent to members of the public in unrestricted areas during the reporting period is greater than the as low as is reasonably achievable (ALARA) design objectives established under § 53.425, the report must specify the Remove the term ALARA from Section 53.1645(a).

Comments on NRCs Proposed Part 53 Page 113 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution causes for exceeding the ALARA design objective and describe any corrective actions.

This language is inconsistent with the proposed language in Section 53.425, which appropriately no longer establishes ALARA design objectives. To ensure consistency, the language in Section 53.1645(a) should be revised to align with the updated provisions of Section 53.425.

J-8 53.1645 Reports of radiation exposure to members of the public Section 53.1645(b) states:

If during any calendar quarter the radiation exposure to a member of the public in the unrestricted areas, calculated on the same basis as the respective ALARA design objective exposure, exceeds one-half of the annual ALARA design objective exposure, the licensee must submit a report as specified in § 53.040. The report shall specify the causes for exceeding one-half the annual ALARA design objective exposure in a quarter and describe corrective actions that the licensee will take to maintain radiation exposure to levels within the ALARA design objectives for the remainder of the year. The report shall be submitted within 30 days from the end of the quarter when one-half of the annual ALARA design objective exposure was exceeded.

This language is inconsistent with the provisions in Section 53.425, which appropriately no longer establishes Remove the term ALARA in 53.1645(b).

Comments on NRCs Proposed Part 53 Page 114 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution ALARA design objectives. To ensure alignment and avoid confusion, the language in Section 53.1645(b) should be revised to reflect the updated approach in Section 53.425.

J-9 53.1670 Financial Qualifications The text for this requirement is at a very high level, simply stating that applicants under this part must possess or have reasonable assurance of obtaining the funds necessary for the activities for which the permit or license is sought. The text notes that electric utilities are assumed to have such reasonable assurance. The details on content for financial qualification are in 50.33(f) but are not included, even in a modified form, in either 53.1109 or 53.1670. Absent some level of detail on application content expected to demonstrate financial qualification, the applicant will be open to individual reviewer expectations, which could be variable from application to application and could exceed current content requirements in 50.33(f).

The language in 53.1670 should be expanded to provide an appropriate level of content requirement for a Part 53 applicant. As a minimum, 53.1670 should reference requirements that provide the technical criteria for reporting (e.g., 53.1366 and 53.1413), similar to the referencing used in 53.1690. Additional details could be in guidance, and the detail in the regulations should not go beyond expectations for a Part 50 or 52 applicant.

J-10 53.1690 Licensees Change of Status; Financial Qualifications The text in 53.1690 is nominally identical to 50.76.

However, 50.76 points to 50.33(f)(2) for details on the financial qualification information to be supplied. 53.1690 requires that the licensee must provide the NRC with the financial qualifications information that would be required for obtaining an initial operating license as specified in 53.1366 or 53.1413.

The inconsistency in which provision(s) provide the requirements for financial qualification information needs to be remedied. This is particularly a problem given the lack of detail in 53.1670.

Comments on NRCs Proposed Part 53 Page 115 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution J-11 53.1720 Insurance Required to Stabilize and Decontaminate Plant Following an Accident 53.1720(a) includes requirements on the minimum amount of insurance required for each reactor station site. In addition to the amounts specified in 50.54(w)(1),

53.1720(a) includes an amount based on plant-specific estimates of costs to stabilize and decontaminate a plant. This additional requirement is a sound addition to 53.1720(a), particularly for SMRs and non-LWRs.

However, there is no discussion of the estimation process or acceptance criteria for this amount. Absent a level of specificity, the acceptance of the estimated costs would be left to the discretion of an individual reviewer.

High-level language on the estimation process requirements and acceptance criteria should be developed and incorporated into the regulation, with more detail provided in regulatory guidance.

J-12 DRO-ISG-2023-01 DRO-ISG-2023-01 provides interim staff guidance to facilitate staff reviews of applications for an operating license or combined license under 10 CFR Part 53. The interim guidance does not provide similar guidance in acknowledging operator licensing or general licensing training programs accredited by the National Nuclear Accrediting Board (NNAB) would be considered a Commission-approved training program.

Provide clarification if facility licensee training programs accredited by the National Nuclear Accrediting Board (NNAB) would be considered a Commission-approved training program. This could be provided in the applicable definition section defining commission approved training programs and including similar wording contained in NUREG 1021, Operator Licensing Examination Standards for Power Reactors.

Provide clarification guidance in DRO-ISG-2023-01, Operator Licensing Programs Draft Interim Staff Guidance 10 CFR 53, October 2024. However, this guidance is applicable to operator licenses to include reactor operator,

Comments on NRCs Proposed Part 53 Page 116 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution senior reactor operator, and general licensed operator and would not include the requirements of 53.830 for general personnel training requirements.

J-13 DRO-ISG-2023-01, Acceptance Criteria Page 8, Criterion 2.3: An instructional designer, industrial/organizational psychologist, or a human factors specialist has determined the relevance of the measurement approach for the particular KSAs. The terms "instructional designer" or "human factors specialist" are not defined.

In the absence of an industrial/organizational psychologist (understood), please clarify whether an instructor who develops training meets the intent.

J-14 DRO-ISG-2023-01, Acceptance Criteria Please explain the purpose for requiring providing information on what is excluded from the test method and examination.

Page 11, Criterion 2.7:...a decomposition of the tasks into the knowledge, skills, cognitive processes, abilities, attitudes, or behaviors that are included in the test method, and those that are excluded from the test method Delete the requirement if there is not a compelling regulatory reason to maintain low to no value-added administrative burden.

J-15 DRO-ISG-2023-01, Acceptance Criteria Please explain the purpose for requiring providing information on what is excluded from the test method and examination.

Page 26, Criterion 9.3: Tasks are decomposed into both those knowledge, skills, cognitive processes, attitudes, or Delete the requirement if there is not a compelling regulatory reason to maintain low to no value-added administrative burden.

Comments on NRCs Proposed Part 53 Page 117 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution behaviors that are included in the examination and those that are excluded from the examination.

J-16 DRO-ISG-2023-01 Page 13 third bullet down, page 14 Criterion 2.18 statement, and page 34 Criterion 2.18: If sampling is not performed because the content domain is small enough to be tested on every examination, the applicant or licensee should state that and explain how examination security concerns related to predictability are addressed.

If a design is simple enough for the content domain to be tested on every examination, then predictability is not a concern but rather a feature. In other words, the applicants will know that every KSA required for performance of the job will be tested, but this is a desirable condition because it ensures there are no KSA gaps due to random sampling.

The final part of the statement, and explain how examination security concerns related to predictability are addressed, should be removed.

J-17 DRO-ISG-2023-02 The background section discusses frameworks A&B of part 53. Is that statement still correct?

Correct the background to align with current revision of part 53.

J-18 DRO-ISG-2023-02 The document is not clear about what on-shift support is for engineering expertise and uses examples of the traditional STA or STA/SRO implying the expertise is at site and on-shift yet makes other statements about flexibility of this person and discusses response time implying remote or even on-call remote.

Make it clear in the document that the engineering expertise will in many cases be located remotely and on-call. Additionally the document ignores NUREG 0737 discussion about the STA being temporary and could be eliminated with improvements to training and Human System Interfaces. This needs to be

Comments on NRCs Proposed Part 53 Page 118 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution part of the document as one advanced reactor design has demonstrated (with NRC SER) under parts 50/52 that an STA is not required to be part of the crew makeup due to improved training of the operators, no significant operator actions and improved HSI rendering the STA not required.

J-19 DRO-ISG-2023-02 On page 11 the commission policy statement on Engineering Expertise on Shift. The bullet language though stating, an STA is not required for Part 53 staffing plans... policy statement provides information about engineering expertise on shift. This type of language using the words on shift is not consistent with earlier language about expected alternatives to engineering expertise in the staffing plan.

Remove the term on shift from the bullet and focus on the policy statement discussion around what engineering expertise looks like as we already state it will be different than what the current fleet looks like.

J-20 DRO-ISG-2023-02 Fourth bullet in Section 7.2 page 16: location and expected response time Is response time in this context meant to be defined by the licensee?

Recommend that response time can be coming on-site or making communication with the control center (control room at site, control room at remote location for micro reactors, or just communication and data systems ready for supporting the on-shift crew.

J-21 DRO-ISG-2023-02 The term Engineering Expertise is not listed in the glossary.

Add the term engineering expertise with definition.

Comments on NRCs Proposed Part 53 Page 119 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution J-22 DRO-ISG-2023-03, Purpose "New reactor technologies are expected to have significant diversity compared to operating reactors.

Accommodating such diversity demands an approach that is flexible yet sensitive to the important differences in these technologies."

The term diversity implies the concept of "diversity and defense-in-depth." It appears this text is referring to the differences in reactor technology that will be submitted for review, not the diversity and defense-in-depth approaches for a plant.

Replace "diversity."

J-23 DRO-ISG-2023-03,

Background

This section states that NUREG-0711 and NUREG-0800 Chapter 18 will produce reviews that "may not be commensurate with the lower risk anticipated from advanced reactors and may not be adequately focused on the unique characteristics of their design and operation."

However, this document appears to be reliant on a NUREG-0711 based HFE program and does not discuss any changes to the acceptance criteria for the elements of the HFE program. It appears this document merely provides guidance for sampling but does not provide guidance for reviewing the HFE program elements.

If the intent is to replace NUREG-0711 guidance as discussed in this section, then acceptance criteria for HFE program elements should be provided.

J-24 DRO-ISG-2023-03, Rationale Paragraph 1 The term diversity implies the concept of "diversity and defense-in-depth." It appears this text is referring to the differences in reactor technology that will be submitted Replace "diversity."

Comments on NRCs Proposed Part 53 Page 120 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution for review, not the diversity and defense-in-depth approaches for a plant.

J-25 DRO-ISG-2023-03, Applicability This section limits the applicability of this ISG to Part 53 applicants and cites DANU-ISG-2022-05 as applicable to Part 50/52 applicants. These 2 documents appear to have different scopes. DRO-ISG-2023-03 describes a sampling methodology that could be used for any advanced reactor applicant regardless of licensing approach.

Expand the applicability of DRO-ISG-2023-03 to include Part 50/52 applicants.

J-26 DRO-ISG-2023-03, Acceptance Criteria Notes 3 through 5 are not used in Table 1.

Remove notes 3 through 5 J-27 DRO-ISG-2023-03, Guidance -

Overview Paragraph 1 The term diversity implies the concept of "diversity and defense-in-depth." It appears this text is referring to the differences in reactor technology that will be submitted for review, not the diversity and defense-in-depth approaches for a plant.

Replace "diversity."

J-28 DRO-ISG-2023-03, Guidance -

Overview "Appendix A to DANU-ISG-2022-01 [] provides guidance for preapplication engagement."

DANU-ISG-2022-01 is applicable to Part 50/52 applicants.

In Applicability, the NRC staff exclude Part 50/52 applicants.

If Applicability is expanded to include Part 50/52, no comment.

If Applicability is not expanded to include Part 50/52, this guidance is not applicable to the applicants and should be removed.

Comments on NRCs Proposed Part 53 Page 121 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution J-29 DRO-ISG-2023-03, Guidance -

Application Acceptance Review The first paragraph and first sentence of the second paragraph are generic guidance statements and not specific to HFE. The NRC should not duplicate generic staff review guidance requirements in multiple documents.

Remove the 1st paragraph and 1st sentence of the 2nd paragraph.

J-30 DRO-ISG-2023-03, Guidance - Facility Characterization "Characteristics identified as important" "Important" is a subjective term that will vary from reviewer to reviewer. This judgement is the basis for downstream activities, yet does not have any definition or criteria.

Define how reviewers will determine "important" characteristics.

J-31 DRO-ISG-2023-03, Guidance - Facility Characterization (2) Safety Analyses Methods and Results "significant facility and external hazards" and "event sequences deemed significant" How is significance determined? Is this referring to safety significance or risk-significance?

Define how significance is determined.

J-32 DRO-ISG-2023-03, Guidance - Facility Characterization (3) Identification of Human Actions Important to Safety "This includes actions credited in analyses of design in depth (DID) and large commercial aircraft impacts" Large commercial aircraft impact will already be considered in the safety analysis as an external hazard.

There is no need to specifically list it in this section.

Remove "and large commercial aircraft impacts."

Comments on NRCs Proposed Part 53 Page 122 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution J-33 DRO-ISG-2023-03, Guidance - Facility Characterization (4) Design Process "Appendix C describes the major elements or activities of HFE, as described in NUREG-0711, Revision 3. The list of activities in appendix C is not all-inclusive, and applicants are not required to conduct the specific activities listed.

Accordingly, the characterization should reflect the application-specific HFE activities, including any not listed in appendix C."

The relationship between this document and NUREG-0711 is very confusing. It is unclear what the use of Appendix C or NUREG-0711 is intended to be if it is neither all inclusive nor a minimum set of criteria.

Additionally, the process steps are confusing.

Characterization is described as the 1st step in developing an application specific review; however, characterization requires "application-specific HFE activities."

It would be helpful to include a process diagram that demonstrates the overall process, what information is needed at each step of the process, and who is responsible for each part of the process.

J-34 DRO-ISG-2023-03, Guidance - Grading

- Process Paragraph 4 and Reviewer Responsibilities (3)

This paragraph gives the NRC reviewer license to selectively choose whatever criteria they see fit. This would result in inefficient and unpredictable reviews.

Suggest developing a process for applicants to propose standards and guidance upon which the HFE program is developed for which the NRC would review for acceptance.

Comments on NRCs Proposed Part 53 Page 123 of 123 Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution J-35 DRO-ISG-2023-03, Appendix A-1 Paragraph 2 Figure A-1 is referenced but the figure does not exist within this document.

Include Figure A-1.

Specific Requests for Comments Part 53 Page 1 of 49 Topic 1 - Part 53, Overall Organization NRC Request Part 53 is structured as one framework with subparts providing technical, licensing, and administrative requirements for the various stages of the life cycle of a commercial nuclear plant. The organization of part 53 in this manner puts a complete set of requirements for each stage of the life cycle in a separate subpart with additional subparts for licensing and administrative requirements.

The NRC is seeking comment on the proposed organization of the requirements in part 53 and possible improvements to how specific requirements (e.g., examples of which specific sections) could be consolidated or otherwise reorganized to make the rule clearer or more concise.

There are numerous references in proposed part 53 to other NRC regulations. Examples of such references include those in proposed § 53.610 to NRC regulations related to radiation protection (part 20), FFD (part 26), physical security (part 73), and MC&A (10 CFR part 74, Material Control and Accounting of Special Nuclear Material) for facilities receiving byproduct or SNMs.

The NRC is seeking comment on whether such references to other regulations in various sections in the proposed part 53 provide benefits to applicants and licensees, or to other stakeholders seeking to understand the regulatory framework under part 53, or whether such references could be removed to reduce the length of part 53.

Industry Response The references are valuable insofar as they point to specific performance-based regulations in other parts. This is done well in 53.860 where specific regulations are pointed out for Part 73 (73.55 & 73.100). Where the reference is general and points to the entirety of a CFR Part, NEI suggests that the reference be removed (i.e., 53.260 & 53.270 references to Part 20). Enclosure 2 contains detailed comments B-7, B-8, C-1, C-3, C-5, E-1, E-4, E-6, J-4, J-7, J-8, on where references could be improved or where they should be removed.

Topic 2 - Part 53, Subpart B - Comprehensive Risk Metrics NRC Request The NRC is proposing to require the use of comprehensive risk metrics and associated risk performance objectives as one of several performance standards in part 53. Comprehensive risk metrics could include a risk metric or a set of risk metrics that approximate the total overall risk from the facility to the extent practicable. Associated risk performance objectives are pre-established values indicative of the comprehensive risk metrics that are used during risk-informed decision-making to gauge plant safety.

Specifically, comprehensive risk metrics and associated risk performance objectives would provide one element of the safety criteria for LBEs other than DBAs in the proposed § 53.220. Comprehensive risk metrics, in the form of the IEFR and the ILCFR, and associated risk performance objectives, in the form of the QHOs of 5x10-7 per year and 2x10-6 per year, respectively, were similarly used in the LMP methodology to ensure that other evaluation criteria were conservatively defined and as a tool for focusing attention on matters important to managing the risks posed by nuclear power plants. The use of such comprehensive risk metrics and associated risk performance objectives in an integrated risk-

Specific Requests for Comments Part 53 Page 2 of 49 informed decision-making process is similar to that used in RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3.

The NRC is seeking comment on the use of comprehensive risk metrics and associated risk performance objectives in part 53 as one of several performance standards. The IEFR and ILCFR and the QHOs represent comprehensive risk metrics and associated risk performance objectives that the NRC has used for decades in a variety of capacities. What other performance standards could be used to address the comprehensive risks posed by proposed commercial nuclear plants? Please provide your considerations and rationale for your recommendation.

If an applicant proposes a novel approach to comprehensive plant risk and the NRC approves the approach, should the resulting NRC-approved comprehensive plant risk metrics and associated risk performance objectives be codified or otherwise memorialized over time and, if so, how?

Industry Response Please see comment B-3 in enclosure 2 of these comments for a more fulsome response to the NRCs inquiry about risk metrics, including the rationale for why 53.220(b) should be assessed in the broader context of defense in depth. To answer the 2nd question first, the comprehensive risk metric should not be codified in the regulation but instead endorsed in regulatory guidance similar to RG 1.174s endorsement of CDF and LERF. This allows flexibility that is essential for the differences between new reactor designs. Some traditional LWRs may choose to keep the CDF and LERF risk surrogates, others may choose to use the risk metrics from LMP and endorsed in RG 1.233, others may chose a technology-inclusive (TI) risk surrogate (see discussion below) and microreactors may chose consequence based limits for a bounding event or set of events as suggested in DG-1414. A set of bounding events is likely more practical, perhaps with a bounding event crediting SR SSCs to address events with frequencies > 1E-4 and a maximum credible accident (MCA) bounding event to address frequencies < 1E-4. A simplified calculation could show that the SR-crediting consequence times a frequency of 1 plus the MCA consequence times a frequency of 1E-4 provides reasonable assurance the QHOs will be met.

Building off research conducted through EPRI, NEI intends to put forward a risk surrogate for Individual Latent Cancer Fatality Rate (ILCFR) for NRC consideration of endorsement in 2025.

Tentatively, the industry is proposing an Offsite Dose Exceedance Frequency (ODEF) representative of the annual frequency of events exceeding to-be-determined dose limit at the Exclusion Area Boundary (EAB). The NEI report intends to calculate an appropriate performance objective for ODEF based on preliminary AR PRA information and methodology like what defined CDF in NUREG-1860.

As industry develops their TI Risk Metric they will consider whether, and under which conditions, a second metric is needed. Many advanced reactors, particularly microreactors, expect an IEFR of 0. The low frequency of ILCFR for such reactors will likely provide confidence that both the ILCFR and IEFR targets will be met. However, there may be reactors where larger releases are possible so it may be valuable to have a small ODEF (SODEF) for the frequency of events with releases between 1 and 200 rem (as an example) and a large ODEF (LODEF) for frequency of events with releases above a higher rem threshold.

Specific Requests for Comments Part 53 Page 3 of 49 Several microreactors expect to have no credible accidents that could result in an offsite dose

>100rem which was used as the lower threshold dose for early fatality in NUREG-1860. 100 rem for a bounding dose is a well-justified surrogate for IEFR. DG-1414 discusses Determination of a demonstrably conservative risk estimate for comparison with the QHOs and puts forth an annual frequency of 1/yr as acceptable with further discussion that multiple postulated bounding events could be combined to support demonstration that the QHOs are met. While NEI is not ready to put forward a methodology for a minimal set of dose criteria that could align with the QHOs, one potential approach would be a dose crediting SR SSCs multiplied by the 1/yr annual frequency summed with a maximum credible accident dose multiplied 1E-4. Industry believes 1E-4 is justified because traditional SR Design criteria ensure that SR SSCs fail at a frequency < 1E-4 or lower. The ASME Code focuses on 1E-4, high winds design criteria is based on a 1E-7 frequency, seismic design criteria factors in margin to the 1E-4 value, etc. As a broader point, the ADVANCE Act directs the commission to consider alternatives to probabilistic risk assessments so the Part 53 language should be broad enough to allow such an approach.

Also important for consideration with respect to comprehensive risk metric is the treatment of event sequences from hazards, non-core sources and lower modes. The PRA standards provide screening criteria for many of these which remain valid for screening hazards, sources, and modes from consideration for a comprehensive risk metric. But the non-LWR PRA Standard also provides the ability to use supplemental evaluations in lieu of PRA. This adds nuance to the calculation of risk metrics that need to be accounted for in regulatory guidance. One means, that has precedent in the operating fleet is the use of hazard penalties. RG 1.76 as an example provides wind speeds that correlate to a 1E-7 annual exceedance frequency. Using design basis accident analysis crediting only SSCs designed to withstand that 1E-7 design basis wind a licensee can provide assurance that high events will contribute no more than 1E-7 to CDF and LERF (or another established risk metric) since there is high confidence that the SR SSCs will perform during the DB wind event. Enclosure 1 provides more examples of this for various hazards.

