ML26013A095
| ML26013A095 | |
| Person / Time | |
|---|---|
| Site: | Nuclear Energy Institute |
| Issue date: | 12/24/2025 |
| From: | Csontos A Nuclear Energy Institute |
| To: | Office of Administration |
| References | |
| NRC-2025-0149, 90R53009 00005 | |
| Download: ML26013A095 (0) | |
Text
PUBLIC SUBMISSION As of: 1/13/26, 9:42 AM Received: December 24, 2025 Status: Pending_Post Tracking No. mjk-8iqd-5q38 Comments Due: December 24, 2025 Submission Type: Web Docket: NRC-2025-0149 Draft Interim Staff Guidance: DSS-ISG-2025-XX, "Treatment of Certain Loss-of-Coolant Accident Locations as Beyond-Design-Basis Accidents" Comment On: NRC-2025-0149-0001 Draft Interim Staff Guidance: Treatment of Certain Loss-of-Coolant Accident Locations as Beyond-Design-Basis Accidents Document: NRC-2025-0149-DRAFT-0005 Comment on FR Doc # 2025-20707 Submitter Information Organization:Nuclear Energy Institute General Comment See attached file(s)
Attachments 12-24-25_NEI Comments on Draft Interim Staff Guidance Docket ID NRC-2025-0149 1/13/26, 9:42 AM NRC-2025-0149-DRAFT-0005.html file:///C:/Users/BHB1/Downloads/NRC-2025-0149-DRAFT-0005.html 1/1 SUNI Review Complete Template=ADM-013 E-RIDS=ADM-03 ADD: Carolyn Lauron Mary Neely Comment (5)
Publication Date:
11/24/2025 Citation: 90 FR 53009
Dr. Aladar Csontos Director, Fuel and Radiation Safety Phone: 202.557.9727 Email: aac@nei.org December 24, 2025 Office of Administration Mail Stop: TWFN-7-A60M U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTN: Program Management, Announcements and Editing Staff
Subject:
NEI Comments on Draft Interim Staff Guidance (ISG), DSS-ISG-2025-XX, Treatment of Certain Loss-of-Coolant Accident Locations as Beyond-Design-Basis Accidents, Docket ID NRC-2025-0149 Project Number: 689 Submitted via regulations.gov
Dear Program Management,
Announcements and Editing Staff:
The Nuclear Energy Institute (NEI)1, on behalf of its members, appreciates the opportunity to review and comment on the subject Draft Interim Staff Guidance (ISG), DSS-ISG-2025-XX, Treatment of Certain Loss-of-Coolant Accident Locations as Beyond-Design-Basis Accidents. The guidance found in this draft ISG describes the mechanistic considerations that the NRC staff may consider in determining whether an applicant has proposed an adequately protective design-basis (DB) LOCA spectrum. As described in the Implementation section, NEI understands this draft guidance to have a limited scope of applicability, specifically to new and advanced reactor design license applications. Consistent with this interpretation, NEI does not provide comments or recommendations addressing the use of analogous guidance or regulatory positions herein for existing or currently licensed facilities.
To avoid ambiguity and potential misapplication, NEI recommends that this limited applicability be stated more explicitly and prominently in the draft ISG. Clear articulation of the intended scope would help ensure the 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.
Office of Administration December 24, 2025 Page 2 Nuclear Energy Institute guidance is applied as intended and would reduce the risk of it being construed as establishing expectations for other licensing or regulatory contexts.
The NRC staff plans to review applications in accordance with the NRC interpretation in this draft guidance until it is incorporated into more durable regulatory guidance, and if the staff determines the application includes adequate justification, an exemption from the LOCA evaluation model requirements of 10 CFR 50.46 would not be needed. The industry supports the intent of the draft guidance to allow new and flexible approaches based on consideration of risk insights rather than deterministic approaches. The industry also appreciates the NRCs efforts to incorporate lessons learned from recent staff review of advanced reactor applications, such as the NuScale Power LLCs US460 small modular reactor design. This ISG should improve regulatory clarity and predictability for future LOCA submittals by advanced reactor vendors with changes to the guidance described below.
Over the past 30 years, NRC has sought to advance a risk-informed and performance-based (RIPB) regulatory framework, consistent with the longstanding Commission policy and RIPB concepts defined in SRM-SECY-98-144, White Paper on Risk-Informed and Performance-Based Regulation. A RIPB approach ensures that regulatory requirements focus on safety outcomes and focus on the most safety-significant aspects of nuclear technologies while providing flexibility in determining how to meet established performance criteria in a way that will encourage and reward improved outcomes and focus on the results as the primary basis for regulatory decision-making.