Topic 3 - Part 53, Subpart B - Defense in Depth NRC Request Proposed § 53.250 would establish requirements based on the longstanding NRC philosophy of providing defense in depth to address uncertainties concerning the design, operation, and performance of commercial nuclear plants during LBEs.

The NRC is seeking comment on the inclusion of the proposed requirements to assess and provide defense in depth. The NRC is also seeking comment on whether to include specific provisions in § 53.250 and subpart B to more explicitly address the possible role of inherent characteristics of some SSCs in preventing or mitigating unplanned events. The proposed § 53.250 is worded to preclude relying on a single engineered design feature to address the range of LBEs other than DBAs, which could possibly allow crediting inherent characteristics without further lines of defense. How could possible inherent characteristics of SSCs be considered in the proposed requirements in § 53.250 or in any alternative requirements for defense in depth provided in response to this item? Please provide your considerations and rationale for your recommendation.

Specific Requests for Comments Part 53 Page 4 of 49 Industry Response NEI appreciates the clarification in the preamble that The phrase engineered design feature would not preclude the possible crediting of inherent characteristics within the design and analysis for commercial nuclear reactors. With this preamble language maintained, the 53.250 requirements may be appropriate, and further guidance can be developed in regulatory guides similar to RG 1.233 which should expand the NRC endorsement of NEI 18-04 as applicable to Part 53. NEI provides in Enclosure 2, comment B-6 rationale for why this requirement is not necessary, but if it is kept, we suggest wording that we believe is clearer without changing the intent.

More clarity could be provided for how this concept can be acceptable to the NRC beyond the methodology endorsed in RG 1.233. There should be clear criteria, building on the information in SECY-18-0096 for a more conservative approach to design criteria for radionuclide retention as an example (see enclosure 6 for more detail on this proposal). An applicant using conservative methodologies should be able to set SR design criteria for the reactor, likely limited to the TRISO fuel particles and fuel form (whether a pebble, compact or other medium), based on the DBA analysis. DID could then be provided via special treatment requirements around the measurement and monitoring of the Specified Acceptable Radionuclide Release Design Limit (SARRDL). The traditional approach to DID should also be allowed. While the traditional approach would require the use of the single failure criterion and have implications across Part 53, it remains a completely valid method for achieving adequate protection of public health and safety and should remain an option Under Part 53.

Topic 4 - Part 53, Subpart C - Probabilistic Risk Assessment NRC Request Current consensus PRA standards provide processes for appropriately defining the scope of a PRA and determining applicability of supporting requirements to suit the specific needs of a given applicant under proposed part 53. In addition to assessing other aspects of PRA acceptability such as PRA peer reviews, NRC determinations of the acceptability of such PRAs would assess the appropriateness of the applicant defined scope as part of determining the applicability of a consensus PRA standard supporting an application. This approach is consistent with the current state of practice and offers appropriate flexibility for PRAs to be developed and assessed based on the application they are used to support, which includes consideration of how PRA results and insights are relied upon, together with factors such as safety margin, simplicity of design, and treatment of uncertainty.

The NRC is seeking comment on what additional guidance, if any, is needed regarding PRA acceptability for Part 53 applicants and licensees.

Industry Response NEI appreciates the acknowledgement that This approach [existing processes in PRA standards for defining PRA scope] is consistent with the current state of practice and offers appropriate flexibility for PRAs to be developed and assessed based on the application they are used to support. We agree that this is aligned with industry methods and appreciate the language in RG 1.253 that The level of detail in a CP PRA should be established using the

Specific Requests for Comments Part 53 Page 5 of 49 process provided in Section 3 of ASME/ANS RA-S-1.4-2021, Risk Assessment Application Process. We believe this flexibility is essential and should be applied to ALL Risk-Informed applications. This means updating RG 1.253 to provide clarity that the PRA scoping process in the standard is acceptable for applications other than construction permits. Specifically, Part 53 should allow a graded approach to PRA which would allow the applicant to choose among the following approaches for the scope of the PRA1:

1. An All-modes, All-sources, All-Hazards PRA that greatly informs the licensing basis.

2. Non-core sources handled through traditional means and regulatory requirements (i.e., Standard Review Plan fuel handling accidents with the traditional regulatory requirements for spent fuel pools and spent fuel storage on site)

3. Lower Modes handled through traditional means and regulatory requirements (i.e., control rod withdrawal DBAs from lower power levels)

4. Hazards handled through traditional means and regulatory requirements (e.g., Appendix S seismic requirements, Appendix R Fire requirements, etc.)

5. DBAs confirmed, as-needed, by Risk Insights.

NEI is concerned by slides shared during a January 16th, 2025 public meeting indicating: As currently envisioned, this flexibility would be used in limited cases (e.g., seismic analysis) where state of the art knowledge of very low frequency events may not yield useful PRA results. We believe the non-LWR PRA standard Section 3 provides an appropriate framework for determining PRA scope. This framework should be acceptable for all Part 53 applications and may justify limiting PRA scope to full power internal events or even a maximum hazard analysis of other supplemental evaluations. We believe this approach, in addition to being aligned with the Commission directive in the SRM, is also supported by the 1995 PRA Policy Statement.

Given the dissimilarities in the nature and consequences of the use of nuclear materials in reactors, industrial situations, waste disposal facilities, and medical applications, the Commission recognizes that a single approach for incorporating risk analyses into the regulatory process is not appropriate. However, PRA methods and insights will be broadly applied to ensure that the best use is made of available techniques to foster consistency in risk-based decision-making. The 1995 Policy Statement also notes it is recognized that there may be situations with material users where it may not be cost-effective to use PRA in their specific regulatory applications.

The argument that DBAs need only be confirmed by risk insights is the most controversial for Part 53 and arguably not in line with the commission directive in the SRM. It is important to note that most of the existing operating fleet was licensed without the use of PRA. With an approach informed by DG-1414 and maximum credible accident analysis, industry believes there is a path to a licensing basis determined by traditional design basis accidents (DBAs) confirmed by a PRA that would entirely remain within the Owner scope and made available for NRC audit. In such a framework, the PRA would serve as a check that the traditional DBAs were in fact bounding (e.g.

that no risk insights would challenge assumptions made in the DBA). It is important to note that reliance on more traditional means of risk assessment would be one means of implementing 1 All of the below may utilize the screening criteria in PRA standards to screen out modes, hazards and sources while still meeting NRC requirements for PRA adequacy or acceptability.

Specific Requests for Comments Part 53 Page 6 of 49 ADVANCE Act Section 208 which directs the Commission to develop risk-informed and performance-based strategies and guidance to license and regulate micro-reactors including strategies and guidance for risk analysis methods, including alternatives to probabilistic risk assessments As an example, the NRC has a well-established set of guidance for seismic design criteria and nuclear codes and standards for structural design. If an applicant follows these traditional standards to design their reactor then, consistent with current precedent under Part 50, a Seismic PRA should not be required. A seismic margins analysis can provide confidence that the QHOs are met, but the potential advantages to a more risk-informed seismic design would be given up.

The above seems to be aligned with the discussion of Section 53.415: These requirements would support either traditional approaches for determining and protecting against external hazards or probabilistic approaches that are being developed for seismic and some other external hazards. Clarity could be gained from explicitly stating the criteria by which NRC determines a hazards PRA or lower modes PRA is necessary. Part of this answer would be the screening criteria in the PRA standard, but the other important consideration was guidance would be appreciated is on the application they are used to support. This should build on NRC precedent where development of a certain PRA scope is an entry condition to specific risk-informed applications:

RICT, SFCP, 50.69 PRA requirements in line with existing guidance.

NFPA 805 would require a Fire PRA, otherwise traditional Fire Protection in line with 50.48(a) is sufficient.

A Seismic PRA would be required to use LMP to deviate from traditional seismic design criteria and allow approaches such as ASCE 43-19.

A high winds PRA could justify deviation from traditional RG 1.76 high winds design criteria.

An external flood PRA could justify deviation from traditional external flooding design in line with RG 1.59.

An internal flood PRA could justify deviation from SRP 3.4.1 for internal flooding design.

To help provide clarity on the various means of RIPB methodologies complying with Part 53, NEI had submitted in September 2021 a White Paper on Technology-inclusive, Risk-informed, Performance-based Approaches for Development of Licensing Bases Under Part 53. NEI has built on this discussion with Enclosure 1 which provides more detail on how existing regulatory guidance can comply with Part 53 and the limited set of Rule language that we recommend facilitating this more flexible approach.

Topic 5 - Part 53, Subpart C and D - Earthquake Engineering NRC Request Proposed § 53.480 would establish requirements related to seismic design considerations. This proposed section is intended to provide a clear connection between siting activities and seismic design activities and to support various approaches to presenting seismic hazards and addressing those hazards

Specific Requests for Comments Part 53 Page 7 of 49 in designs. The proposed requirements are intended to provide sufficient flexibility to allow approaches like those currently in parts 50 and 100 or approaches that might be endorsed by the NRC in the future that could incorporate more risk insights from PRAs.

The NRC is seeking comment on whether the proposed requirements for earthquake engineering provide appropriate flexibility in addressing seismic risks while also ensuring that the regulations continue to adequately address seismic hazards.

Please provide your considerations and rationale for your recommendation.

Industry Response While the proposed requirements provide appropriate flexibility in most cases, there needs to be significant supporting regulatory guidance that is not currently available. DG-1410 is currently pre-decisional and should be published to support the Part 53 Rulemaking. As NEI has commented on that pre-decisional draft guide, Options 2 and 3 are promising and could allow appropriate flexibility. Publishing DG-1410 as RG 1.251 will not be sufficient to provide appropriate flexibility under Part 53. As part of the endorsement of ASCE 43-19 in RG 1.251, the NRC should consider, in line with the industry comments on pre-decisional DG-1410, the endorsement of ANS 2.26 which would allow selection of return periods more frequent than 1E-4 is justified by consequence analysis.

NRCs Nth of a Kind (NOAK) microreactor white paper (ML24268A310) provides a graded approach to site characterization and uses SSHAC as an example. This is a step in the right direction, but it does not go far enough. A truly graded approach could use USGS NHSM and not require a site-specific SSHAC. RG 1.208 currently is the only approved methodology for establishing the design basis seismic hazard and does not allow such flexibility. NEI has proposed further flexibility in a report on rapid high-volume deployment of microreactors (ML24213A337). To support Part 53, NRC, working with industry should endorse a graded approach to site characterization necessary and sufficient to establish the design basis seismic hazard.

The requirement in 53.480(c) (vi) on soil structure interaction is overly prescriptive and may not be appropriate for some sites and designs. This is recognized in NUREG-0800, 3.7.2: for sites where SSI effects are considered insignificant and fixed base analyses of structures are performed, bases and justification for not performing SSI analyses are reviewed by the NRC on a case-by-case basis. If the SSI analysis is not required, the input motion at the base of the structures will be the design motion reviewed in SRP Section 3.7.1" Since Part 53 is intended to be performance-based and minimize exemptions, the NRC should consider language such as should consider, if appliable soil-structure instead of must take into account Finally, 53.440(b) and numerous requirements in 53.440 and 53.480 reference both SR and NSRSS SSCs when SR would be more appropriate. 53.415 correctly limits the ability to survive the DBHLs (including seismic events) to SR SSCs and the equipment qualification requirements in 53.440 and 53.480 should align with 53.415. 53.440(b) is particularly important as it requires generally accepted consensus codes and standards that have been endorsed or otherwise found acceptable by the U.S. Nuclear Regulatory Commission (NRC). for systems that are SR or

Specific Requests for Comments Part 53 Page 8 of 49 NSRSS. The NRC has not endorsed codes such as ASCE 7, AISC360 and ACI 318 for NSRSS structures and such endorsement is essential for appropriate requirements for NSRSS structures. As discussed elsewhere, NEI would prefer that 53.440(b) be limited to SR structures, but if the language remains as-is, industry will require the endorsement of ASCE 7, AISC360 and ACI 318 for NSRSS structures.

Topic 6 - Part 53, Subpart E - Construction and Manufacturing NRC Request Proposed § 53.610(b)(1)(iii) would require procedures that describe how construction will be controlled so as not to impact other features important to the design (e.g., dewatering, slope stability, backfill, compaction, and seepage).

The NRC is seeking comment on whether such specific requirements are useful or whether these requirements could be met through other requirements proposed in Part 53 or already present in other relevant regulations (e.g., quality assurance requirements in appendix B to part 50).

Industry Response The requirements in 53.610, specifically (a)(1)-(4) & (6) which themselves are arguably redundant with Appendix B QA requirements are sufficient to meet the intent of 53.610(b)(1)(iii). Specifically, Criterion III for Design Control requires translation of design basis requirements for SSCs are translated into procedures. Therefore, features like dewatering, slope stability, backfill, compactionall features that relate to the protection of the plant from external hazardswould be required to be addressed as a matter of addressing that requirement. 53.610(b)(1)(iii) should be deleted as duplicative.

Topic 7 - Part 53, Subparts E and H - Manufacturing Licenses NRC Request 1 The proposed requirements governing manufacturing are set forth in subpart E, and the proposed requirements governing the licensing processes are contained in subpart H. Some of the proposed requirements, including provisions related to the loading of unirradiated fuel into a manufactured reactor, are intended to cover a factory fabrication model that has been suggested for some micro-reactor designs. However, as written, the proposed provisions are not limited to any size or type of reactor.

The NRC is seeking comment on whether the proposed regulations are sufficient to govern various scenarios for the possible manufacturing and deployment of manufactured reactors.

If a comment indicates that the proposed regulations are not sufficient, please describe the reasons why, including, if applicable, any plausible scenario for which the commenter believes the proposed regulations are not sufficient.

Industry Response NEI believes the proposed regulations are mostly sufficient, but there is a gap with regards to factory testing of fueled microreactors. NEI appreciates the efforts to develop a white paper for

Specific Requests for Comments Part 53 Page 9 of 49 factory testing of fueled microreactors under Part 53, but NEI would propose a different approach that we believe is more efficient relying on existing guidance instead of new requirements.

The rule package should provide more clarity on the relationship between Part 53 and Part 70 requirements. In our view, the Part 70, Subpart H possession and use license is essential and has been tested over time to implement the appropriate safety measures up to and including fuel loading. The provisions of Part 70 provide performance requirements for radiological, chemical, fire, and criticality safety. These performance requirements are implemented through an Integrated Safety Analysis that forms a fundamental part of the licensing basis for a facility and operations within it, management measures, and change management processes. Additionally, Part 70 invokes other regulatory requirements that provide adequate assurance of public health and safety (e.g., Parts 20, 26, 51, 73, 74, etc.). Attempts to either bypass or restate the comprehensive requirements of 10 CFR 70 in Part 53 have the potential to miss requirements that are essential to safety and safeguards of the activities in the factory, create contradictions between requirements of Part 53 and Part 70 (the case with the subcriticality language regarding at least two independent physical mechanisms versus double contingency principle), and create confusion regarding which regulation is operative.

Duplicative requirements include:

Requiring licenses issued under 10 CFR Part 70 for loading fuel into a manufactured reactor to comply with 10 CFR Part 70, Subpart H, Additional Requirements for Certain Licensees Authorized to Possess a Critical Mass of Special Nuclear Material. Subpart H in 10 CFR Part 70 already contains criteria for the types of licenses that must comply with Subpart H. In addition, since Subpart H was codified prior to conceptualizing factory-fabricated reactors, the regulation contains the phrase or any other activity that the Commission determines could significantly affect public health and safety, which should be the standard for being subject to Subpart H, not requiring a compliance with Subpart H in 10 CFR Part 70 via 10 CFR Part 53. In addition, the staff who review applications under 10 CFR Part 70 would have a much better understanding of whether performing an integrated safety analysis would be beneficial for each specific manufacturing facility.

The requirement in § 53.620(e)(2) appears to be duplicative with § 53.620(e)(4). The regulations in § 53.620(e)(2) state that any contract to transport the reactor must contain language requiring the person transporting the manufactured reactor comply with all shipping requirements in applicable NRC regulations, certificates of compliance, and NRC-issued licenses. The requirements in § 53.620(e)(4) require compliance with 10 CFR Part 71. By including certificates of compliance in § 53.620(e)(2), this part of the proposed rule appears to indicate that the manufactured device contains fuel, and therefore, the shipper (the licensee possessing the material) must include in a contract that the carrier must comply with all shipping requirements in applicable NRC regulations, certificates of compliance, and NRC-issued licenses. The licensee is responsible for ensuring that the shipment, regardless of which carrier transports the radioactive material, complies with all NRC requirements in 10 CFR Part71, including 10 CFR 71.5, which requires compliance with the requirements of the Department of

Specific Requests for Comments Part 53 Page 10 of 49 Transportation, which included package, marking and labeling, placarding, and road, rail, air, and vessel modal requirements.

Unduly restrictive requirements include:

Limiting shipment of a manufactured and loaded microreactor to destinations with a COL. If a reactor vendor wanted to export a manufactured reactor, it would need an export license and, if the proposed rule becomes final, an exemption to the requirement in § 53.620(e)(1), since deployment sites outside the U.S. would not have a COL issued by the NRC. This requirement unfairly limits U.S. reactor vendors or requires them to have fabrication facilities outside the U.S. The requirement should be revised to include deployment sites for which NRC has issued an export license, under 10 CFR Part 110.

Limits the duration of a manufacturing license (ML) to between 5 and 15 years, where the license duration for fuel fabrication facilities can be up to 40 years, which is the same license duration as operating reactors. The duration of an ML should be extended to coincide with Commission-approved proposed rule for Part 52 (see item 32 in SRM-SECY-22-0052, Proposed Rule: Alignment of Licensing Processes and Lessons Learned from New Reactor Licensing (RIN 3150 AI66).)

Absent a safety finding, fabrication should be able to start any time when the ML is in timely renewal. The NRC does not require fuel fabrication facilities to stop making fuel or a reactor to shut down when it is within 30 days of expiration of the license, therefore, the manufacturing license should be able to be in timely renewal if submitted within the timeframe provided in the rule for renewal and the ML holder should be able to continue fabrication of reactors authorized under the ML, with the understanding that it would be fabricating at risk if the ML is not renewed.

A reactor in which there are two independent features to preclude criticality is not designed or used to sustain nuclear fission, The NRC should not limit the number of microreactors that a manufacturing facility can produce over its lifetime, but rather, based on safety and security analyses, limit the quantity of special nuclear material onsite and the number of reactors fabricated and being loaded with fuel at a given time.

The proposed Part 53 does not appear to give holders of an ML authority to make changes to the ML similar to the requirements in § 52.171(b)(1). Not authorizing changes to an ML would not allow for technological innovation after feedback from reactor operations.

NEI suggests that any attempt to control the possession and use of SNM up to and including fuel loading be relegated, by reference in Part 53, to the requirements of 10 CFR 70 and the Part 53 manufacturing license findings be constrained to those attributes that are relevant to the ultimate reactor operational safety and safeguards. This might include in-factory inspections (e.g., ITAAC) of safety significant features operating the reactor. NEI believes that ITAAC that may currently require an operating license under Part 50 or 52 should be carried out with a class 104 license under Part 53. This allows the use of well-established guidance without being overly burdensome for reactor operation that is very limited in scope.

Specific Requests for Comments Part 53 Page 11 of 49 Additional clarification in Subpart E could also be provided to differentiate the Part 53 requirements for manufactured reactors that do not include in-factory fueling activities (and therefore would not require a 10CFR70 and supporting licenses) from the requirements addressing SNM and radioactive material activities for a manufactured reactor fueled at a factory.

While NEI disagrees with the language that defines a fueled reactor as a utilization facility (Item 7 below), even if this position is adopted the comment remains valid regarding separation of the regulatory finding into Part 70 and Part 53 where the Part 53 finding would only authorize assembly and fueling at a Manufacturing facility licensed under Part 53 but would require a separate safety finding consistent with Part 70 for all SNM operations in a factory setting, including reactor fueling.

NRC Request 2 The proposed regulations in subpart H allow holders of or applicants for a COL to reference an ML but do not include such a provision for the holder of or applicant for a CP or OL. This proposed change from the current relationship between subparts in part 52 and the part 50 licensing process was made to simplify the provisions in the proposed part 53 for licensing and deploying manufactured reactors.

The NRC seeks comment on whether part 53 should include provisions for an applicant for or a holder of a CP or an OL to reference an ML and, if so, how this should be done.

Industry Response Given the variety of business models that might exist in the future it seems prudent to allow an applicant for, or holder of, a CP and particularly for an OL to reference the ML. Given the design maturity and finality provisions of an ML, it would be expected that a CP referencing that ML would have preliminary information only associated with minor site-specific matters, but this may be of commercial interest and benefit to certain projects. If a provision to reference an ML in a CP application is provided, and subsequent construction is managed while an OL is reviewed and issued, there would need to be provision for managing the ITAAC provisions in the ML in an alternative fashion (e.g., those inspections, tests, and analyses being incorporated into a sites construction quality assurance program and general NRC construction oversight, but not used as a license condition requiring closure notification and associated finding.) Based on initial concepts developed by NEI and its members for RHDRA reactors, there are deployment scenarios where use of the CP/OL licensing approach appeared to provide benefit in efficient organizing regulatory decisions with site-specific work, which still presumed use of an ML as the authority for manufacturing the microreactor ahead of final site selection, largely for perceived administrative challenges in ITAAC management.

NRC Request 3 Proposed § 53.1295 states that the holder of an ML could not begin manufacture of a manufactured reactor less than 6 months before the expiration of the license. This limitation is similar to the current restriction in § 52.177, which states that the manufacture of a reactor cannot begin less than 3 years before the expiration of the license. The restriction was revised from 3 years in part 52 to 6 months in

Specific Requests for Comments Part 53 Page 12 of 49 the proposed part 53 in recognition of the likely use of MLs for a factory-fabrication model for microreactors.

The NRC seeks comment on whether it is necessary or appropriate to revise the 3-year restriction in part 52 on when manufacturing activities could begin in relation to license expiration and, if so, what that restriction should be.

Industry Response Six months may be appropriate. However, this may not be the most relevant metric. The more important question might be related to whether the ML holder (presumably coupled with a Part 70 possession license) can go into a timely renewal period (as is currently done with possession licenses) allowing continued possession of the SNM while the license renewal is under review.

This question becomes related to Item 7 below where not defining the fueled module as a utilization facility could facilitate continued possession under the Part 70 license even after the ML has expired. It is also relevant to the response in Item 1 above where we believe a Part 70 license is essential for safe and secure possession and use. Rather than a discrete temporal limit, the renewal activity could be managed with generic timely renewal period language or as a license condition based on the MLs specific technology implementation. There is also an opportunity under Part 53 to extend the duration of an ML to 40 years, to be consistent with current Part 70 license terms and SRM-SECY-22-0052, which proposes to extended MLs under Part 52 to allow for 40 year licenses.

NRC Request 4 Proposed § 53.1288 provides the finality provisions for MLs and includes, as does existing § 52.171, limitations on the NRCs imposition of new requirements on either the design or the requirements for the manufacture of a manufactured reactor. No MLs have been issued under part 52 and there is no practical experience with the proposed finality sections. While the implications of the finality provisions related to the design of a manufactured reactor can reasonably be inferred from experience with DCs and COLs, there is no experience or available guidance regarding finality for requirements for the manufacture of the manufactured reactor.

The NRC is seeking comment on the proposed finality provisions for MLs and specifically if and how finality for manufacturing processes might be requested and used.

Industry Response It is unclear from our reading of the draft regulation what requirements are actually being imposed by Part 53 for the manufacture of the manufactured reactor beyond the ability to load fuel in the factory and requirements to maintain subcriticality by at least two independent physical mechanisms. In our response to Item 1 above we state our view that the Part 53 license finding be constrained to the safe and secure operation of the reactor and that safely handling and loading fuel in the Factory be managed under the Part 70 license. If that is done, there are essentially two safety findings required, one to handle and load SNM (Part 70) and the second related to the safety of the reactor in operation (Part 53) and the finality is cleanly associated with the two distinctly different findings. Regarding finality of reactor design-related

Specific Requests for Comments Part 53 Page 13 of 49 matters, it appears logical that the similarity between DC and COL experience would carry into a Part 53-issued ML from the language selected.

NRC Request 5 The NRC is seeking comment on the proposed regulations for the loading of fresh (unirradiated) fuel into a manufactured reactor for subsequent transport to a site for which the Commission has issued a COL that authorizes construction and operation of a commercial nuclear plant using the manufactured reactor. The proposed regulation includes provisions for loading of fuel into manufactured reactors at a manufacturing facility prior to transporting the fueled reactor to its deployment site, as suggested by some stakeholders. The NRC has historically viewed reactor operation as including fuel load, and existing NRC regulations reflect this view. While the Act authorizes the NRC to issue licenses to manufacture production or utilization facilities, it does not contain specific provisions on fueling or operating facilities licensed under an ML, and existing ML regulations under part 52 do not include provisions for fuel load.

The proposed rule addresses this matter by allowing an applicant to combine an ML with a part 70 license, which would authorize possession of a manufactured reactor in which the licensee has loaded unirradiated fuel provided at least two independent criticality prevention mechanisms are in place, each of which is sufficient to prevent criticality assuming optimum neutron moderation and neutron reflection conditions. This requirement would limit the possibility of creating fission products and allow the control of SNM, so that the loading of the fuel into a manufactured reactor could be governed primarily via a part 70 license and associated regulations (including those in subpart H of Part 70).

A specific topic on which the NRC is seeking comment is on the potential benefits of and issues with including the requirements of subpart H of part 70 within the proposed regulations for loading fuel into manufactured reactors at the manufacturing facility. For example, should the NRC include a threshold for including the requirements of subpart H of part 70 and, if so, what factors and decision criteria should be considered in such a threshold? If a comment indicates that the proposed regulations are not sufficient, please describe the reasons why, including the plausible scenarios for which the proposed regulations would not work or could be made to work better.

Industry Response While the request for comment states [e]xisting ML regulations under part 52 do not include provisions for fuel load, these regulations also do not expressly prohibit use of a Part 70 SNM license to load fuel into an existing reactor. The staff deems itself unable to authorize fuel load due to the Commissions past view that operations start at fuel load, which does not inhibit manufacture of large light-water reactors and thus issued SECY-24-0008, Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory, to authorize fuel loading in a factory. However, for microreactors, with two methods to control criticality, which satisfy the double contingency principle, the safety evaluation should show that there is no possible way for an inadvertent criticality event, even during an accident or natural phenomena, as such, the NRC should not consider the apparatus to be a nuclear reactor.

NEI believes that the requirements of 10CFR70 should be adequate for safely loading fuel into a reactor at the factory and that there is no need for the additional requirement of at least two independent criticality prevention mechanisms are in place, each of which is sufficient to

Specific Requests for Comments Part 53 Page 14 of 49 prevent criticality assuming optimum neutron moderation and neutron reflection conditions.

This is based on decades of experience from licensees handling SNM and assembling reactors and components for the Government under a Part 70 license. Adding additional requirements beyond the long-established criticality prevention performance criteria of 10CFR70 and described in detail in Appendix A of NUREG-1520 is not only unnecessary but adds significant and potentially confusing layers of requirements above long-established double contingency principle practices proven over decades of operations with SNM. These additional requirements could also preclude the otherwise safe loading of certain reactor types without the need to disassemble at the use site to remove physical mechanisms.

As an example of potential confusion and misunderstanding, even the language in the NRCs document (ML24283A027) becomes inconsistent. In this question posed by NRC they refer to, at least two independent criticality prevention mechanisms are in place, each of which is sufficient to prevent criticality assuming optimum neutron moderation and neutron reflection conditions., however in the balance of the document they refer to at least two independent physical mechanisms in place, each of which is sufficient to prevent criticality assuming optimum neutron moderation and neutron reflection conditions. Whether the mechanisms are physical or not, the wording seems to unnecessarily exceed the double contingency principle in that it requires two separate mechanisms. 10CFR70.4 defines the double contingency principle as process designs should incorporate sufficient factors of safety to require at least two unlikely, independent, and concurrent changes in process conditions before a criticality accident is possible. As described in NUREG-1520 Revision 2, one mechanism can provide control of two separate parameters, or a single process can be controlled such that at least two independent failures involving the parameter would be required before a criticality accident is possible.

Material licensees have successfully utilized the double contingency principle throughout their operations for decades. While implementation of 10CFR70 for fuel cycle facility activity management provides a significant base of positive operational experience, the implementation of those principles for reactor fueling, fresh and spent fuel management, and related activities does have precedent in the community. We believe adequate guidance for factory fueling activities can be developed as necessary to provide clarity for 10CFR70 license holders who will manage them.

NRC Request 6 Section 170, Indemnification and Limitation of Liability, of the Act states that each license under section 103 shall have as a condition of the license a requirement that the licensee have and maintain financial protection of such type and in such amounts as the NRC shall require.

The NRC is seeking comment on whether the proposed regulations should include amounts of required financial protections for MLs for fueled manufactured reactors, and, if so, what would be appropriate amounts of required financial protection.

Industry Response NEI believes indemnification under Price Anderson is warranted to flow directly to the Manufacturing License necessitating the need for the licensee to provide financial protection, however the amounts of protection required should be commensurate with the factory

Specific Requests for Comments Part 53 Page 15 of 49 fabrication activities and limitations preventing criticality. This is predicated on the fact that the ultimate reactor operator will be afforded indemnification and will be required to provide financial protection commensurate with the operating reactor license. Precedent from existing CX license holders may provide benefit in establishing reasonable limits and amounts of financial protection commensurate with such operations.

NRC Request 7 Some stakeholders have suggested that a fueled manufactured reactor with appropriate protections against criticality should not be categorized as a utilization facility under NRC regulations or Section 11cc. of the Act.

The NRC is seeking comment on possible approaches where the NRC could find that a fueled manufactured reactor would not be a utilization facility, the basis for such a finding, and the potential benefits of and potential issues with such a finding.

Industry Response In SECY-24-0008 and in the proposed Part 53 rule (SECY-24-0008: Micro-Reactor Licensing and Deployment Considerations: Fuel Loading and Operational Testing at a Factory (nrc.gov),

ML24283A027), NRC Staff proposed options for Commission consideration regarding factory fueling of micro-reactors. Staff proposes that fueling of a reactor in a factory setting requires a utilization facility license even if the measures described in the SECY and Proposed Rule to preclude criticality are present and maintained such that achieving or sustaining nuclear fission in a self-supporting chain reaction is not credible.

NEI believes this conclusion overly complicates the licensing processes that are currently employed and provide adequate protection. Specifically, loading fuel under a possession and use license (10CFR70), transporting fuel under 10CFR71, and storing used fuel under an ISFSI license (10CFR72) can be done safely and securely (specifically preclude criticality under these regulations) without the additional administrative burden of requiring a utilization facility license.

Relevant Language from the Atomic Energy Act The term "utilization facility" means (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 peculiarly adapted for making use of atomic energy 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. This is consistent with the definition proposed for Part 53.

Note: Reactor or nuclear reactor do not appear to be defined in the AEA.

53.020 Commercial nuclear reactor means an apparatus, other than an atomic weapon, designed or used to sustain nuclear fission. For the purposes of requirements in this part that

Specific Requests for Comments Part 53 Page 16 of 49 reference requirements in 10 CFR part 50, a commercial nuclear reactor is equivalent to a nuclear reactor as defined in 10 CFR 50.2.

The Commission has the authority, through its forthcoming direction to the NRC staff on SECY-24-0008, to clarify as a policy matter that a device specifically designed and controlled to preclude criticality (as described in Staffs paper and the proposed rule) is neither a nuclear reactor nor a utilization facility until such time as it is intentionally manipulated to sustain nuclear fission in a self-supporting chain reaction. Up to that point, criticality prevention is adequately addressed in NRC regulations in 10CFR70.61d, the risk of nuclear criticality accidents must be limited by assuring that under normal and credible abnormal conditions, all nuclear processes are subcritical, including use of an approved margin of subcriticality for safety.

Preventive controls and measures must be the primary means of protection against nuclear criticality accidents. NEI believes that Part 70 offers a better and safer framework for preventing inadvertent criticality than any other regulation and guidance suite the NRC currently has in place. It is also fully incorporated into ANSI Standards that have been accepted by NRC.

If the staff felt constrained by the definition of nuclear reactor in 50.2 when writing SECY 0008, Part 53 allows for an updated definition that would allow greater flexibility. As the staff has previously noted, Section 11cc. of the Atomic Energy Act of 1954, as amended (AEA),

specifies that the Commission may determine by rule what constitutes a utilization facility.

Definition of a Utilization Facility; Direct Final Rule, 79 Fed. Reg. 62,329, 62,331 (Oct.17, 2014)

Thus, we suggest removing designed or from the Part 53 definition of a commercial nuclear reactor, which should support a conclusion that the fueled manufactured reactor is not a utilization facility until measures to preclude criticality (in place in accordance with part 70 regulations) are removed. By not defining fueled manufactured reactors as utilization facilities, the NRC would remove redundancies with Parts 70, 71 and 72 requirements and facilitate the transportation of microreactors in line with congressional directives in Section 208 of the ADVANCE Act.

Similarly, Parts 71 and 72 provide adequate prevention against inadvertent criticality without the administrative burden of issuing a utilization facility license.

Absent an explicit Commission policy determination on this issue in response to SECY-24-0008, the Part 53 rulemaking offers a pathway to establish by rule of the Commission that a utilization facility license is not required. That would simplify the licensing process and associated hearing-related requirements when it comes to factory assembly and transportation.

During the ACRS meeting regarding SECY 24-0008, the Committee also appeared to question the purported need to designate a microreactor fueled at a manufacturing facility as a Utilization Facility.

Why does this Matter?

The designation of a vessel containing fuel as a utilization facility ripples into various parts of the life cycle of a factory manufactured/fueled module. This is illustrated in the table below.

While the operational reactor phase is clearly a utilization facility, the classification of the other phases as a utilization facility brings additional regulatory burden and uncertainty about what

Specific Requests for Comments Part 53 Page 17 of 49 needs to be done either technically or administratively to adapt current existing and adequate regulatory frameworks to the designation as a utilization facility.

Phase of Lifecycle of Manufactured Reactor Regulatory Framework Attribute Assured by Regulatory Framework Relevant to Utilization Facility Designation Approval of reactor design 10CFR50/52/53 Approving the design of the manufactured reactor which will be operated as a utilization facility Initial fuel and fueled component fabrication and assembly (assures subcriticality) 10CFR70 Features to Preclude Criticality and Subcriticality assured by 10CFR70 Fuel loading of the factory fabricated module 10CFR70 Features to Preclude Criticality and Subcriticality assured by 10CFR70 Transport of unused reactor to use site 10CFR71 Features to Preclude Criticality and Subcriticality assured by 10CFR71 Pre-use storage at the use site Combination of 10CFR 50/52/53 and 10CFR70 Features to Preclude Criticality and Subcriticality assured by 10CFR70 and fresh fuel storage conditions Use of the reactor at the site 10CFR50,52 or 53 Safe and secure operation of a utilization facility Post-use storage of fueled reactor 10CFR50,52 or 53 Safe and secure handling of radioactive material under Parts 50.

52 or 53 Long-term storage of used factory fabricated fueled module on site after use 10CFR72 (with exemption as needed for cooling periods < 1 year) and fuel stored within the reactor Subcriticality assured by 10CFR72 Transport of used factory fabricated fueled module on site after use 10CFR71 Subcriticality assured by 10CFR71 Interim or long term Storage 10CFR72, 10CFR60/63 Subcriticality assured by 10CFR72/60/63

• Note that the table above lists potential options based on microreactor deployment model.

Additionally, as a practical matter, it is possible that a facility that fuels a reactor may also be more vertically integrated. Due to the cost of licensing and maintaining an NRC-licensed facility, some businesses may combine fueling a reactor with a fuel fabrication facility capability that could begin at the early stages of fabrication and transition to fueling a factory fabricated module. These activities would necessitate a license under 10CFR70, including Subpart H, which already has all necessary provisions and processes in place to assure subcriticality. Therefore,

Specific Requests for Comments Part 53 Page 18 of 49 requiring a utilization facility license for such facilities provides no safety or security benefit and would impose unnecessary administrative burden. Such a result would be inconsistent with the intent of ADVANCE Act Section 208,Regulatory requirements for micro-reactors, which directs the NRC to develop and implement risk-informed and performance-based strategies and guidance in eight areas, specified in subparagraphs 208(a)(1)(A)-(H), for the licensing and regulation of micro-reactors. Throughout development and implementation of these strategies and guidance, the NRC must consider the unique characteristics of micro-reactors, including physical size, design simplicity, and source term; opportunities to address redundancies and inefficiencies; and other policy and licensing issues, including those discussed in SECY-20-0093, Policy and Licensing Considerations Related to Micro-Reactors. By logical extension, this directive also should apply to SECY-24-0008, in which the NRC staff explicitly acknowledges that

[t]he requirements for 10 CFR Part 70 licenses better match the technical and safety aspects of loading fuel into a micro-reactor with features to preclude criticality than the requirements of 10 CFR Part 50 and 10 CFR Part 52 that apply to an operating utilization facility.

NRC Request 8 Proposed requirement § 53.620(d)(2)(i) would require a security program, including a physical security plan, for any ML authorizing possession of a manufactured reactor into which fuel has been loaded at the manufacturing facility. Currently, requirements in § 73.67(c)(1) only require that a physical security plan be submitted for those licensees who possess, use, transport, or deliver to a carrier for transport SNM of moderate strategic significance, or 10 kg or more of SNM of low strategic significance.

The NRC is seeking comment on whether the proposed requirement: (1) should be specific to the facility type (i.e., manufacturing facility) or be specific to the category of material being used at the facility; (2) should apply to all manufacturing plants, including those at which licensees may only possess SNM of low strategic significance (i.e., category III), or only those facilities for which an applicant must submit a physical security plan per § 73.67(c)(1); or (3) should include more specific requirements on the supplemental security measures that may be needed for licensees possessing SNM of moderate strategic significance (i.e., category II)?

Industry Response Similar to our response in Item 1 above, NEI believes the safeguards and security aspects associated with the manufacturing facility that is authorized to possess SNM are best handled independently of Part 53 and incorporated into the Part 70 possession license through current regulation language. Our understanding is that Category III safeguards measures for protecting SNM are currently adequate in 10 CFR 73/74 and that NRC is treating Category II possession on a case basis due to the small number of applicants seeking to possess Category II quantities. While long-term changes to Parts 73 and 74 may be warranted, we do not believe this should be codified on Part 53 and is better left treated by existing regulations related to material possession.

NRC Request 9 Proposed requirement § 53.620(d)(2)(i) would require a cybersecurity program. The proposed general cybersecurity performance requirements would be to provide reasonable assurance that a cyberattack

Specific Requests for Comments Part 53 Page 19 of 49 could not adversely impact the functions performed by digital assets used by the licensee for implementing the physical security, radiation monitoring, and criticality requirements.

The NRC is seeking comment on the following: (1) to what extent stakeholders envision physical security controls, radiation monitoring, and criticality controls at a manufacturing facility being digital; (2) to what extent should the ML holder be required to protect digital computer and communications systems that impact safety and security functions from a cyberattack at a manufacturing facility authorized to load fuel; and (3) whether the rule provides sufficient clarity on the cybersecurity measures needed for license issuance or if additional detail should be included either in the rule or in guidance?

Industry Response NEI believes that Cyber Security requirements for possession licenses are currently being adequately implemented and there is minimal value in codifying these into Part 53 for the possession and fuel loading activities conducted under a Part 70 license at the Manufacturing Facility. We do acknowledge there may be components of cyber security that need to be implemented at the manufacturing facility to prevent issues from manifesting at the reactor operating site under the OL or COL.

NRC Request 10 Proposed requirement § 53.620(d)(2)(i)(B) would require that the physical security program be designed to prevent unintended and uncontrolled criticality events. This would include criticality events that are initiated maliciously.

The NRC is seeking comment on whether the ML holder should be required to design its security program to protect against radiological sabotage (i.e., an unintended criticality event leading to unacceptable radiological consequences), in addition to theft and diversion. For example, should the NRC establish security requirements to prevent an adversary, including an insider, from tampering with the reactor at a manufacturing facility or during transport in such a way as to cause an inadvertent criticality event? If so, should the NRC consider factors such as the category of fuel and the number of reactors at a factory that can simultaneously be loaded with fuel in establishing the security requirements?

Industry Response

§ 70.22(j)(1) requires a general licensee to a licensee safeguards contingency plan for dealing with threats, thefts, and radiological sabotage, as defined in part 73 of this chapter for possession of special nuclear materials licensed under Part 70. Physical security requirements can provide reasonable assurance that a malevolent act by an intruder or insider threat would not lead to a criticality event. The security requirements should be revised to ensure that an individual cannot either make sufficient changes to the reactors control systems or move a sufficient quantity of special nuclear material to an unsafe, moderated geometry to achieve an unintended criticality event. During other phases should fall under separate licenses, such as Part 73 for transportation.

NRC Request 11

Specific Requests for Comments Part 53 Page 20 of 49 Proposed requirement § 53.620(d)(2)(i) would require an ML holder to meet the performance objectives in § 73.67. Requirements § 73.67(e) and § 73.67(g) include provisions for security of category II and category III quantities of SNM, respectively, during transportation.

The NRC is seeking comment on the extent to which the ML should require ASMs (i.e., security measures above those required by § 73.67(e) and § 73.67(g)) for transportation of a fueled reactor to its place of operation. What should those measures be?

Industry Response Similar to previous responses, codifying transportation security requirements should be done in Parts 71 or 73 respectively.

NRC Request 12 Proposed requirement § 53.620(d)(2)(i) would require an ML holder to meet the performance objectives of § 73.67. For licensees utilizing a category II quantity of SNM, the requirement in § 73.67(d)(4) would have the ML holder conduct a screening to confirm the identity of an individual prior to granting unescorted access to the controlled access area where the material is used or stored. The purpose of this requirement is to both confirm the identity of the individual and support a determination that the individual is trustworthy and reliable.