The Nuclear Energy Innovation and Modernization Act (NEIMA) of 2019 mandates RIPB and technology-inclusive approaches to licensing. Under NEIMA, a light-water reactor (LWR) with significant improvements over large LWRs under construction in January 2019 is an advanced reactor. As such, the draft ISG should incorporate RIPB attributes that are inclusive of advanced LWRs. However, the draft ISG is not performance-based as written. Even within its intended and explicitly limited scope for LWRs (including new and advanced LWRs), it retains prescriptive language. Pages 4-5 set out specific design features (e.g., reduced penetrations, large water inventory, passive injection), inspection expectations, and fracture-mechanics or probabilistic methods to justify treating certain LOCA locations as beyond-design-basis (BDB). This anchors regulatory acceptability to a particular LWR accident construct and prescribes how applicants must demonstrate acceptability, rather than establishing performance criteria tied to safety functions. As a result, the ISG does not reflect a performance-based framework; instead, it reinforces event-specific analytical obligations that may not be necessary when broader safety-function performance could provide a sufficient safety margin and basis for regulatory decisions.
We recognize another challenge with NRC staff practices. In several sections of the draft ISG, the guidance refers broadly to reactor coolant pressure boundary ruptures and primary system component failures. Given that the stated purpose of the guidance is to address alternative compliance demonstrations for 10 CFR 50.46, this terminology could be misinterpreted and lead to regulatory ambiguity. The LOCA definition in 10 CFR 50.46 is explicitly limited to postulated breaks in reactor coolant system piping. The regulation does not require demonstration of compliance for failures of vessels or other non-piping primary system components. As such, references to reactor coolant pressure boundary ruptures or primary system component failures may be read as
Office of Administration December 24, 2025 Page 3 Nuclear Energy Institute expanding the scope of 10 CFR 50.46 beyond its current regulatory basis. NEI recommends that the draft ISG clearly distinguish between piping breaks, which are within the scope of 10 CFR 50.46, and other types of reactor coolant pressure boundary failures, which are not. Clarifying this distinction ensures that the guidance is interpreted and applied consistently with the regulation and does not inadvertently imply new or expanded compliance expectations without conducting the appropriate 10 CFR 50.109 backfit compliance evaluations.
Although the process for determining BDB break location is overly prescriptive, the guidance proposed for the elimination of exemptions is helpful and would benefit from additional explanation and clarification. Within the Rationale section, the draft ISG briefly references three closely related regulatory positions. Greater exposition of these positions would promote consistent interpretation and application by both NRC staff and licensees.
Industry fully supports these positions and seeks clearer articulation to ensure a common and durable understanding.
- 1. Certain LOCA break locations may be characterized and analyzed as BDB without the need to obtain an exemption from the LOCA evaluation model requirements of 10 CFR 50.46.
- 2. Once a LOCA break location is appropriately characterized asBDB, the consequences associated with LOCAs at that break location are excluded from the DB considerations of all other applicable requirements, including 10 CFR 50.49 and GDCs 19, 35, 38, 41, 44, and 50. Accordingly, no additional analyses or demonstrations (e.g., best-estimate containment accident pressure analyses) are required to establish compliance with these requirements for BDB LOCA break locations.
- 3. Consistent with this framework, an exemption from the requirements of 10 CFR 50.49 or GDCs 19, 35, 38, 41, 44, and 50 is not required in order to exclude LOCA break locations that have been properly characterized as BDB.
Explicitly articulating these positions in the ISG would reduce the potential for misinterpretation, reinforce the intended RIPB regulatory framework, and support predictable and consistent application across licensing actions.
Within the guidance section, there is a potentially significant regulatory position that warrants clarification.
Specifically, with respect to demonstrating compliance with 10 CFR 50.46, the guidance states that the staff determination may be based on the use of more restrictive criteria than those in 10 CFR 50.46(b) (such as no core uncovery or no departure from nucleate boiling), but the BDB (realistic) evaluation methodology results may also be compared to the 10 CFR 50.46(b) acceptance criteria.
NEI understands the intent of this guidance is to allow applicants the flexibility to propose figures-of-merit and acceptance criteria that are more conservative than those specified in 10 CFR 50.46(b), for either design-basis accident (DBA) or beyond-design-basis accident (BDBA) break locations. Under this interpretation, the NRC would accept an applicants voluntary use of these more restrictive criteria in lieu of an explicit demonstration against the 10 CFR 50.46(b) acceptance criteria.
Office of Administration December 24, 2025 Page 4 Nuclear Energy Institute However, as currently written, the guidance could be interpreted as authorizing the staff to impose acceptance criteria that are more restrictive than those established in regulation, rather than limiting such criteria to those voluntarily proposed by the applicant. This ambiguity should be resolved.