The NRC is seeking comment on whether the ML requirements should include ASMs (i.e., measures beyond those required by § 73.67(d)(4)) in order to provide reasonable assurance of identity confirmation and trustworthiness and reliability.

Industry Response Relegate to the Part 70 license and existing requirements and methods. No need for duplication and potential contradictions.

NRC Request 13 The NRC is seeking comment on whether provisions regulating the testing of fueled manufactured reactors in the manufacturing facility should be included in part 53 and, if so, what would be practical for the holder of an ML while also providing adequate protection of public health and safety. One possibility could be COLs that would be issued to the holders of an ML to cover low power (e.g., <5%

rated thermal power) nuclear physics testing of fueled manufactured reactors within the manufacturing facility prior to the manufactured reactors being transported to and incorporated into a commercial nuclear plant for the purpose of energy production. The NRC recognizes configuration changes are needed to perform nuclear physics testing and is seeking comment on what requirements should apply to the manufactured reactors and the manufacturing facility during such testing (e.g., limiting power levels). If a comment indicates that the regulations should address limited operations at manufacturing facilities, please describe the likely scenarios that would need to be addressed and suggest what would be appropriate requirements for such scenarios.

While an ML holder could accomplish nuclear physics testing by applying for a COL under the proposed subpart H of part 53, stakeholders have indicated that many of the requirements would likely be unnecessary, given the reduced risk profile posed by such activities. Therefore, the NRC is seeking

Specific Requests for Comments Part 53 Page 21 of 49 comment on what requirements in subpart H of part 53 should apply to applicants for a COL who would perform testing of fueled manufactured reactors at the manufacturing plant. Examples of proposed requirements that might be relaxed or modified for applications for low power testing at manufacturing plants include those related to selection of LBEs to reflect limited inventory of radionuclides and decay heat, aircraft impact assessments, and earthquake engineering.

Additionally, the NRC is seeking comment on whether several other requirements in part 53 could be modified for applications for a low power testing COL at a manufacturing facility. For example, the NRC is seeking comment on how portions of the ML facility used to support testing should fall within the requirements for construction activities under § 53.610; whether §§ 53.710 and 53.715 (SSC configuration control) must be implemented to ensure portions of the ML facility relied on to limit potential radiological consequences from LBEs are available to perform their safety functions; and whether the requirements of § 53.730 could be modified to reflect the conditions of low power physics testing. If a comment indicates that some design and analysis requirements and related application requirements in subpart H of the proposed part 53 are not needed for the testing of fueled manufactured reactors, please provide a rationale supporting your comment and, if applicable, what alternate requirements would be appropriate.

Moreover, the licensing mechanism for the facility could present unique challenges. One option could be to issue a low power testing COL for each fueled manufactured reactor to be tested. This would comport with the agencys practice of issuing one license per reactor but could prove prohibitive from a cost standpoint and may provide very little safety benefit if all manufactured reactors are the same.

Alternatively, one low power testing COL could be issued for the portions of the ML facility used to test the fueled manufactured reactors and allow multiple fueled manufactured reactors to be completed and tested over the course of the ML. Under this approach, any ITAAC related to testing of the fueled manufactured reactors would need to be closed after they were manufactured but prior to testing, and the NRC would issue a notice of intended operation and provide the public an opportunity to request a hearing on whether each fueled manufactured reactor as constructed complies, or on completion will comply, with the acceptance criteria of the license. The NRC is seeking comment on the potential benefits and issues with having a COL for each fueled manufactured reactor to be tested versus having a COL cover the testing of multiple fueled manufactured reactors. If a comment indicates a preference for a particular approach, please provide a rationale supporting the comment and describe the specific scenarios that the regulations need to address.

Industry Response It is important to note that some testing may be achievable under a Part 70 license and a clear delineation between which testing is achievable under a Part 70 vs which would require a Part 50, 52 (or 53) license is worth further exploration. Subcritical testing of reactivity controls, reactivity manipulation, or possibly even neutron multiplication measurements with external neutron sources that can meet the Part 70 performance requirements for subcriticality should fall into the Part 70 regime. Subcritical Testing that cannot meet the performance requirements of Part 70 for subcriticality would enter the regime of a utilization facility and require a Part 50/52 or 53 license. Critical and lower power testing would be in the utilization facility regime.

Similar to NEIs position on Part 70 discussed in Topic 7, request 1, we believe that an applicant who wants to perform critical testing on the site should obtain license based on guidance in

Specific Requests for Comments Part 53 Page 22 of 49 NUREG-1537. There is a long history to the regulation and guidance of test reactors including experience with non-LWRs. Given Part 53 as proposed does not currently cover Class 104 licenses one path would be to obtain a Part 50 license that can be combined with a Part 70 and Part 53 license. This would require the ability to transition between Part 50 and 53 licenses. A Part 50 Class 104 license developed using NUREG 1537 as guidance is specifically suited to the unique activities of the critical experiments and provides appropriate assurance of safety and safeguards for the experiment or test. We recognize this may lead to a combined license at a particular site with up to 3 safety findings (ML FSAR for reactor design and operational plant safety under Part 53, Possession license for safely and securely manufacturing and fueling under 10CFR70, and critical experiment/test (non-power) reactor license under Part 50). Another option is adding class 104 licenses to Part 53, but again relying on proven existing guidance. An applicant would need to lay out which testing will be performed under each regime with limitations for each and a link to ITAAC closure. Similarly, programs should address the overlap of licensing regimes in the given application as needed.

Industry is aligned that new license categories such as an operating license for testing a manufactured reactor at the manufacturing facility increase regulatory uncertainty, add undue burden on an applicant and are unnecessary given existing guidance for Part 70 licenses (NUREG-1520), research and test reactors (NUREG-1537) and the licenses already proposed in Part 53 or existing in Parts 50 or 52. We do appreciate the idea that one low power testing COL could be issued for the portions of the ML facility used to test the fueled manufactured reactors and allow multiple fueled manufactured reactors to be completed and tested over the course of the ML and that should be explored, but in the context of a NUREG-1537 class 104 license.

Topic 8 - Part 53, Subpart F - Staffing and Generally Licensed Reactor Operators Under the Act Sections 106 and 107, the NRC is proposing to group commercial reactors into classes upon the basis of the similarity of operating and technical characteristics of the facilities, and then to prescribe uniform conditions for licensing individuals as operators of any of the various classes; determine the qualifications of such individuals; and, for certain classes of commercial reactors, issue general licenses (i.e., licenses for which no application is needed) to such individuals allowing the individuals to operate the commercial reactor.

NRC Request 1 Categories of Individuals Who May Manipulate Facility Controls: The NRC is proposing requirements that would allow the manipulation of the controls of certain facilities by GLROs in lieu of specifically licensed reactor operators and senior reactor operators. Reactor operators and senior reactor operators are the only categories of individuals currently allowed to be licensed to manipulate the controls of utilization facilities under part 55.

The NRC is interested in public perspectives on this proposed addition of the GLRO category, particularly in light of new reactor technologies and concepts of operations.

Industry Response

Specific Requests for Comments Part 53 Page 23 of 49 NEI Supports this concept. If a plant meets the criteria specified in 53.800(a)(1) - (5), then there are no required operator actions or credible human errors that would impact the safe operation of the plant. Having the licensee hold the General License is sufficient.

Guidance should be provided on what meets the Defense in Depth (DiD) criteria under 53.800(a)(5). We would recommend a time frame of 7 days with no operator intervention needed as the first layer of DID, as the ARs are less complex, have improved HSI and significantly more time (on the order of days or weeks, vs hours for current fleet) for accident sequences to progress without risk of core damage.

NRC Request 2 Criteria for GLRO Staffing: The NRC is proposing criteria under which facilities would be staffed by GLROs in lieu of specifically licensed reactor operators and senior reactor operators. These criteria establish a new class of self-reliant-mitigation facilities, as defined in part 53, for which distinct GLRO licensing and staffing requirements would apply.

The NRC is soliciting public feedback regarding whether these proposed criteria are appropriate and what, if any, alternative criteria should be considered. Please provide your considerations and rationale for your answer.

Industry Response In general, NEI is supportive of the concept, however there are two points that need to be made. The first is the criteria for being a self-reliant mitigation facility are not time bound with respect to human interaction. In other words, if human interaction is needed in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> vs 24 days, the criteria would not be met for Defense in Depth as an example. Historically mission times for systems were 7 days or less with a few out to 14 days. If defense in depth has a clear time-based limit defined either by the NRC or by the licensee, such that actions taken after that would be additional positive actions to keep the plant in a safe condition, then more plants could take advantage of the GLRO. This time-based criteria should be established via guidance.

The second is there does not appear to be a significant reduction in the administrative burden to licensing operators as a GLRO as opposed to a RO/SRO. Unless there is more clarity provided about the delta, it is unclear why additional effort would be expended to make a change.

NRC Request 3 Medical Requirements for GLROs: Based on the proposed criteria that a self-reliant-mitigation facility, as defined in part 53, must meet, the NRC is proposing not to subject GLROs to requirements for medical fitness and medical examination. This is in contrast with the proposed requirements associated with specifically licensed reactor operators and senior reactor operators, as well as the existing requirements for reactor operators and senior reactor operators under part 55.

The NRC is soliciting public feedback regarding whether GLROs should be subject to medical fitness and/or medical examination requirements like reactor operators and senior reactor operators. Please provide your considerations and rationale for your answer.

Industry Response

Specific Requests for Comments Part 53 Page 24 of 49 It is reasonable to not require medical requirements through regulation. If the plant meets the criteria set to be a Self-reliant-mitigation facility (SRMF), then the GLROs actions are not needed to protect the health and safety of the public. The GLROs should be subject to the same requirements as the remaining plant personnel, who are not subject to additional medical requirements. Facilities should still be able to apply their own fitness/medical criteria based on the job demands. The general fitness for duty requirements should be sufficient to address the fitness for duty of a GLRO. To maintain the trust of the public, it is expected that GLROs will have fitness for duty requirements imposed by the licensee.

NRC Request 4 Onshift Engineering Expertise: The NRC is proposing to require that engineering expertise be accounted for within facility staffing plans. This proposed requirement would be in lieu of the traditional position of the Shift Technical Advisor. The NRC is further proposing that individuals providing such engineering expertise would need, among other things, to possess either a qualifying 4-year degree or licensure as a Professional Engineer.

The NRC is interested in feedback from the public regarding the appropriateness of this requirement, including any alternatives that should be considered. Please provide your considerations and rationale for your answer.

Industry Response The industry and new designs have progressed to the point that the STA position (or equivalent onshift engineering expertise) could be sunset. The regulation should offer a path to do so through training programs and Human System Interface design. This would be consistent with NUREG 0737 language that identified the STA as a temporary position until both training and HSIs were improved. The training programs have improved and the new design HSI is improved along with the less complicated accident sequences for Advanced Reactors (AR). In fact, there is precedence for an AR vendor under part 50/52 licensing who demonstrated no STA needed for the control room crew due to the improvements mentioned above as part of an NRC observed staffing validation plan at the AR simulator facility. The staff agreed that no STA was needed based on their observation and the associated SER discussed the reasons for that conclusion.

NEI agrees with broadening engineering expertise (see discussion on training below, which is not necessarily a degree), however, disagrees with phrasing that suggests or requires that expertise to be continuously onshift. Recent small Light Water Reactor precedent suggests this is not required due to reduced reliance on operator actions, results of a task analysis and validation activities, and industry upgrades to qualifications of operators. Quick access to Engineering Expertise should not require the actual physical presence of a person on site. NEI proposes that Engineering Expertise as a support function be available to assist the operations, however, this can be done remotely. In todays technological designs, monitoring the plant from a remote location would provide the necessary information to an off-site engineer and allow them to communicate with the plant operators. Additionally, if the plant design is such that no actions are needed for a specified period of time, then that should be factored into when that assistance would be required to be on site. For example, if there are no actions needed for 7 days, then engineering expertise should not be required to be on site for 7 days.

Specific Requests for Comments Part 53 Page 25 of 49 We also disagree with the proposed requirement that engineering expertise can only be acquired by a qualifying 4-year degree. Many components of the training upgrades that were implemented throughout the nuclear industry due to the TMI action plan provide a better foundation for critical thinking in a nuclear power plant than any reliance on the widely varied curriculum of a 4-year degree. These are:

Math, physics, thermodynamics, and component design topics that are of specific relevance to the operation of a nuclear power plant Training for mitigating core damage Plant-specific training on the following topics:

o Plant systems o Plant specific reactor technology (including core physics data) o Plant chemistry and corrosion control o Reactor plant material o Reactor plant thermal cycle o Transient/accident analysis o Emergency procedures (including basis documents)

As noted above, we recommend this be part of the support organization and not be on shift staffing nor require a 4-year degree. No credit has been considered in this proposed rule for the NRCs post TMI criterion needed to remove the onshift engineering expertise.

NRC Request 5 Use of Simulation Facilities as HFE Testbeds: The NRC is proposing to establish regulations pertaining to the use of simulation facilities within the context of the licensing programs both for specifically licensed reactor operators and senior reactor operators as well as for GLROs. However, these regulations, as currently proposed, do not address the use of simulation facilities within the context of serving as testbeds for HFE-related analyses and assessments. Rather, the NRC currently envisions that the use of simulation facilities as HFE testbeds is more appropriately addressed via guidance documents.

The NRC is soliciting public feedback regarding whether simulation facility requirements should also address the use of simulation facilities as HFE testbeds. Please provide your considerations and rationale for your answer.

Industry Response NEI agrees that establishing a standard is needed. The current lack of clear guidance adds additional work to efforts to perform Independent Staffing Validation. It may be better served in guidance vs regulation as methods change and improve over time to allow the staff to be more efficient with making changes in the future to match the industrys ability to change.

Topic 9 - Part 53, Subpart F - Emergency Preparedness and Security Programs NRC Request 1

Specific Requests for Comments Part 53 Page 26 of 49 The proposed framework for part 53 would incorporate the changes to NRC regulations from the final rulemaking on Emergency Preparedness for Small Modular Reactors and Other New Technologies (the EP for SMR/ONT rule) by including references to § 50.160, Emergency preparedness for small modular reactors, non-light-water reactors, and non-power production or utilization facilities, and by making conforming changes within § 50.160. The proposed framework for part 53 would also introduce a graded approach to physical protection requirements that includes the criterion in § 53.860(a)(2)(i) to establish a class of licensees that would not be required to protect against the design-basis threat (DBT) of radiological sabotage. The NRC is soliciting public comment relating to these topics, which could include ways that graded approaches for both emergency preparedness and security programs might be assessed and considered during the licensing process.

The NRC is seeking comment on the sufficiency and clarity of requirements in proposed part 53 related to the assessments needed to support graded emergency planning and security. If a comment indicates that there is an issue with the sufficiency or clarity of the proposed regulations, please describe the reasons why, including, if applicable, any scenario for which the proposed regulations are not sufficient and possible ways to clarify the requirements. The NRC is specifically seeking comment on possible challenges arising from the interactions between the proposed regulations and related assessments for grading the requirements for emergency planning and security.

Industry Response Please see our response to the question below.

NRC Request 2 The NRC is preparing various guidance documents to support this rulemaking and other ongoing or recently completed rulemakings related to emergency preparedness and security. DG-5076, Guidance for Technology-Inclusive Requirements for Physical Protection of Licensed Activities at Commercial Nuclear Plants, has been issued along with this proposed rulemaking and public comments are requested via this notice on that draft guidance. The NRC is also planning to issue a draft revision of RG 1.242, Performance-Based Emergency Preparedness for Small Modular Reactors, NonLight-Water Reactors, and Non-Power Production or Utilization Facilities, for public comment. The planned revision to RG 1.242 would add guidance for part 53 applicants and licensees.

In the staff requirements memorandum to SECY-23-0021, the Commission directed the NRC staff to address the consideration of security-related events for an advanced reactor that addresses security through design and engineered safety features when it harmonizes this rulemaking with the EP for SMR/ONT rule. In the EP for SMR/ONT rule, the NRC established an alternative performance-based and risk-informed approach for emergency planning, including determining the need for and size of an emergency planning zone (EPZ) to support predetermined, prompt protective actions. The NRC has incorporated the relevant rule language from the EP for SMR/ONT rule into this proposed rule and is seeking stakeholder feedback as to whether additional rule language changes or additional guidance would be beneficial.

In light of the Commission direction and the above considerations, the NRC is assessing how best to address the treatment of security-related events in emergency planning, including in the determination of EPZ size, for reactors licensed under Part 53. Part 53 is introducing an alternative approach to

Specific Requests for Comments Part 53 Page 27 of 49 meeting security regulations that should be taken into consideration under § 50.160. Stakeholders are encouraged to take a holistic view of the various activities and opportunities to provide comments on this rulemaking and related guidance supporting this rulemaking (e.g., DG-5076 on physical protection requirements, future revisions to RG 1.242). In developing comments, the NRC urges stakeholders to consider various scenarios that might arise when implementing graded approaches for security and emergency planning for various reactor designs. Scenarios could include the following:

the potential consequences from security events up to and including the DBT of radiological sabotage are bounded by unlikely and very unlikely event sequences such that security events do not need separate analyses in the EPZ size determination; the potential consequences from security events up to and including the DBT are not bounded by unlikely and very unlikely event sequences but could otherwise support a reduced EPZ size consistent with considerations discussed in RG 1.242 and NUREG- 0396, Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants; or the potential consequences from security events up to and including the DBT are not bounded by unlikely and very unlikely event sequences and warrant consideration of increasing the size of the EPZ.

The NRC is interested in comments on the need for additional rule language or guidance to address graded approaches for emergency planning and security programs under the scenarios described above for part 53 applicants and licensees. Please address within the comments any technical, policy, or legal issues that are associated with your suggestions.

Industry Response The NRC should establish clear rule language and guidance on addressing interactions between the graded approaches for emergency planning and security programs. NEI recommends adopting the following approach:

1. For an EPZ sizing analysis, the potential consequences from security events, including those involving the DBT, are adequately bounded if the licensee meets the regulatory requirements to protect against the DBT found in Part 73. This position is reasonable given the following points:

In Parts 50 and 52, and the proposed Part 53, there are numerous NRC requirements concerning the identification and analysis of design basis (unlikely) and beyond-design-basis (very unlikely) accident/event sequences. In accordance with the requirements and guidance found in the Emergency Preparedness for Small Modular Reactors and Other New Technologies Final Rule, these accidents/events are considered when determining the size of a facilitys EPZ.2 It is extremely unlikely that a DBT adversary force could create conditions and consequences markedly more severe than those identified in the analyses of accidents/events used to size the EPZ.

Power reactor security plans have, as a general performance objective, a requirement to provide reasonable assurance against attacks leading to significant 2 For example, see here: https://www.federalregister.gov/d/2023-25163/p-110

Specific Requests for Comments Part 53 Page 28 of 49 core damage and spent fuel sabotage,3 a significant release of radionuclides from any source,4 or unreasonable risk to the public health and safety.5 While acknowledging that the wording for the second and third performance objectives are taken from in-progress rulemakings and could be changed, the final wording will certainly require the same overall levels of protection. It is reasonable to expect that facilities with these types of NRC-approved security plans can prevent a successful attack by a DBT adversary force, including attacks with consequences greater than those resulting from the beyond-design-basis (very unlikely) events examined in an EPZ sizing analysis. These security capabilities include intrusion detection and assessment, delay features, blast protection, an insider mitigation program and, depending on the facility, onsite armed responders and/or armed responders from an offsite location (e.g., law enforcement support).

2. When a licensee does not need to protect against the DBT of radiological sabotage (i.e.,

meets the provisions of §53.860(a)(2)), then the potential consequences from security events up to and including the DBT should be assessed as part of the EPZ sizing analysis.

The determination of the EPZ boundary location would consider the offsite doses resulting from analyzed security events and whether pre-determined, prompt protective measures are necessary.

Also, in this request and elsewhere in the SOC, the NRC references a future revision of Regulatory Guide 1.242, Performance-Based Emergency Preparedness for Small Modular Reactors, Non-Light-Water Reactors, and Non-Power Production or Utilization Facilities. An understanding of the proposed guidance in this revision is necessary to develop a fully informed response to the above NRC question. For this reason, NEI asks that this revision be made available for public comment prior to the Part 53 rule being finalized.

Topic 10 -Part 53, Subpart F - Integrity Assessment Program Requirements NRC Request Decades of operating experience with LWRs suggests that phenomena such as environmentally assisted fatigue and chemical interactions could impact certain SSCs during the life of a commercial nuclear plant. Under the existing regulatory framework, historically, some of these phenomena were not addressed during early licensing reviews but were identified and addressed later when significant safety issues arose (e.g., see numerous generic letters, bulletins, orders, and development and implementation of vessel integrity and materials reliability programs) or a licensee voluntarily pursued renewal of an OL under Part 54. The NRC is proposing to include a new set of programmatic requirements for an Integrity Assessment Program that would ensure these phenomena are addressed early in the life of a commercial nuclear plant licensed under Part 53. The requirements would be provided in § 53.870.

3 From 10 CFR 73.55(b).

4 From the change to 10 CFR 73.55(b) described in the proposed rule, Alternative Physical Security Requirements for Advanced Reactors.

5 From the addition of 10 CFR 73.100(b) described in the proposed rule, Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors.

Specific Requests for Comments Part 53 Page 29 of 49 The NRC is seeking comment on whether the proposed requirements under the Integrity Assessment Program appropriately complement design requirements to address concerns regarding aging, cyclic or transient load limits, and degradation mechanisms related to chemical interactions, operating temperatures, effects of irradiation, and other environmental factors. In addition, the NRC is interested in views on whether, and if so how, degradation mechanisms are or could be addressed in other programs.

Industry Response As noted by NRC, operating experience with LWRs suggests that phenomena have adversely impacted certain SSCs during their licensed life. NRC notes that some of these phenomena were not addressed during early licensing reviews but were identified and addressed later when significant safety issues arose or a licensee voluntarily pursued renewal of an OL under Part 54.

When degradation phenomena were not addressed during the licensing reviews it was a consequence of the phenomena not being recognized or that the potential extent of the degradation was not anticipated. Examples include intergranular stress corrosion cracking in BWR piping and the role played by heat affected zones in sensitizing that material, and the role of material composition, particularly the effect of copper and nickel content in welds in reactor pressure vessels, in exacerbating the effects of neutron irradiation in reducing the fracture toughness of those welds.

While the integrity of the affected plant components was degraded over time by these phenomena, the required inspection, testing, and surveillance programs, including leak detection requirements, identified the degradation and led to repair and mitigation strategies before safety was compromised.

The overall requirements in Part 53 addressing design and inspection provide a sound basis for seeking to identify potential degradation phenomena, ensure that materials used in the SSCs are qualified for used in the operating environments, provide margins against failure, and require assessments of operating experience. Specific examples of the relevant Part 53 requirements include:

§ 53.220 Safety criteria for licensing-basis events other than design-basis accidents, which requires, in part design features and programmatic controls must be provided for each commercial nuclear plant such that identification and analysis of licensing-basis events (LBEs) other than DBAs in accordance with § 53.240 demonstrate (a) Plant SSCs, personnel, and programs provide the necessary capabilities and maintain the necessary reliability to address LBEs other than DBAs in accordance with §§ 53.240 and 53.450(e), and provide measures for defense in depth in accordance with § 53.250;

§ 53.250 Defense in Depth, which requires (a) measures must be taken for each commercial nuclear plant to ensure appropriate defense in depth is provided to compensate for uncertainties in the analysis of the safety criteria such that there is reasonable assurance that the safety criteria in this subpart are met over the life of the plant, and (b) the uncertainties that must be addressed under paragraph (a) of this section include those related to the state of knowledge and modeling

Specific Requests for Comments Part 53 Page 30 of 49 capabilities, the ability of barriers to limit the release of radioactive materials from the facility during LBEs other than DBAs, the reliability and performance of plant SSCs and personnel, and the effectiveness of programmatic controls.

§ 53.440 Design requirements which requires:

(1) Analysis, appropriate test programs, prototype testing, operating experience, or a combination thereof must demonstrate that each design feature required by § 53.400 meets the defined functional design criteria required by §§ 53.410 and 53.420. This demonstration must consider interdependent effects throughout the commercial nuclear plant and the range of conditions under which the design features required by § 53.400 must function throughout the plants lifetime.

(2) The design processes for SR and non-safety-related but safety-significant (NSRSS)

SSCs under this part must include administrative procedures for evaluating operating, design, and construction experience and for considering applicable important industry experiences in the design of those SSCs.

(b) The design features required by § 53.400 must, wherever applicable, be designed using generally accepted consensus codes and standards that have been endorsed or otherwise found acceptable by the U.S. Nuclear Regulatory Commission (NRC).

(c) The materials used for each SR and NSRSS SSC must be qualified for their service conditions over the design life of the SSC. (emphasis added)

(d) Possible degradation mechanisms related to aging, fatigue, chemical interactions, operating temperatures, effects of irradiation, and other environmental factors that may affect the performance of SR and NSRSS SSCs must be evaluated and used to inform the design (emphasis added).

§ 53.450 Analysis requirements, (b) Specific uses of analyses, (5) To identify and assess events that challenge plant control and safety systems whose failure could lead to the uncontrolled release of radioactive material to the environment. These include internal events, such as human errors and equipment failures, and external events identified in accordance with subpart D of this part.

§ 53.710 Maintaining capabilities and availability of structures, systems, and components, (a) Technical specifications must be developed, implemented, and maintained that define conditions or limitations on plant operations that are necessary to ensure that safety-related (SR) SSCs can fulfill the safety functions identified under § 53.230 and support meeting the safety criteria of § 53.210. The technical specifications must describe the following requirements:

o (ii) Surveillance requirements. Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained and that the limiting conditions for operation will be met. (emphasis added) o § 53.715 Maintenance, repair, and inspection programs, (a) A program to control maintenance activities and monitor the performance or condition of SR and NSRSS SSCs must be developed, implemented, and maintained.

Specific Requests for Comments Part 53 Page 31 of 49 (b) Whenever a licensee determines through activities related to maintenance, repair, and inspection of SSCs, the activities under § 53.710, or otherwise that the performance or condition of an SR or NSRSS SSC does not demonstrate compliance with established special treatments or performance goals related to capabilities, availability, or reliability, the licensee must take appropriate corrective action.

(c) Performance and condition monitoring activities and associated goals and preventive maintenance activities must be evaluated at least every 24 months. The evaluations must take into account, where practical, industry-wide operating experience. (emphasis added) Adjustments must be made where necessary to ensure that the objective of preventing failures of SSCs through maintenance is appropriately balanced against the objective of minimizing unavailability of SSCs due to monitoring or preventive maintenance.

§ 53.865 Quality assurance, Each holder of an OL or COL under this part must develop, implement, and maintain a quality assurance program in accordance with appendix B of part 50 of this chapter.

§ 53.880 Inservice inspection and inservice testing, (a) Each holder of an OL or COL under this part must develop, implement, and maintain a program for inservice inspection (ISI) and inservice testing (IST) prior to receiving an OL or COL. The ISI/IST programs must, wherever applicable, be in accordance with generally accepted consensus codes and standards that have been endorsed or otherwise found acceptable by the NRC. The ISI/IST program must include all inspections and tests required by the codes and standards used in the design and be supplemented by risk insights that identify the most important SSCs to plant safety. The types of testing and inspections and their frequency should be informed by risk insights to maintain the reliability and performance of SSCs consistent with the associated design and analyses activities involving those SSCs. Risk insights must also be used to determine when to conduct the inspections and tests (e.g., full power, shutdown, refueling) to minimize risk to the plant workers and the public. The ISI/IST program must be documented in a written manual and managed by qualified personnel reporting to the Plant Manager.

This comprehensive set of requirements provides a robust approach to ensuring the integrity of SSCs during the operational life of the commercial nuclear plant. This is further reinforced, if under 53.440(b) applicants commit to ASME Section XI Division 2 which requires a degradation mechanism assessment and performance monitoring to identify degradation.

The requirement for an Integrity Assessment Program under §53.870 both duplicates these other requirements and increases regulatory burden without an increase in safety.

In Commissioner Caputos comments on SECY-23-0021, Proposed Rule: Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors (ML23199A289) it is noted the draft proposed rule fails to recognize the inherent safety benefits afforded by advanced reactor designs. This is evident in the draft proposed operational requirements for the facility safety program and the integrity assessment program The Commissioner provides additional

Specific Requests for Comments Part 53 Page 32 of 49 thoughts and insights, noting that the need for an integrity assessment program should come from the risk evaluation of a plant. She concludes Staff should remove the proposed integrity assessment program from the regulatory text. Staff should address integrity assessment programs in guidance for cases where an applicants risk evaluation indicates a need for such a program. This is a sensible approach and would eliminate the duplication in requirements in Part 53 regarding SSC integrity and yet provide a means for NRC to provide guidance, as warranted by operating experience, to support individual licensee activities.

Topic 11 - Part 53, Subpart G - Decommissioning NOTE: The NRC is currently pursuing another rulemaking, Regulatory Improvements for Production and Utilization Facilities Transitioning to Decommissioning, which was published as a proposed rule for public comment on March 3, 2022 (87 FR 12254). As these rulemakings progress, the NRC will consider revisions to part 53 to align the two rulemaking efforts. (see p. 100 of the FRN)

NRC Request 1 On March 3, 2022, the NRC published the proposed rule entitled Regulatory Improvements for Production and Utilization Facilities Transitioning to Decommissioning (87 FR 12254). This rulemaking would amend the NRCs current regulations to provide an appropriate regulatory framework for nuclear power reactors transitioning from operations to decommissioning. The rulemaking would address lessons learned from licensees that have completed or are currently in the decommissioning process.

The NRC staff sent a draft final rule to the Commission for its consideration on January 31, 2024, in SECY-24-0011, Final Rule: Regulatory Improvements for Production and Utilization Facilities Transitioning to Decommissioning (3150-AJ59; NRC-2015-0070).

What aspects of this draft final rule, if any, should be incorporated in a Part 53 final rule and why?

Industry Response Generally speaking, the regulations for transitioning from operations to decommissioning set forth in SECY-24-0011 should be incorporated in the Part 53 rulemaking, taking into account the differences in reactor and fuel designs that may require somewhat different language or approaches to account for those differences. The primary objective of a Part 53 rule should be to eliminate the need for licensees to request license amendments and exemptions from regulations to make changes to their operating licenses as the radiological risk at the site decreases during decommissioning. A graded approach to implementing these changes by establishing discrete levels or phases of decommissioning should be established like what is defined in SECY-24-0011. This graded approach would need to reflect the variety of reactor designs and fuel types that may be licensed under Part 53. For example, the condition defined in SECY-24-0011 for an elapsed spent fuel decay period may not be appropriate for some fuels licensed under Part 53, but an analogous condition could be developed for these fuels.

With this and other adaptations a Part 53 rule would provide an efficient framework for transitioning site plans for emergency response, staffing and training, physical security, cyber security, drug and alcohol testing, and financial protection as the site progresses through these decommissioning levels, while maintaining adequate levels of safety and security.

Specific Requests for Comments Part 53 Page 33 of 49 NRC Request 2 Proposed § 53.1060(b) in subpart G would require that, No later than 30 days after the Commission publishes notice in the Federal Register under § 53.1452(a), the licensee must submit a report containing a certification that financial assurance for decommissioning is being provided in an amount specified in the licensee's most recent updated certification, including a copy of the financial instrument obtained to satisfy § 53.1040. This is similar to the current requirement in § 50.75(e)(3) for part 52 COL holders. The NRC is seeking comment on whether commercial nuclear plant COL holders under part 53 should have the same requirement as COL holders under part 52 to demonstrate that they have financial assurance in place no later than 30 days after the Commission issues the notice of intended operation under § 53.1452. Please provide your considerations and rationale for your answer.

Industry Response The proposed §§ 53.1010(b)(1)(ii) and 1060(b) would require a COL holder under Part 53 to update its decommissioning report to certify that it has provided financial assurance for decommissioning in the amount proposed in the application and approved under § 53.1020.

Consistent with § 50.75(e)(3), this update would be required 30-days from issuance of the notice of intended operation required under the proposed § 53.1452(a). In turn, the notice of intended operation must be issued by NRC not less than 180 days before the date scheduled for initial fuel load. This provision provides consistency with the existing requirements applicable to COL holders under Part 52 and would be appropriate for reactors licensed pursuant to Part 53 that are fueled at the reactor site. However, it would be of greater value to adjust this provision to require applicants to submit this update no later than 30-days after initial fuel load, since it will be far more likely that an initial deposit will have been made by that time. A conforming change to Part 52 would achieve the desired consistency.

On a related topic we note that § 50.1060 would require COL holders under Part 53 that are deploying fueled manufactured reactors to submit a decommissioning report updating the DCE and providing a copy of its financial assurance instrument 2 years and 1 year before the date for initiating physical removal of the physical mechanisms to prevent criticality. This requirement may not be practical considering the sequence in which manufactured reactors will be produced by manufacturers, purchased by a COL holder, and then deployed to a site, potentially in a brief time from the date of purchase.

Topic 12 - Part 53, Subpart H - Licenses to construct and operate commercial nuclear plants of identical design at multiple sites NRC Request In addition to including provisions in part 53, subpart H, for referencing ESPs, standard design approvals, and design certifications in applications for commercial nuclear plants, the proposed § 53.1470 provides optional requirements related to the submittal and NRC review of CP, OL, and COL applications to construct and operate commercial nuclear plants of identical design at multiple sites, similar to requirements found in appendix N in both 10 CFR parts 50 and 52. This section would set out the particular requirements and provisions applicable to situations in which applications for CPs and subsequent OLs, or COLs, under this part, are filed by one or more applicants for licenses to construct

Specific Requests for Comments Part 53 Page 34 of 49 and operate nuclear power reactors of identical design ("common design") to be located at multiple sites. Hearings for applications filed under appendix N in both parts 50 and 52 are governed by subpart D of part 2, as would be the case for future part 53 applications under proposed § 53.1470.

Under the proposed requirements in this section, each application is to be treated as a separate application, with the exception of the common design, and so would require separate applications, separate determinations of sufficiency for docketing, separate notices of docketing, and so forth.

Proposed § 53.1470 would also require that each application list all the applications that are to be treated together to ensure that the NRC is clearly informed of the intentions of all applicants. Ordinarily, the NRC would publish in the Federal Register a separate notification of docketing for each application, so that delays in the docketing of one application would not delay the docketing and subsequent technical review of other applications. However, if circumstances allow (e.g., sufficiency review for multiple applications are completed simultaneously), the NRC could publish a single notice of docketing for multiple applications.

With regard to how the NRC would fulfill its obligations under the National Environmental Policy Act of 1969, as amended, the NRC staff would prepare a separate environmental document for each application, but the NRC could conduct joint scoping on environmental issues related to the common design. If the applications reference a standard design certification or the use of a manufactured reactor, then the environmental document would need to incorporate by reference the environmental assessment (EA) prepared for either the design certification or the ML, as applicable. In addition, § 53.1470 would require the ACRS to report on each of the applications, as would be required by provisions in subpart H of part 53. Each ACRS report would be limited to the safety matters which are not relevant to the common design. In addition, the ACRS would need to issue a report on the safety of the common designexcept for those matters relevant to the safety of a referenced design certification or manufactured reactor.

Given this synopsis of how the requirements in proposed § 53.1470 would be implemented as currently written, the NRC is seeking comment on whether there are opportunities to allow added flexibility for applicants under these provisions. This could include consideration of whether applications for which the common design is not completely identical could be evaluated under this provision and, if so, what the process would be for determining the appropriateness of a common review. In addition, the NRC is interested in feedback about the pros and cons of requiring that applications under these proposed provisions be submitted at the same time versus allowing them to be submitted on a staggered basis.

Industry Response NEI supports the NRCs efforts to identify potential opportunities for added flexibility in the 10 CFR 53.1470 process. As the staff notes, section 53.1470 provides optional requirements related to the submittal and NRC review of CP, OL, and COL applications to construct and operate commercial nuclear plants of identical design at multiple sites (similar to requirements found in Appendix N in both 10 CFR parts 50 and 52). We believe that providing flexibility in this process is important given (1) the NRCs and industrys ongoing work related to nth-of-a-kind (NOAK) micro-reactors and other advanced reactors that may be rapidly deployed in high volumes in remote applications (i.e., RHDRA), respectively; and (2) certain provisions in the ADVANCE Act of 2024 aimed at improving the efficiency of the NRCs new reactor licensing process (see, e.g.,

Sections 207, 208, and 505 of the ADVANCE Act).

Specific Requests for Comments Part 53 Page 35 of 49 In our July 31, 2024, RHDRA Proposal Paper, NEI identified the Appendix N process (as currently set forth in Parts 50 and 52) as one potential tool that could be used to reduce schedule and cost risks associated with licensing reviews for each site-specific RHDRA application. Indeed, one of the key concepts proposed in our paper is a rapid efficient and repeatable licensing (ReLic) process that can achieve 6-month deployment times by:

1. completing all, or as much as possible, of the safety and environmental reviews and public engagement processes at one-time, prior to the identification of a specific site, and

2. performing a simplified site license (SSL) application review that focuses on verifying that the site characteristics, which are based on pre-existing generic data as much as possible, conform to the minimum set of site parameters in the envelope established in the one-time up-front reviews.

Given that individual applicants may seek deploy to numerous micro-reactors/RHDRA (which are expected to be fabricated and primarily assembled in a manufacturing facility as opposed to at the deployment site) in a given area or region within a relatively short time frame, the section 54.1470 common design procedures could prove useful in meeting these objectives. In to its September 2024 Preliminary White Paper titled Nth-of-a-Kind Micro-Reactor Licensing and Deployment Considerations, the NRC staff expressed its agreement. Specifically, it noted that NRC staff review of COLs involving multiple fixed sites could be standardized by using 10 CFR Part 52, Appendix N, Standardization of Nuclear Power Plant Designs: Combined Licenses to Construct and Operate Nuclear Power Reactors of Identical Design at Multiple Sites.

While the RHDRA report focused on microreactors, in line with the ADVANCE Act, some of these efficiencies could be gained for any new reactor application that endeavors to resolve generic design issues early through processes such as SDA, DC and ML, and is not limited solely to micro-reactors.

To support ADVANCE Act implementation, proposed 10 CFR 53.1470 should be modified to expressly allow for the possibility of staggered license applications. As with many NRC regulations, the Appendix N process in Parts 50 and 52 (on which 53.1470 is based) was apparently developed with large LWRs in mind, and based on the premise that a utility applicant(s) might seek to license and build several large LWRs on different sites concurrently -

but over an extended, multi-year horizon - using a standard design certification or design-centered review approach. As such, the requirement in proposed section 53.1740 that [e]ach application must list each of the other applications to be treated together under this section should be modified to allow for the staggered or sequential submittal of applications that will rely on a common design, perhaps over a specified time frame (e.g., 12 or 18 months). We do not believe this would inhibit the NRCs administrative handling and technical review of the applications or preclude the efficient conduct of the hearing process.

In this regard, we encourage the NRC staff to review the guidance provided by the Commission in its 2008 Final Policy Statement on the Conduct of New Reactor Licensing Proceedings, 73 Fed.

Reg. 20,963 (Apr. 17, 2008). In that policy statement, the Commission sought to identify procedural measures within the existing Rules of Practice to ensure that particular issues are

Specific Requests for Comments Part 53 Page 36 of 49 considered in the agency proceeding that is the most appropriate forum for resolving them, and to reduce unnecessary burdens for all participants. Among other things, it noted that:

Consideration of generic matters common to several applications may be possible in several contexts, including, for example, the NRC staffs review of a corporate or operational program (e.g., quality assurance, security) that is common to several applications. Another example is staff review of design acceptance criteria (DAC), a special type of ITAAC that are used to verify the resolution of design issues for which completed design information was not provided in a design certification application.

If contentions on such a program or DAC are admitted with respect to more than one application, consolidation of such contentions before a single licensing board may result in more efficient decision making, as well as conserving the parties resources.

While applicants should seek to submit their common design applications simultaneously, Subpart D to 10 CFR Part 2 nonetheless affords the licensing board discretion to consolidate applications filed close in time, if this will be more efficient and otherwise provide for a fair hearing.

An applicant(s) may request variances or exemptions from a design certification, but such variances or exemptions should be common to all applications for which the applicant(s) is seeking a common design review.

The Commission does not require, but rather encourages, applicants to standardize the balance of their plants to the extent practicable.

An applicant requesting treatment of its application under the design-centered approach may seek to submit separate portions of the application at different times, pursuant to 10 CFR 2.101(a)(5) or an exemption from 10 CFR 2.101. Under such circumstances, the Commission would issue a Notice of Hearing for the portion of the application to be reviewed under the design-centered approach, and a second notice limited to the portion of the application not treated under the design-centered review approach upon submission of the complete application. Such a procedure would not affect any prospective intervenors substantive rights to seek intervention on issues that are material to the Commissions decision on each individual application.

The Commission is willing to consider other methods of managing proceedings involving consideration of information common to several applications. For example, the Commission does not intend to foreclose the Chief Judge of the Atomic Safety and Licensing Board Panel from designating a licensing board to preside over common portions of applications on the motion of the applicants, even if separate proceedings have already been convened on one or more of the applications involved. In such a case, however, the applicants should jointly identify the common portions of their respective applications when requesting the Chief Judge to take such action.

Revisions of specific applications during the review process could result in formerly common issues being referred to the licensing board presiding over a specific

Specific Requests for Comments Part 53 Page 37 of 49 portion of one or more applications, and would be resolved in the normal course of adjudication.

In short, the Commissions 2008 Policy Statement suggests that it intends for the common design in Appendix N to 10 CFR Part 52 (and related heading procedures in Subpart D of 10 CFR Part 2) - which the staff now proposes to include in 10 CFR 53.1470 - to be applied in a flexible manner. That flexibility should accommodate, to the extent practicable, the possibility of sequential or staggered license applications, as well as potential variances or exemptions from design certifications (or other NRC generic approvals) and revisions to specific applications.

Topic 13 - Part 53, Subparts H and I - Probabilistic Risk Assessment Information NRC Request Proposed § 53.1239(a)(18) in subpart H and the related references to this proposed requirement for the holders of OLs and COLs would require a description of the PRA required by § 53.450(a), and its results to be included in FSARs. However, guidance documents may further clarify the division of PRA-related information needed to be in the FSAR, in other possible licensing basis documents, and controlled as plant records subject to inspections and audits. For example, a possible approach could be to include a summary of the PRA results in the FSAR and control that information under § 53.1545 and create a separate document related to the broader PRA analyses and related processes as a program document under § 53.1560. The program document would provide more detail than the summaries in the FSAR but still be a much-condensed source of information in comparison to the documentation of the PRA.

This possible approach would reflect the role of the PRA in the licensing process under part 53 and in maintaining margins to the safety and evaluation criteria in subparts B and C but may allow a more appropriate evaluation process to address the particulars and complexities of the PRA-related documents.

The NRC is seeking comment on the appropriate placement of PRA-related information among various licensing basis documents and plant records. In addition to the placement of PRA-related information, the NRC is seeking comment on the appropriate control of that information and on the routine submittal of updates to the NRC. Please provide your considerations and rationale for your answer.

Industry Response This response is informed by the 2007 updates to Part 52 regarding PRA maintenance. In that rulemaking, the Commission decided that PRAs and upgrades are not required to be submitted to the NRC, but instead should be maintained by the licensee for NRC inspection. As discussed in existing guidance, licensees would maintain a PRA configuration control procedure, in accordance with the non-LWR PRA Standard which would drive an emergent update for significant changes that would impact risk-informed decision-making. Plant changes would continue to be tracked through the 50.59 process and NEI 22-05 for a Technology-Inclusive Risk-Informed Change Evaluation provides guidance on when plant changes would require a license amendment request prior to implementation.

Specific Requests for Comments Part 53 Page 38 of 49 The guidance in NEI 21-07 as endorsed by RG 1.253 provides sufficient detail on the level of information to be captured in the SAR for applicants following LMP. RG 1.253 should be updated to endorse RG 1.253 for Part 53 and NEI is happy to discuss any changes to NEI 21-07 that would facilitate that endorsement. Similar guidance could inform other TI-RIPB methodologies for licensing basis development, and NEI is willing to either update NEI 21-07 or develop guidance documents in concert with industry for NRC endorsement of the appropriate PRA information to be captured in the DAR for various licensing approaches.

The language in 53.1545 for PRA updates seem appropriate. By pointing to the 53.450 requirements, NEI interprets this requirement to mean that when the PRA is updated every 5 years in accordance with 53.450(c), that updated information needs to be incorporated into the SAR update which is required every 2 years. NEIs interpretation is that this requirement would not drive a PRA update every 2 years as that would be unnecessarily burdensome and out of line with existing precedent.

It would NOT be appropriate to have a separate document related to the broader PRA analyses and related processes as a program document under § 53.1560. Such a document would be informed by:

1. PRA changes which are adequately covered under the maintenance requirement in 53.450(c).

2. Design Changes and Changes to the Method of Evaluation that are sufficiently covered by 53.1550. Note that this requirement should be updated to align with NEI 22-05 once updated for NRC endorsement with consideration of flexibility to allow different licensing approaches.

3. Changes to Special Treatments which may impact DID adequacy which is already covered through 53.1560 for updates to programs.

Topic 14 - Part 53, Subparts H and I - Changes to Manufacturing License NRC Request Proposed § 53.1530 would not allow the holder of an ML or the holder of a COL using a manufactured reactor to make changes to the design of the manufactured reactor without requesting a license amendment from the NRC. The proposed requirements do not include a specific mention of the manufacturing processes for which the NRC could possibly provide finality under proposed § 53.1288.

The NRC is seeking comment on the appropriate change control provisions for MLs, including whether criteria could be developed to determine when a license amendment request would not be required and whether those criteria should address changes in manufacturing processes as well as changes in the design. Please provide your considerations and rationale for your recommendation.

Industry Response Guidance on the licensing requirements of design changes should be developed in collaboration with industry, considering whether the design change results in a safety significant impact. Cask manufacturing licenses and NEI 12-04 might provide a worthwhile framework.

Specific Requests for Comments Part 53 Page 39 of 49 Topic 15 - Financial Qualifications NRC Request Utility new reactor applicants are exempt under § 50.33(f) from financial qualification reviews because they are generically presumed to be financially qualified for operations. In contrast, merchant power plant new reactor applicants are required under § 50.33(f)(2) to submit information that demonstrates they possess or have reasonable assurance of obtaining the funds necessary to cover estimated construction and operating costs for the period of the license. A merchant power plant new reactor applicant is a non-rate-regulated entity (e.g., a nonutility) that engages in the business of production, manufacturing, generating, buying, aggregating, marketing, or brokering electricity for sale at wholesale or for retail sale to the public. Over the past decade, the agency has heard some concerns about the challenges that merchant power plant applicants face in demonstrating compliance with the current financial qualification requirements.

NRC Question: Does this standard continue to pose challenges for merchant power plant applicants? If so, please provide a detailed explanation of these challenges.

Industry Response The current reasonable assurance financial qualifications (FQ) standard imposes unnecessary requirements for burdensome submittals of information and a detailed FQ review that is not necessary to protect public health and safety. This FQ standard also creates unreasonable impediments to the development of new reactor projects, including advanced reactors.

There is no direct correlation between financial qualifications and safety performance, and even if there is an indirect impact, there is no evidence that the safety performance of non-electric utility operators is affected by the NRC staffs FQ reviews. Moreover, if there were a correlation of safety performance with financial resources, one might expect that regulated utilities without the financial pressures of unregulated utilities would have better safety performance. However, the safety performance of non-utility reactors compares favorably with regulated reactors over the last two decades. Moreover, all evidence suggests that the NRCs safety requirements and related review and oversight processes are the most effective tools for assuring safety performance.

There is no evidence that NRC Staff FQ Reviews Are Necessary to Assure Safety Performance.

The NRC staff has long understood that there is no direct correlation between licensees pre-licensing FQ reviews and later safe construction or operating performance. See SECY-79-299, Generic Issue of Financial Qualifications: Licensing of Production and Utilization Facilities, dated April 27, 1979.6 When nuclear industry restructuring unfolded in the early 2000s, the NRCs regulations required detailed FQ reviews by the NRC staff in order to approve transfers of commercial nuclear power reactors from traditional electric utility owners and operators to non-electric utility power companies (merchant owners and operators). As set out in the Abstract for a 2014 study of industry experience following the wave of deregulation, some policy-makers raised concerns that these [merchant] corporations would ignore safety, while 6 Cited by Financial Qualifications for Reactor Licensing Rulemaking, Regulatory Basis Document, at 11 (October 2016)

(ML5322A185).

Specific Requests for Comments Part 53 Page 40 of 49 others claimed that these transfers would bring improved reactor management, with positive effects on safety. American Economic Journal: Economic Policy, 6(3), 178-206, Corporate Incentives and Nuclear Safety, by Catherine Hausman (August 2014) (Hausman). Hausmans findings confirm that transferring a reactor to merchant plant status does not have an adverse impact on safety performance:

Using data on a variety of safety measures and a difference-in-difference estimation strategy, I find no evidence that safety deteriorated; for some measures, it even improved following divestiture. Moreover, for given levels of generation, safety substantially improved. Ownership transfers led to the alignment of private incentives to increase operating efficiency, and these gains do not appear to have come at the cost of public safety.

Hausman, at 1.

Merchant plants have certainly faced financial challenges in the years since the Hausman analysis was published in 2014, and several merchant plants have faced potential and actual shutdowns due to financial issues in the past ten years. However, there is no evidence to suggest that these issues have caused safety performance to suffer. To the contrary, the NRCs and industrys overall experience in the reactor oversight process, for example, does not show that merchant plants have performed differently than regulated plants.

This background suggests that the NRC could eliminate the FQ review requirements altogether.

However, in the prior rulemaking to change the FQ standards in Part 50 and 52, the NRC staff determined that a more limited FQ review using the appears to be financially qualified standard should be retained. See SECY-18-0026, Proposed Rule: Financial Qualifications Requirements for Reactor Licensing (Feb 26, 2018) (ML17172A656).

New Level of FQ Review. Under the revised FQ standard, as previously proposed by the NRC staff, the voluminous information requirements in 10 CFR Part 50, Appendix C would be deleted.

Instead, an applicant would need to submit a plan showing that it had financial capacity to obtain funding for construction. For an operating license, the applicant would need to provide estimates of the total annual operating costs for each or the first 5 years of operation and an Applicant Financial Capacity Plan describing how the applicant intends to cover the estimated operating costs, including documentation of sources of funds to cover each of the first 5 years of operation. Presumably, the NRCs FQ requirements for operations would be similar to the 5-year pro forma reviews that the NRC staff historically has used for its license transfer FQ reviews.

While such an FQ review arguably is unnecessary, the NRC staff is presumably concerned that an outlier applicant for an operating license or applicant for the transfer of an operating license could present itself. By retaining an FQ review under the new standard, the NRC staff would be able to screen an applicants capability and ensure that each such applicant appears to have capabilities similar to existing merchant plant owners and operators.

NRC Safety Review Process and Programs. FQ issues that might arise during the construction or operation of any reactor are best identified and addressed in connection with NRCs existing

Specific Requests for Comments Part 53 Page 41 of 49 safety review, oversight, and inspection processes and programs. As described in the NRC staffs Regulatory Basis for the 2018 Proposed Rulle, these include:

NRC Licensing Review Process

1. Construction Reactor Oversight Process

2. Reactor Oversight Process

3. Resident Inspectors Program

4. Reactor Operating Experience Program

5. Vendor Inspection Program

6. Quality Assurance Inspection Program

7. NRC Oversight of Non-power Production or Utilization Facilities A reduced level of FQ review would not compromise public health and safety, because the NRC maintains a number of oversight programs and processes that directly ensure safe plant construction and operation. These programs are well established and provide reliable direct mechanisms for identifying potential safety issues during power reactor construction and operation. With these programs, the NRC can readily identify any degradation in licensee performance during construction and operation, independent of root causes and independent of status as a merchant plant or rate-based utility. Thus, the more detailed review required in the current FQ requirements is not necessary to protect public health and safety.

NRC Question: Should part 53 have the same financial qualification requirements as parts 50 and 52?

Why or why not?

Industry Response Yes, Part 53 should have the same financial requirements as parts 50 and 52, which should be based in each of these parts on the appears to be financially qualified standard in part 70 (Section 70.23(a)(5)). Utility applicants with cost of service recovery should remain generically presumed to be financially qualified if they are applicants for a license under Part 50, 52 or 53.

For nonelectric utility applicants, a detailed review of financial qualifications is unnecessary.

However, to provide an initial screening review to avoid projects by applicants clearly incapable of being successful, it may be appropriate that the NRC staff conduct a limited inquiry regarding financial qualifications under the appears to be financially qualified standard. If necessary to assure that projects do not begin safety-related construction activities without adequate funding, the NRC staff could impose appropriate license conditions or verify an applicants documentation demonstrating commitments of financing equal to more than 50 percent of the construction cost estimate to assure that it appears that adequate funding will be available. As the NRC staff explained in the proposed rule attached to SECY-18-0026 as Enclosure 1:

The proposed rule would provide a new process for FQ reviews for the various classes of applicants while maintaining public health and safety. The purpose of the staffs FQ review is to ensure that an applicant has the financial capacity to obtain funding, not to ensure that the project is completed. The staff contemplated various levels of funding that would indicate that an applicant would be able to obtain the remaining funds needed. The NRC staff has determined that an applicant with commitments for greater

Specific Requests for Comments Part 53 Page 42 of 49 than 50 percent funding for proposed licensed activities has made a reasonable and sufficient demonstration of financial capacitysuch an applicant would not be subject to license conditions for future verification.... Under the proposed rule, initial applicants with 50 percent or less of the necessary funding at the time of application would be subject to license conditions for future verification that sufficient funding is available prior to the start of licensed activities.

The staff further noted that any license conditions would be sufficient and specific to allow the NRC staff to conduct a ministerial review to verify that the licensee's financing plan is executed and funding is obtained prior to the start of licensed activities.

NRC Question: Are there categories of merchant new reactor applicants for which a part 70 appears to be financially qualified standard would be more appropriate? If so, please explain what types of applicants should be able to use the part 70 financial qualification standard and what distinguishes these applicants from ones that should not be able to use this standard.

Industry Response There is no apparent basis for distinguishing between any traditional nonutility applicants for large commercial reactors, i.e., these applicants all have the same basic characteristics. Each of these applicants lacks cost-of-service treatment and plans to sell power into the market under appropriate power sales arrangements (short or long-term power sales contracts or real-time market sales of electricity in a competitive market). There is no reason to believe that any nonutility advanced reactor applicant would be any different. With an adoption of the appears to be financially qualified standard, however, the NRC staff would have the flexibility to address any unique circumstances involving a nonelectric utility applicant that might arise.

NRC Question: If a part 70 financial qualification standard were to apply to a category of merchant new reactor applicants, should it also apply to pre-construction license transfer applications for these reactors? Why or why not?

Industry Response Yes, the appears to be financially qualified standard is appropriately applied to both preconstruction and operating license transfers. To the extent the NRC staff finds it necessary to impose preconstruction or pre-operation conditions, those conditions should be assessed and either revised or reimposed in connection with any license transfer. Presumably, this would also include a review of financial projections (cost and revenue) for the initial five years of operation.

In connection with any operating license for an operating facility, the NRC staff would also review five-year financial projections in connection with license transfers.

NRC Question: Is there another standard the agency should consider for financial qualification of merchant new reactor applicants? Commenters are encouraged to provide specific suggestions and the basis for those suggestions.

Industry Response For the reasons discussed above and in the NRCs Regulatory Basis document and Proposed Rule in SECY-18-0026, the Part 70 appears to be financially qualified standard offers sufficient

Specific Requests for Comments Part 53 Page 43 of 49 flexibility for a limited NRC staff review of the financial qualifications without imposing an undue burden upon applicants and without creating unacceptable impediments to the development of new reactor projects.

Topic 16 - Recent Legislation NRC Request On July 9, 2024, the President signed into law the Accelerating Deployment of Versatile, Advanced Nuclear for Clean Energy Act of 2024, also referred to as the ADVANCE Act. Section 203, Licensing Considerations Relating to Use of Nuclear Energy for Nonelectric Applications, and Section 208, Regulatory Requirements for Micro-Reactors, of the ADVANCE Act specifically mention the technology-inclusive regulatory framework to be established under section 103(a)(4) of NEIMA as a potential vehicle to be considered for the report to Congress required under section 203 and a potential vehicle to implement strategies and guidance for the licensing and regulation of micro-reactors required under section 208. This proposed rulemaking is, in part, how the NRC is implementing section 103(a)(4) of NEIMA.

The NRC is seeking comment on how part 53 could be revised to better enable its potential use to implement the ADVANCE Act. Specifically, Section 208 of the ADVANCE Act requires the NRC to develop and implement risk-informed and performance-based strategies and guidance in several areas for the licensing and regulation of micro-reactors, including with respect to licensing mobile deployment. The ADVANCE Act requires the NRC to consider the unique characteristics of microreactors, including physical size, design simplicity, and source term; opportunities to incorporate specific improvements related to streamlining the review process; and other policy and licensing issues. With regard to implementation, the ADVANCE Act provides the NRC with three options. The NRC may implement the developed strategies and guidance, as appropriate, via (1) the existing regulatory framework, (2) the Part 53 rulemaking, or (3) a pending or new rulemaking. Given the language included in Section 208, the NRC is seeking comment on how part 53 could be revised to better address the ADVANCE Acts requirements related to strategies and guidance for micro-reactors.

Industry Response NEI appreciates this request for public comment to address the ADVANCE Act. Many of the items highlighted in the ADVANCE Act are addressed to some extent under Part 53 including:

Section 203 Flexible operation including load following The use of advanced reactors for non-electric applications A definition of construction that allows building the energy island without an NRC license, which supports colocation with industrial facilities Section 208 Flexibility in staffing and operations, safeguards and security and emergency preparedness.

SRM-SECY-20-0045 to the extent it was addressed in RG 4.7 R4

Specific Requests for Comments Part 53 Page 44 of 49 The ADVANCE Act should be implemented for Parts 50, 52 AND 53 since applicants for microreactors intend to pursue both approaches. There are several regulatory issues that NEI believes are not yet adequately addressed in Part 53. Part 53 needs additional changes to address:

Section 203 Guidance for various non-electric applications such as chemical production, water desalination, industrial heat, energy storage, isotope production and particularly district heating.

Section 208 develop risk-informed and performance-based strategies and guidance to license and regulate micro-reactors including strategies and guidance for. risk analysis methods, including alternatives to probabilistic risk assessments; the transportation of fueled micro-reactors; and the related licensing mobile deployment.

siting, including in relation to(i) the population density criterion limit described in the policy issue paper on population-related siting considerations for advanced reactors dated May 8, 2020, and numbered SECY-20-0045; (ii) licensing mobile deployment; and (iii) environmental reviews; other relevant considerations discussed in the policy issue paper on policy and licensing considerations related to micro-reactors dated October 6, 2020, and numbered SECY-20-0093.

Section 203 While many of the issues identified under section 203 of the ADVANCE act are addressed by more flexibility in part 53, some regulatory uncertainty remains about what requirements might be placed on adjacent facilities (radiation protection, security, including access authorization and cybersecurity, etc.). The most problematic issue with Part 53 with respect to section 203 is the inclusion of population-related siting criteria which duplicates the Part 100 language. Since Part 53 intends to be performance-based and should consider deployment models as directed by the ADVANCE Act, the population-related siting criteria should not be prescriptive and exclusionary.

Proposed section 53.530 would establish requirements for population-related considerations and maintain requirements and definitions similar to those currently in part 100 for an exclusion area, low population zone, and population center distance. Although the NRC recognizes that some applicants may propose to essentially collapse the exclusion area and low population zone to the site boundary, as permitted by the NRCs current EP regulations, it nonetheless proposes to maintain the NRCs long-standing preference for siting reactors in areas of low population density using the current language from part 100 in proposed § 53.530(c). As a result, for purposes of Part 53, the NRC plans to retain the definition of population center distance in 10 CFR 100.3 - i.e., the distance from the reactor to the nearest boundary of a densely populated center containing more than about 25,000 residents. It also

Specific Requests for Comments Part 53 Page 45 of 49 proposes to retain the related requirement in 10 CFR 100.21(b) that [t]he population center distance must be at least one and one-third times the distance from the reactor to the outer boundary of the low population zone. Significantly, the proposed rule provides no technical or regulatory basis for the NRCs decision to retain these Part 100 siting requirements in Part 53.

This issue is of critical importance to the industry given its adverse implications for the siting of advanced reactors (SMRs, non-LWRs, micro-reactors) closer to, or even within, larger population centers to support various emerging industrial and commercial applications beyond wholesale electricity. These use cases include, for example, providing electricity for data centers, clean tech factories, and hydrogen production facilities; supplying industrial process heat and steam; and replacing fossil fuels or retired coal-fired generation facilities. Given the likely need to deploy advanced reactors at sites that are located closer to large population centers than traditional large LWRs, a more flexible, risk-informed approach to siting based on the specific characteristics advanced reactors is warranted.

NEI believes that simply importing Part 100s prescriptive population siting requirements into Part 53 is fundamentally inconsistent with the purpose of Part 53, as mandated by Congress in the Nuclear Energy Innovation and Modernization Act of 2019 (NEIMA). Specifically, Part 53 is intended to provide a risk-informed, performance-based, and technology-inclusive regulatory framework for new commercial nuclear plants. As discussed in other portions of our comments, the ADVANCE Act of 2024 further reinforces that purpose. Moreover, notwithstanding these congressional directives, the NRC has been developing and implementing various technology-inclusive, risk-informed, and performance-based licensing approaches for advanced reactors, as evidenced by its 2023 final EP rule, 2024 proposed Part 73 physical security rule, and endorsement of the Licensing Modernization Project (LMP) methodology. The NRCs approach to population siting in Part 53 should not be an outlier or exception to the agencys efforts to provide risk-informed and performance-based approaches for licensing and regulating new nuclear power plants. To the contrary, the NRCs approach should provide new nuclear plant applicants with the flexibility to justify siting their plants closer to, or within, densely populated centers (including those with more than 25,000 residents) based on their specific reactors designs and safety attributes.

In NEIs view, the NRCs 2023 rulemaking establishing alternative EPZ sizing requirements for SMRs and other new technologies provides the necessary technical and regulatory bases for enabling the siting of new advanced reactors closer to population centers, without increasing risk to the public (including societal risk) or reducing defense-in-depth, which are considerations historically cited by the NRC as the primary function of its prescriptive population center distance requirements. Indeed, in SECY-16-0012, Accident Source Terms and Siting for Small Modular Reactors and Non-Light Water Reactors (Feb. 2016), the staff noted that [t]he dose criteria which would allow for a smaller LPZ would also potentially allow the reactor to be considered for a location at a distance that is relatively close to a population center with at least 25,000 people. In SECY-20-0045, Population Related Siting Considerations for Advanced Reactors (May 2020), the staff noted that its efforts to address population-related siting considerations are an important part of the integrated approach to help inform the design and siting processes for advanced reactors. It also recognized that NEIMA requires the NRC to develop and implement risk-informed, performance-based methods to resolve policy issues

Specific Requests for Comments Part 53 Page 46 of 49 such as siting considerations that may unnecessarily restrict the development of commercial advanced nuclear reactors. Notably, in voting on SECY-20-0045, Commissioner (now Chairman)

Wright emphasized that [t]his technology-inclusive, risk-informed approach moves away from deterministic siting criteria, consistent with the NRCs approach to establish emergency preparedness requirements for small modular reactors.

Accordingly, NEI recommends that the NRC not use the current language from part 100 in proposed section 53.530. Instead, we recommend that NRC remove 53.530(c) and rely on updated guidance (RG 4.7).

Performance-based guidance on potentially collocating the low population zone and population center distance with a site boundary EPZ should be developed for the population density criterion to potentially allow siting of low-consequence reactors close to or within densely populated centers containing more than about 25,000 residents. NEI intends to further justify this position in an upcoming White Paper to the NRC.

Section 208 NEI submitted NEI Proposal Paper Regulations of Rapid High-Volume Deployable Reactors in Remote Applications (RHDRA) and Other Advanced Reactors (ML24213A337) which provides a roadmap for meeting many of the issues pertaining to section 208 of the ADVANCE Act. That report should be considered when updating Part 53 to be inclusive of microreactor technologies.

The following items in particular need to be addressed in Part 53.

Alternatives to PRA Section 208, directs the NRC to develop risk-informed and performance-based strategies and guidance to license and regulate micro-reactors including strategies and guidance for. risk analysis methods, including alternatives to probabilistic risk assessments; This is in direct conflict with the language in 10 CFR 53.450(a) which requires a PRA. Directionally the ADVANCE Act is aligned with the Commission directive in SRM-SECY-23-0021: The staff should not apply consensus probabilistic risk assessment (PRA) standards as a strict checklist of requirements for Part 53 PRA acceptability determinations. Framework B was intended to provide such a pathway for risk insights to be considered without a PRA, and the SRM proposed an option c for an option to create a less prescriptive regulation where methods of compliance, similar to Framework B, could be located in guidance. NEI suggests that to meet the intent of the ADVANCE Act, Part 53 should be revised to require a risk evaluation, but that the risk evaluation may be qualitative and is not required to be a PRA. Language for Risk Evaluations in line with Commissioner Caputos vote would allow the approaches originally envisioned for Part 53, would align with option c provided to the staff for Framework B and meet the intent of the ADVANCE Act. Language elsewhere in the proposed rule would have to be changed to remove other than DBAs from the phrase LBEs other than DBAs. As discussed in the detailed comments, LBEs without the addition of other than DBAs allows the methodologies initially envisioned but would also allow a more traditional approach.

Transportation of Microreactors

Specific Requests for Comments Part 53 Page 47 of 49 While transportation of microreactors will primarily be governed by 10 CFR 71, the part 53 rulemaking missed an opportunity to address transportation of microreactors. By addressing fuel loading with a pathway for the reactor to be fueled but not in operation Part 53 begins with a pathway for the transport of a fueled microreactor. That provides a much better starting point than the existing Part 50 or 52 pathways. From SECY-24-0008: A fundamental subject of this paper is whether the commissions historical position that operation begins with the loading of fuel into the reactor is applicable to factory-fabricated modules.8 As an outgrowth of this position, each factory-fabricated micro-reactor module that is fueled in the factory would be required to have a facility operating license or a combined license, regardless of whether it is operated for testing at the factory. This position also means that a factory-fabricated module would be considered to be in operation when loaded with fuel. The NRCs current regulatory framework does not provide for authorizing transportation of utilization facilities that are in operation, meaning that a factory-fabricated module could not be transported to a deployment site when loaded with fuel without changes to the current regulatory framework. In this paper, the NRC staff proposes that a factory-fabricated module with features to preclude criticality would not be in operation when loaded with fuel.

NEI agrees that a fueled microreactor, with feature to preclude criticality, should not be considered in operation. NEI provides more detail in our response to the manufacturing license questions, but more could be done to increase regulatory certainty for the transport of fueled microreactors and the licensing of mobile microreactors. Part 53 fails to be truly technology-inclusive without being inclusive of these innovative technologies.

Siting Issues The ADVANCE Act explicitly calls out the population density criterion limit described in the policy issue paper on population-related siting considerations for advanced reactors dated May 8, 2020, and numbered SECY-20-0045; with Part 53 as one means of implementing the change. See the Section 203 response to this question for the proposed changes to address this clause of the ADVANCE Act.

Other microreactor considerations There are several other microreactor licensing considerations from SECY-20-0093 that could be addressed by a more technology-inclusive Part 53. One of the more important items to address is a pathway to testing of a fueled microreactor at the manufacturing facility. As described in the response to the manufacturing license questions, NEI believes that a Part 104 license, using methodologies similar to NUREG-1537, would be appropriate for such an application. Other issues remain, but the performance-based requirements of Part 53 are better-suited to potential microreactor deployments so it is essential that as many of the ADVANCE Act issues as possible be addressed as part of the Part 53 Rulemaking.

Decommissioning Part 53 maintains the same decommissioning regulations for microreactors as large LWRs. The regulation does not adequately address the microreactor paradigm where the reactor would be transported back to the microreactor manufacturer for decommissioning or refurbishment. The

Specific Requests for Comments Part 53 Page 48 of 49 decommissioning funding requirements in Subpart G still seem to place the onus of the funding on the holder of an OL or COL. The regulations should consider that two decommissioning funding plans could be submitted for a reactor siteone by the microreactor manufacturer for removal and eventual decommissioning of the reactor and one by the OL or COL holder for removal of radioactive material that remains onsite after removal of the microreactor. In addition, the decommissioning regulations should allow for their use, if chosen, to remove the microreactor from the site of operations to a facility for decommissioning. This recommendation should be accommodated as a change to Part 53 and should not delay the issuance of a final rule on the transition to decommissioning currently before the Commission in SECY-24-0011.

Other There is potential under Part 53 to create a General License process for certain reactors, including microreactors. As discussed in NEIs rapid high-volume deployment of microreactors (RHDRA) report (ML24213A337), the Atomic Energy Act may need to be amended to authorize a general license approach for reactors and would require the development of appropriate NRC policy, regulations and guidance. However, Appendix 5 of that report proposes a pathway for minimizing the scope and level of information required in a site-specific license application, and Part 53 should facilitate such an approach. Implementing such a pathway will require further engagement between NRC and industry.

53.1449(c)(3) requires 225-day notification if ITAAC have not yet been closed for COL.

53.1452(a) requires notification of plans to operate a minimum of 270 days before start of operation. As NEI and industry put forth in the RHDRA paper, future microreactors of a standard approved design will be looking to operate within 180 days of identification of a need for power.

These regulations and timeframes do not support the needs for microreactors and appear to be focused on proposed timelines for large LWRs. It should be considered whether these timelines should be graded based on reactor being licensed.

While not called out explicitly in the ADVANCE Act, Part 53, section 53.1118 should be updated to reflect Section 301 of the ADVANCE Act removing certain limitations on foreign ownership of some types of licensed facilities. Furthermore, the NRC should perform a systematic and aggressive search for potential changes, in requirements, policy and guidance, to reduce unnecessary regulatory burden that have not yet been fully considered. A more thorough consideration of changes to reduce regulatory burden is consistent with the intent and direction of the ADVANCE Act.

Topic 17 - Regulatory Analysis NRC Request The NRC requests public comment on the draft regulatory analysis, which is available as indicated in the Availability of Documents section of this document.

Industry Response We request that NRC perform a line item cost/benefit analyses by assessing each change (as compared to the current Parts 50 and 52) individually to avoid inadvertently including

Specific Requests for Comments Part 53 Page 49 of 49 burdensome requirements through cost/benefit analyses of the aggregated changes. The Integrity Assessment Program, any changes to ALARA requirements, the 2 physical independent mechanism requirement (as opposed to traditional double contingency principle, NEI acknowledges the net benefit for a change to the definition of in operation), extension of SR requirements to NSRSS SSCs, and other changes seem unlikely to provide net benefit for NRC or industry.

Specific Requests for Comments Part 26 Page 1 of 10 Topic 1 - Part 26 - Fitness for Duty Program NRC Request The proposed rule under § 26.603(c) would enable a licensee or other entity to implement an FFD program under proposed § 26.604, FFD program requirements for facilities that satisfy the § 26.603(c) criterion, if the licensee or other entity performs a site-specific analysis to demonstrate that the facility and its operation satisfy the criterion in § 53.860(a)(2).

Should the NRC consider replacing its proposed § 26.603(c) criterion referencing § 53.860(a)(2) with an alternative requirement that if the commercial nuclear plant is of the class described in § 53.800, Facility licensees for self-reliant-mitigation facilities, and either § 53.800(a)(1) or (2) is satisfied, then drug and alcohol testing would not be required? This proposal would align the § 26.603(c) criterion with that proposed in the NRC-licensed operator regulatory framework of part 53. Please provide your considerations and rationale for your recommendation.

Should the NRC also consider making a conforming change to the proposed § 73.120 criterion used for the AA program? Please provide your considerations and rationale for your recommendation.

Comment The foundation of the Fitness for Duty program is predicated on behaviors that are observed, documented and reported with the goal of mitigating anomalous behaviors and thereby protecting the health and safety of the public and onsite staff. An intact behavior observation program provides reasonable protections regarding aberrant behaviors that could contribute, directly or indirectly to a deleterious event.

26.604 should still require a BOP for plants meeting the criteria of 53.860(a)(2).

A new section should allow flexibility for plants meeting the criteria of § 53.800(a)(1) or (2) and would remove the requirement for random testing as well as allow scaling back other FFD requirements. With behavior observation firmly established; other tenets of the Fitness for Duty mandates could be stood down or minimized. There remains a concern with, even with reduced risk to the public in cases of self-reliant mitigation facilities. The insider threat risk remains in all circumstances as the program element that is most effective in mitigating insider threats.

Additionally, even when considering reduced risk to the public, the concern with onsite safety and reliability remains. Protecting on-site staff and equipment is a major concern. Personnel who are experiencing mental health conditions, substance abuse disorders, and other comorbid conditions should not be operating safety or sensitive material and should, through the behavior observation program, be subjected to For Cause testing and subsequent referral to an evaluator assessment and potential treatment. Random testing and other FFD elements could be scaled back with conditions of testing that align with For Cause or Post Event testing. An effective behavior observation program allows for, in any state of reduced risk, the ability to respond and interdict in the best interest of the individual worker and the installation.

Specific Requests for Comments Part 26 Page 2 of 10 Suggested changes These changes would more appropriately scale the FFD requirements for facilities will maintain a BOP that benefits all licensees.

Topic 2 - Part 26 - Technology-Inclusive Approaches to Fatigue Management Requirements Applicable to Unit Outages NRC Request In establishing the outage minimum days off requirement of § 26.205(d)(4), the NRCs objective was to ensure that individuals performing the duties described in § 26.4(a)(1) through (a)(4) have sufficient periodic long-duration breaks to prevent cumulative fatigue from degrading their ability to safely and competently perform their duties. In addition to the science of fatigue management, the NRC considered several factors in establishing the existing requirements. These additional factors were practical and safety considerations associated with the management of refueling outages for large LWRs, including the following: (1) the typical duration and frequency of outages; (2) the availability of contract personnel to perform the work; (3) the risk presented by the outage work while the reactor is shut down; and (4) the controls applied to the work that may limit the potential for latent errors to challenge reactor safety when the reactor is returned to power. The details of such considerations may differ for new reactor technologies or designs. Such considerations may not be relevant for some reactor designs (e.g., reactors capable of on-line refueling) and there may be additional, more pertinent factors to consider for other designs.

The NRC is seeking stakeholder input on whether alternative fatigue management requirements applicable to outages should be adopted to support technology-inclusive approaches that would be appropriate to support the licensing and regulation of future commercial nuclear plants. Please provide your considerations and rationale for your recommendation.

Comment The proposed fatigue management approach is appropriate as-is in consideration of the "science of fatigue management" and there would be little difference to be earned by considering alternative requirements.

The one change industry would propose is generic approval of waivers of requirements for natural phenomena instead of requiring plants to request on an ad hoc basis if in hurricane prone areas.

Topic 3 - Part 26 - Draft Regulatory Guidance Approach for Fatigue Management NRC Request In support of this proposed rule, the NRC has issued DG-5078, Fatigue Management for Nuclear Power Plant Personnel at Commercial Nuclear Plants Licensed Under 10 CFR Part 53. This DG describes

Specific Requests for Comments Part 26 Page 3 of 10 methods the NRC staff considers acceptable for addressing certain aspects of FFD programs at commercial nuclear facilities licensed under part 53.

The NRC staff also intends to eventually transition this draft guide into an update to RG 5.73, Fatigue Management for Nuclear Power Plant Personnel, or the development of a new RG. At this point, NRC staff are considering four options for future RG development:

Option 1: Amend the existing RG. The NRC may develop an updated version of RG 5.73 that continues to endorse (with clarifications, additions, and exceptions) the guidance contained in NEI 06-11, Managing Personnel Fatigue at Nuclear Power Reactor Sites, Revision 1, and incorporates the topics discussed within DG-5078 as new NRC staff positions in section C of RG 5.73.

Option 2: Issue a new RG specific to part 53 licensees. The NRC may develop an entirely new RG applicable specifically to facilities licensed under part 53. This new RG would capture the guidance contained in DG-5078 and incorporate existing guidance (e.g., selected guidance in RG 5.73 and NEI 06-11) that is considered to be technology inclusive in nature. The existing guidance (i.e., RG 5.73) would remain in place as the guidance for facilities licensed under parts 50 and 52.

Option 3: Review and potentially endorse new or revised industry-developed guidance. The NRC may engage with the industry regarding a potential update to industry guidance document NEI 06-11 or the development of new, separate industry developed guidance specific to facilities licensed under part 53. The NRC would then review the new or revised industry-developed guidance within the NRCs RG process, which includes opportunities for public participation.

New or revised industry-developed guidance could incorporate DG-5078 or propose alternatives for the NRC to consider.

Option 4: Develop a comprehensive revision of the existing RG. The NRC may develop a more comprehensive revision of RG 5.73 that would explicitly detail all NRC positions reflected in the existing RG (including those endorsed positions currently contained in NEI 06-11, Revision 1),

along with the guidance of DG-5078. Such a revision would thereby be a stand-alone document, without reference to or explicit endorsement of separate, industry-developed guidance.

The NRC is seeking stakeholder input regarding which of the four options listed above would be optimal (or whether there are other options that the NRC should consider). Please provide your considerations and rationale for your recommendation.

Comment The industry fatigue group has elected option 3 to revise the current industry guidance and incorporate DG-5078. This will align with the initiative that is currently in progress to revise NEI 06-11 to better align with the existing NRC endorsement in R.G. 5.73 and correct generic formatting and sentence structure within the document.

Detailed Comments Proposed Part 26 Changes Page 4 of 10 Fitness for Duty/Access Authorization General Comment: The current operating fleet FFD program, while detailed, is structured in an easy-to-follow format. It is easy to discern where specific information is located based on the titles of the subparts. It is difficult to navigate the Part 53 FFD rule. While in one section, references are made to other sections, that may or may apply the program mandate you are researching, depending on the program you have implemented. It is very difficult to follow and make the connection with the required mandate. I am concerned that new inexperienced program personnel attempting to implement Part 53 will find it difficult to understand and implement. It seems that a detailed rule as found in Part 26 has been made more difficult to interpret and understand in Part 53. Part 53 should have been, in my opinion, much easier to read and interpret given the reduced risk represented by the advanced technology.

26.603 (d) 3 Performance Monitoring Performance Monitoring requires a documented review that monitors performance measures and thresholds on site and licensees fleet-level program performance for monitoring FFD program effectiveness.

Comment: This is a new requirement in comparison to Part 26 and seems to add a level of complexity that only complicates the administration of the program. The operating current fleet under Part 26 has been monitoring FFD program effectiveness for many years. 26.603 recognizes this. In those years sites have learned by operating experience and the data gathered to inform on actions needed to improve and adjust our programs. The data reported to the NRC annually informs operating sites of trends and the potential need for altering programs to meet societal and program challenges. The corrective action program advises sites on trending issues and informs program personnel regarding the need to alter the program based on trends. There is a performance mechanism in place that works well. I see no need to insert an additional burden for a program of lesser risk.

Employee Assistance Programs Comment: The Employee Assistance Program appears to be absent from the rule. Providing protections for employee self-referral and brokering services to help employees with mental health, substance abuse, stress, anxiety is a well-established benefit for employees. Self-referral also enhances the FFD program where employees can seek help confidentially. Employee assistance is mentioned in Draft Guide 5073 as means to help employees and improve the FFD program. The Draft Guide also mentions a group consortium EAP configuration where sites could collectively join but seems to mention EAP as an option. I believe an EAP consortium would be an excellent method for addressing the issues employees face for this unique population. Additionally, an EAP program, in a consortium or independent of a consortium, should be considered for the contract vendor population. This is an unaddressed issue in the current Part 26 program. Contract vendors who are not covered by an employer provided EAP are not afforded the same services and protections afforded a licensee employee. This gap in services has been an ongoing issue with the current operating fleet. Providing for EAP services as an option for contract employees would be an added much needed enhancement to the Fitness for Duty program.

The contract vendor would be responsible for any costs incurred, but the services and protections would be equal to employees and should be considered.

Detailed Comments Proposed Part 26 Changes Page 5 of 10 26.606 (a) Contents of Written Policy 1 (c) The consumption of alcohol within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of performing or directing the performance of work making the individual subject to the FFD program. Individuals should be informed that alcohol metabolism is a function of many variables and that 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> may not be sufficient to reduce in situ alcohol concentrations below the cutoffs in 10 CFR 26.101, Conducting a confirmatory test for alcohol, and 10 CFR 26.103, Determining a confirmed positive test result for alcohol. Regardless of when alcohol or any other substance was consumed, inhaled, or injected, a confirmed positive test result will result in an FFD policy violation and sanction in accordance with the FFD policy and procedures.

Comment: Operating experience has informed that there are numerous impairing substances exceeding that of alcohol that can impair. The five-hour rule for alcohol is well established for current operating sites, but other impairing substances should be included in this discussion.

26.609 Behavioral Observation Licensees and other entities must ensure that the individuals who are subject to Part 53 are subject to behavioral observation and that behavioral observation is performed by all individuals subject to this subpart. DG 5073 stipulates that all personnel responsible for performing BOP functions should be trained to have sufficient awareness to detect degradation in performance that may be the result of being under the influence of any substance, legal or illegal, or a physical or mental impairment that in any way may adversely affect their ability to perform their duties safely and competently.

Training must communicate the requirement to promptly report any onsite or offsite behaviors or activities by individual subject to the FFD program that may constitute an unreasonable risk to the safety or security of the NRC-licensed facility or SNM, or may cause harm to others, to the licensee-designated personnel for appropriate evaluation and action in accordance with the FFD policy.

This reporting must include any information on character or reputation indicating that the observed individual cannot be trusted or relied on to perform those duties and responsibilities or maintain FFD authorization (i.e., have access to NRC-licensed facilities or sensitive information).

This section does seem to mandate training but does not mandate a frequency for Fitness for Duty training. As compared to the operating sites, the frequency of FFD training requires an annual frequency. As written, any frequency could be adopted. Fitness for Duty training is essential to keeping the population awareness keen concerning societal changes that impact trustworthiness and reliability.

Fitness for Duty is a program that must be reinforced and is vital to safe competent operation. A training frequency that is biannual may make sense given the framework of the Part 53 program.

26.603 Determination of Fitness Details of the Determination of Fitness process and the use of professionals to assess substance abuse related issues include Medical Review Officer, Psychologist, SAE. Respective to SAE, the Part 53 rule uses the following words to describe the qualifications: accredited, educated, or trained. These words are amenable because I believe it refers to individuals with advanced education and knowledge in the substance abuse field. Individuals with advanced (masters level) university degrees whose field of study focuses on diagnosis and treatment of substance abuse should be used in the industry. Part 26 has restrictive language that does not include this important population. As a result, individuals who are

Detailed Comments Proposed Part 26 Changes Page 6 of 10 highly trained and capable of performing high quality assessment are not considered. Personnel who possess advanced degrees include administrators of drug and alcohol programs, university professors who possess years of advanced training that exceed that of other program professionals (MROs, Psychologist) who have a license to practice medicine or psychology but lack the experience and training of specially trained substance abuse personnel. Part 53 should recognize this disparity and allow for the use of these personnel in the Fitness for Duty program under Part 53. Referring to the verbiage in 10 CFR 26.187 would only replicate the problems experienced by the current operating fleet.

26.619 Determination of Fitness, For Cause 26.619 would explicitly permit determinations to be performed via electronic means, so long as those determinations are supported by an appropriately trained individual who is present in-person with the individual being assessed.

Comment: This is viewed as a significant enhancement to the existing Part 26 requirements which require that these determinations be conducted face to face. However, the phrase appropriately trained individual should be defined. This could manifest as a limiting and burdensome requirement just as the current face to face requirement is. Additionally, considering the advancement to telehealth and medicine, it seems reasonable to allow these determinations, without a third party, if the integrity and security of the interview can be assured. It should be noted that behavior observation under the current behavior observation program can be conducted electronically. Additionally, clinical interviews conducted by psychologists are routinely conducted and completed electronically. Additionally, current APA guidelines for psychologists may prohibit an additional person in the room based on data that suggest the interviewing individual could be guarded or reticent to speak about personal information with a third party in the room. Protection of information is also a consideration. Therefore, the introduction of a third party in an interview typically reserved for the licensed professional and interviewee should weigh the potential compromise of program integrity.

DG 10 CFR 26.603(a)(4), Drug and Alcohol Testing and Fitness Determinations A determination of fitness is typically the process entered when there are indications that an individual specified in 10 CFR 26.4 may be in violation of the licensees or other entitys FFD policy or procedure; For example, the individual was identified as using an illegal substance or is a member of a group acting or advocating for an unlawful change to the U.S. government or violence to a particular ethnic, religious, or cultural group. A determination of fitness may require the use of a medical or clinical professional, called on by the licensee or other entity, to evaluate the individual, formulate a treatment plan, and recommend whether the individuals authorization should be reinstated. The types of professionals called upon to make this determination should be educated, accredited, or trained in the specific area(s) of concern (e.g., drug or alcohol abuse, psychosis, etc.).

Comment: This language allows for personnel who have advanced training in the focus areas such as substance abuse to administer services to nuclear programs. Personnel with advanced degrees who have completed years of substance abuse training should be utilized and recognized as experts in their field to administer services to nuclear personnel.

Detailed Comments Proposed Part 26 Changes Page 7 of 10 26.607 Random Drug Screening Part 53 stipulates random drug and alcohol testing is conducted at an annual random testing rate greater than or equal to 50 percent for the population of individuals subject to testing and there is no prior notice to report for random testing.

Testing 50 percent of a population that is estimated to be substantially smaller than a current operating site may be difficult to implement and manage. Adjustment of percentage for random testing with smaller populations should be considered. Example: A smaller percentage, 10-20 percent quarterly for random screening. Perhaps a consortium could be employed where licensees would participate with other licensees in a combined pool. Perhaps the annual rate of testing could be adjusted annually as seen in the DOT program.

Draft Guide 10 CFR 26.4, FFD program applicability to categories of individuals.

The MRO should be considered as FFD program personnel if involved in the day-to-day operations of the program, regardless of full-or part-time employment, or location of employment because the MRO must review positive test results obtained from an HHS-certified laboratory and should assist FFD program staff in the evaluation of subversion attempts.

Comment: The MRO is typically considered FFD program personnel by virtue of the importance of duties assigned as the Medical Review Officer. Clarity in language would help staff assign the MRO as FFD staff, if appropriate. There is also confusion as to whether the MRO should be assigned as a critical group member. Clarity in language and flexibility in administration is recommended.

Draft Guide 5073, 10 CFR 26.607 For Cause The Draft Guide states (p76) that, For indications of possible impairment that do not create a reasonable suspicion of alcohol or substance abuse, the licensee or other entity should permit the individual to return to work only after the physiological or psychological condition is evaluated by the MRO and the MRO has determined that the individual is fit to perform his or her duties safely and competently (see 10 CFR 26.619, 10 CFR 26.77, and 10 CFR 26.189).

Comment: This is inconsistent with the tenet that only licensed professionals with the requisite knowledge of their training and experience. 26.189 states, A determination of fitness must be made by a licensed or certified professional who is appropriately qualified and has the necessary clinical expertise, as verified by the licensee or other entity, to evaluate the specific fitness issues presented by the individual. In this section, referencing the MRO as the evaluator of psychological conditions without clarification that the medical official has or must have the required psychological credentials is confusing Additionally, this section states, For indications of possible impairment that do not create a reasonable suspicion of alcohol or substance abuse, the licensee or other entity should permit the individual to return to work only after the physiological or psychological condition is evaluated by the MRO and the MRO has determined that the individual is fit to perform his or her duties safely and competently. (see 10 CFR 26.619, 10 CFR 26.77, and 10 CFR 26.189).

For cause conditions for observed behavior do require a drug and alcohol screen. However, operating experience from the current reactor fleet current has revealed that substance use can alter behavior in a

Detailed Comments Proposed Part 26 Changes Page 8 of 10 manner that does not readily manifest itself as a symptom for substance abuse. These behaviors can be subtle and do not overtly present as substance use or abuse. Operating experience has revealed substance use/abuse as a contributor to behaviors that do not overly present as anomalous. Removing drug use from the contributors of possible impairment is an essential component for establishing the cause of impairment. Programs need to eliminate all potential contributors of anomalous behaviors when entering into a Determination of Fitness evaluation. Even when an identified issue is identified where drug and alcohol use is not overtly suspected but may be contributing to the impairment or behavior.

10 CFR 26.610 Sanctions Section 12 of the 5073 Draft Guide states the following:

The guidance below is based on lessons learned from the LLWR community and informed by the expectation that facilities licensed under 10 CFR Part 53 may be licensed and operated with staff sizes markedly smaller than those of current LLWR facilities, and these facilities may be sited in geographically remote locations. Facilities with small staff sizes set up the paradigm that the relative contribution of an individual to safety and security may be greater than at facilities with larger staff, despite advances in passive and automated technologies. For example, if a 10 CFR Part 53 CNP has one individual performing radiation protection, chemistry, or health physics activities, two NRC-licensed individuals per shift, and a few onsite NRC-required security officers, it is possible that if one person in a particular labor category is impaired, the effectiveness of the team may be diminished. For this reason, the regulations and guidance escalate the severity of sanctions based on risk.

In determining the schedule of sanctions for violations of the FFD policy, licensees or other entities could assign individuals to one of three sanction groups (Group 1, 2, or 3), based on the risk significant level of their assigned duties and responsibilities. Group 1 could be the highest risk significance level and Group 3 the lowest. Licensees or other entities should perform a risk-significance assessment and assign individuals to groups based on the results of that assessment. For guidance, individuals who perform the following duties and responsibilities, regardless of their other job duties, should be assigned to Group 1.

These are the individuals who are empowered, entrusted, assigned, and possibly licensed by the NRC to perform or direct those duties and responsibilities that make them subject to the FFD program.

Comment: It is not entirely clear why a Part 53 site requires a more stringent sanction for an FFD policy violation than a current LLWR. Part 53 sites are projected to operate with minimal personnel as noted in bold print above, but this does not constitute a higher sanction. The intent is to sanction the individual for violating the fitness policy. Increasing the sanction based on factors other than the individuals failure to meet fitness for duty mandates and expectations seems out of place.

Additionally, due to the remote location of some Part 53 sites, it seems reasonable, based on earlier comments respective to psychological evaluation, to consider remote mental health and substance abuse treatment. Telehealth options that meet treatment protocols and provide maximum flexibility for the individual to receive treatment is a prescription for restoring the employee to work productivity as soon as reasonably achievable. Treatment plans are varied and some, depending on the diagnosis, will require more time and intensive treatment. I recommend consideration for the minimum sanction to remain at fourteen days and a repeat violation at three-five years. A third violation would mean a user

Detailed Comments Proposed Part 26 Changes Page 9 of 10 should undergo a permanent sanction, having attempted on two occasions to rehabilitate unsuccessfully. Personnel with chronic addictive and comorbid diagnosis could be out of work for months with some requiring inpatient treatment requiring extended leave. Increasing the sanction for those who make a first-time wrong decision involving experimental drug use should not be subjected to more than the minimum sanction of fourteen days. In my experience, these types of program offenders recognize the gravity of their decision and do well with education and counseling and are back in the workplace performing well within the minimum sanction period.

Access Authorization Section §73.120(c)(2) establishes behavioral observation requirements, which are an awareness initiative for recognizing behaviors adverse to the safe operation and security of the facility through observing the behavior of others in the workplace and reporting aberrant behavior or changes in behavior that might reflect negatively on an individual's trustworthiness or reliability. Maintaining behavioral observation would assist and/or improve worker safety and reduce the risk of an insider threat. This proposed requirement in §73.120(c)(2) would be a scaled version of the full BOP required under §73.56(f).

Comment: It seems unreasonable and unnecessarily risky to ask a population of personnel to observe behaviors without any training that informs of what to look for. Human behavior is precarious and factors such as mental health and substance abuse can alter human behavior. Understanding these characteristics is essential in identifying behavior that can change subtly and become volatile if not detected early. Additionally, insider threat is a component that raises the level of necessity for training in behavioral observation. Recommend consideration for a training frequency of bi-annual without an exam requirement and could be framed as awareness training.

DG 5074, Reviewing Official If an individual who has UA has not entered the protected area, material access area, or controlled access area for more than 30 continuous days, the individuals UA status should be suspended until the reviewing officials have evaluated the lapse in time to reinstate UA status.

Comment: Under the current 73.56 program, the standard for suspension of access authorization is someone that has not been under behavior observation for greater than 30 days. Work related reasons for not entering the protected area may include temporary reassignment and other circumstances.

However, the individual may remain observed under these conditions.

Recommendation: The DG should consider linking the suspension of access to incidences where the individual is unobserved. The 30-day period of not entering the protected area may inform that the individual is no longer properly observed and should be evaluated by the Reviewing Official to determine if the individual is no longer compliant.

5074 Behavior Observation Paragraph 112. Integral to the licensees BO elements for all employees is an annual management review of employee behavior, conducted by a reviewing official and the employees supervisor. This review enables interaction between the reviewing official and their designee (i.e., the employees immediate supervisor) and the employee, letting the supervisor become aware of any condition that

Detailed Comments Proposed Part 26 Changes Page 10 of 10 may cause the employee to act or behave unconventionally. The review also lets supervisors consider whether it may be necessary to refer the employee for additional medical or psychological assistance.

Comment: Behavior Observation is a continual requirement. Adverse behaviors determined to be aberrant and concerning are to be reported when detected. The requirement places an additional burden on program personnel, as an annual requirement mandates tracking and very often chasing Supervisors to complete a form. There is little value in this since as demonstrated by the requirement for the current operating fleet.

Recommendation: Under the current operating program, personnel are reporting behaviors that are acted on and addressed by program personnel, only to repeat the same reporting at the end of the year for the annual submission for form. Supervisors and all personnel should be reporting behaviors immediately to access authorization personnel and documenting appropriately when doing so. There does not seem to be high value in the annual reporting form and recommend consideration for not requiring its use.

Specific Requests for Comments Part 73 Page 1 of 6 Topic 1 - 73.100 - Physical Security NRC Request The proposed § 73.100 would identify the proposed performance-based physical security requirements with which future commercial power reactor applicants or licensees physical protection programs would need to demonstrate compliance, without prescribing the specific methods that must be used to satisfy them. Applicants and licensees would have increased flexibility regarding the modern technologies and methods that they could use. Implementing guidance in DG-5076 (proposed RG 5.97),

Guidance for Technology Inclusive Requirements for Physical Protection of Licensed Activities at Commercial Nuclear Plants, would be available to assist applicants and licensees. For example, DG-5076 provides detailed guidance, including performance standard recommendations, on the probability of detection and alternative sources of power for exterior intrusion detection systems (subsection 4.1.1.1.A), interior intrusion detection (subsection 4.1.1.1.B), intrusion assessment (subsection 4.1.1.2.A), security response/neutralization subsection (4.1.1.4.A), security communication (subsection 4.1.1.3.A), and security delay (subsection 4.1.1.4.C).

Does the NRCs proposed approach in § 73.100 provide a sufficient level of detail to be readily understood and easily applied to the licensing and oversight of new and advanced power reactors, or should the NRC consider moving some objective and measurable security performance standard recommendations from the draft implementing guidance in DG-5076 into proposed § 73.100? If so, which objective and measurable security performance standard recommendations should be moved from DG-5076 to § 73.100? Please provide the basis for your response.

Comment NEI response - Yes, § 73.100 provides a sufficient level of detail to be readily understood and easily applied to the licensing and oversight of new and advanced power reactors. No objective and measurable security performance standard recommendations need to be moved from the draft implementing guidance in DG-5076 into proposed § 73.100. NEI finds the requirement descriptions in § 73.100 adequate for an applicant or licensee to understand what is necessary to achieve NRC-approved security plans. Maintaining the recommended performance attributes and implementing instructions (methods) in guidance will allow designers maximum flexibility to use modern technologies and methods to meet the requirements in § 73.100.

Topic 2 - 73.110 - Cybersecurity NRC Request The proposed § 73.110 would require licensees to demonstrate protection against cyberattacks in a manner that is commensurate with the potential consequences from those attacks, without prescribing the specific methods that must be used to demonstrate protection. Under proposed § 73.110(a),

licensees would need to ensure that digital computer and communications systems are adequately protected against a potential cyberattack that would, for example, result in adverse impacts to the physical security digital assets used by the licensee to prevent unauthorized removal of material per

§ 53.860(a). Protecting against such a potential cyberattack would involve requiring cybersecurity for

Specific Requests for Comments Part 73 Page 2 of 6 SNM at a commercial nuclear reactor licensed under part 53. Applicants and licensees would have increased flexibility regarding the modern technologies and methods that they could use for protecting against such a potential cyberattack. Detailed implementing guidance in DG-5075 (proposed RG 5.96),

Establishing Cybersecurity Programs for Commercial Nuclear Plants licensed under 10 CFR part 53, would be available to assist applicants and licensees. For example, DG-5075 provides guidance on the implementation of security by design features (e.g., facility design) for negating the potential consequences from such a potential Cyberattack.

If a cyberattack were to compromise the availability, integrity, or confidentiality of data or systems associated with security systems/measures for the protection of SNM at a commercial nuclear reactor licensed under part 53, do the potential consequences warrant requiring cybersecurity for such material? Please provide the basis for your response including a detailed explanation of challenges, if any, posed by requiring cybersecurity for SNM at a commercial nuclear reactor licensed under part 53.

Comment The cyber security program, as an integrated component of the physical protection program has, as its objective, protecting against the design basis threat (DBT) of radiological sabotage.

Proposed 10 CFR 73.110 provides the programmatic requirements to protect against the radiological sabotage cyberattack. Licensees may rely on certain functions of structures, systems and components (SSCs) to prevent radiological sabotage. These functions must be protected against the DBT cyberattack. Licensees would analyze these functions and the associated digital assets to identify whether a cyberattack would prevent the satisfactory completion of those functions and result in radiological sabotage. Security controls would be implemented to ensure those digital assets are protected with reasonable assurance.

As sites perform cybersecurity analyses for the functions that need to (or not) be protected, and potential consequences from cyberattacks that would result from compromise are identified, the outcome of this analysis will inform the control measures necessary to protect those functions from adverse impacts. It would be expected that the approach would be risk-informed and consequence-based, addressing the range of technologies being developed within the advanced reactor community.

The industry believes the graded approach identified within 10 CFR 73.110, focused on radiological consequences, supports a positive step towards architecting a cyber security plan for protecting the future reactor fleet.

Detailed Comments for Proposed Revisions to Part 73 Page 3 of 6 Rule Language Comment Number NRC Preliminary Requirement Industry Comment Proposed Resolution 73-1 73.100 Technology-inclusive requirements for physical protection of licensed activities at commercial nuclear plants against radiological sabotage.

The proposed general performance objective presented in § 73.100(b)(1) is overly broad, i.e., lacks specificity as to the actual performance-based requirement. NEI recommends that the NRC develop a new performance objective with a clear performance-based requirement, and offers two options for consideration:

Preferred approach - As noted in the SOC, the proposed revisions to part 73 are intended to establish a new consequence-based approach for a range of security areas. Given this purpose and the bedrock role of target sets in determining facility protection requirements, an ideal general performance objective would be one that incorporates these two concepts. For example, a single performance objective could read:

The licensee must establish, implement, and maintain a physical protection program and a security organization providing reasonable assurance that (1) no target set objective can be achieved by the design basis threat of radiological sabotage; or (2) measures can be taken to prevent a release of radionuclides resulting in consequences exceeding the offsite dose reference values defined in

§53.210 of this chapter in the event a target set objective is achieved.

Alternative approach - A general performance objective could also be based on text found on page 86976 of the Federal Register Notice. As shown below, this statement of the objective directly reflects the intended consequence-based approach to security:

The licensee must establish, implement, and maintain a physical protection program and a security organization that provides reasonable assurance of adequate protection against any deliberate act by the design basis threat of radiological sabotage which could directly or indirectly

Detailed Comments for Proposed Revisions to Part 73 Page 4 of 6 endanger the public health and safety by exposure to radiation.

Under either approach, the necessary consequence limits could be included in the SOC and guidance.

73-2 73.100(b)(3)(iv)

The proposed regulations require the physical protection program to be designed to provide timely security response to interdict and neutralize adversary attacks up to and including the design basis threat of radiological sabotage. The NRC should reconsider the interdiction and neutralization requirement in light of the following discussion.

NRC policies encourage designers of new reactors to incorporate features that can provide a more robust and effective security posture with less reliance on operational programs, i.e.,

incorporate security-by-design.

Recognizing that many different advanced reactor designs and approaches to security-by-design are possible, the NRC staff developed the proposed § 73.100 requirements with a goal to make them technology-inclusive. While NEI supports this goal and appreciates the NRC staffs efforts toward it, we believe there is a category of advanced reactor that was NEI recommends the NRC amend the proposed § 73.100 requirements concerning armed responders to recognize and accommodate facilities that will not require an armed response to maintain releases below those with consequences exceeding the offsite dose reference values defined in §53.210(b). These changes would incentivize more robust approaches to security-by-design, consistent with NRC policy, and ensure that no unnecessary regulatory burden is imposed on an applicant or licensee (i.e., only require what is necessary for adequate protection). In the approach envisioned by NEI, the licensee would be relieved of requirements related to an interdiction and neutralization capability, but, as a defense-in-depth measure, still be required to notify a supporting offsite agency or organization capable of providing an armed response. Because an armed response by a certain time would not be necessary to ensure that a release does not exceed a consequence limit, the timeline of an actual response to an attack would be determined by the incident commander (i.e., the lead law enforcement official dispatched to the site) based on public and plant worker safety considerations.

An alternative approach to addressing this comment would be to retain the requirement for an applicant or licensee to have an interdiction and neutralization capability but remove the requirements in § 73.100(b)(3)(iv)(A)(2) through (5). Again, this option could be available if a threat-based analysis, with no consideration of on-site or offsite armed responders, demonstrates that the consequences do not exceed the offsite

Detailed Comments for Proposed Revisions to Part 73 Page 5 of 6 not adequately considered in the development of the proposed rule.

These are facilities that, if attacked by the Design Basis Threat (DBT) of radiological sabotage, possess the safety and security design features necessary to preclude an offsite consequence (dose) limit from being exceeded without any actions by an armed response force. As currently proposed, this type of facility would be required to have an interdiction and neutralization capability, even though actions by armed responders would not be necessary for adequate protection of public health and safety.

This outcome is a disincentive for designers to incorporate features that maximize security-by-design.

dose reference values defined in §53.210(b). As above, the licensee would be required to notify a supporting offsite agency or organization capable of providing the armed response, but the timeline of the response would be determined by the incident commander (i.e., the lead offsite official on the scene) based on public and plant worker safety considerations.

Detailed Comments for Proposed Revisions to Part 73 Page 6 of 6 Guidance Cmt.

Section/

Paragraph Comment Justification

1.

A3.2 of DG-5075 CDAs classified as Most Critical should only need to implement the security controls needed to protect the CDA from cyber attacks (from all attack pathways). This includes the necessary defense-in-depth. Or be able to define certain controls as additional margin available.

This provides a more reasonable approach to protecting the CDA aligned with the operating fleet. Focus should be performance driven rather than compliance driven.

2.

A3.2 of DG-5075 In the discussion of Most Critical CDAs, the term near real-time assurance is used. This is vague and will lead to a variety of interpretations.

3.

Definitions in DG-5075 Definition of SIEM - also uses the term near real-time

4.

Overall Comment Three-Tiered Review - should allow if an alternate method is available and trained upon, that the digital asset does not need to be included This approach would be consistent with how the current operating fleet addresses Emergency Preparedness digital assets in NEI 10-04, Revision 3. An alternate method that is adequately independent and diverse credited for performing the EP function must be available in sufficient time to detect the compromise of the DA. Detecting the compromise in sufficient time ensures the licensee can implement an alternate method to perform the EP function(s).

5.

DG-5075 Several areas of the guidance state that the analyses should be revisited periodically, NRC should provide guidance on what is meant by periodically.