Accordingly, the guidance should clearly state that:
- 1. It is acceptable for an applicant to propose and self-impose more restrictive figures-of-merit and acceptance criteria;
- 2. Given the risk-insignificant nature of BDBA break locations, the guidance should clarify that regimented or prescriptive reporting requirements analogous to those in 10 CFR 50.46(a)(3) are not necessary for BDBA evaluations; and
- 3. These approaches can be implemented within the existing regulatory framework, without the need for an exemption from 10 CFR 50.46.
This clarification would better align the guidance with regulatory intent, preserve appropriate applicant flexibility, and avoid implying that the staff may impose requirements beyond those prescribed in regulation.
The comments provided in this letter and attachment are from several NEI members to include advanced reactor developers, fuel vendors, and operating light-water-reactor utilities. We trust that you will find the attached questions and comments useful and informative as you work on finalizing the draft ISG. If you have any questions or require additional information, please contact me at aac@nei.org Sincerely, Aladar Csontos Director, Fuel and Radiation Safety Attachment C:
Carolyn Lauron, NRR//DNRL/NLIB, NRC Vic Cusumano, NRR/DSS, NRC NRC Document Control Desk
- Consolidated Comments on draft Interim Staff Guidance (ISG), DSS-ISG-2025-XX, Treatment of Certain Loss-of-Coolant Accident Locations as Beyond-Design-Basis Accidents Comment Number Section Comment/Basis Recommendation
- 1.
Purpose The wording would normally be analyzed implies that the certain break locations should be DB, but an allowance is being granted by the ISG. It may be beneficial to indicate that the ISG guidance is helpful for addressing locations where the need to assume a break and provide supporting analysis is nuanced or unclear.
Suggested rewording would normally be analyzed to "may be considered"
- 2.
Background
The emergency core cooling system (ECCS) performance requirements in Title 10 of the Code of Federal Regulations (10 CFR), Section 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, assume as their starting point that a LOCA has occurred. Such an approach is called non-mechanistic and presumes reactor coolant pressure boundary (RCPB) rupture without regard to cause.
This description of the text of 10 CFR 50.46 is misleading since it refers to RCPB ruptures generally and 10 CFR 50.46 only applies to a specific type of RCPB rupture: ruptures in pipes.
10 CFR 50.46 states:
As used in this section: (1) Loss-of-coolant accidents (LOCA's) are hypothetical accidents that would result from the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant makeup system, from breaks in pipes in the reactor coolant pressure boundary up to and including a break equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant system.
While the introductory section of 10 CFR 50.46 could be interpreted to apply to any RCPB rupture this definition of a LOCA clarifies that it is only applicable to pipe ruptures.
Revise the text to apply only to pipe ruptures, for example:
Such an approach is called non-mechanistic and presumes reactor coolant pressure boundary (RCPB) rupture in a pipe without regard to cause.
Similar characterizations appear throughout the document, all of which should be addressed.
The basis of this proposal is to make the text consistent with 10 CFR 50.46.
It is clear from the text of 10 CFR 50.46 that breaks in locations other than pipes are already outside of the regulatory requirements.
- 3.
Background, 2nd paragraph, 4th sentence One of the issues that drove the need for this ISG was disagreement between some members of the NRC technical staff and industry applicants regarding whether Suggested rewording of does not apply to has not been previously applied
Comment Number Section Comment/Basis Recommendation probability could be applied. The wording does not apply implies an NRC belief that the answer is a definitive no. It is not clear why that should be the case from a safety perspective, which is part of the change being allowed under the ISG. It may be beneficial to indicate that the ISG guidance is helpful for addressing this as an evolving position.
- 4.
Background, 2nd paragraph, last sentence Some readers may interpret "relaxed" as less safe when that is not the case. It may be beneficial to re-word consistent with later discussion (e.g., pages 3 and 5) that use the term "realistic" in this context.
Suggested rewording relaxed to the term more realistic
- 5.
Background, 3rd paragraph, 1st sentence Similar to the earlier comment in the Purpose section, part of the disagreement between the NRC and industry applicants was what constituted a break location or not. Calling this an alternative interpretation of the design-basis" seems to imply that there is a "normal" interpretation that was mutually agreed upon at some point. Since that is not the case, rewording may be beneficial.
Reword an alternative interpretation of the design-basis to a different approach to the design-basis
- 6.
Rationale, 1st paragraph, last sentence Although 10 CFR 50.46 is at the center of many prior discussions between NRC and industry, NRC technical staff has also cited the 10 CFR Appendix A definition of LOCA, GDC 35, and 10 CFR Appendix K as related regulations for LOCA break designations. The NRC technical staff has then requested exemptions from these other regulations in addition to 10 CFR 50.46 when using a different approach. If the vision of the ISG is to avoid unnecessary exemptions under this process, then the other regulations should be mentioned as well.
Include other regulations such as 10 CFR 50 Appendix A definition, 10 CFR 50 GDC 35, and 10 CFR 50 Appendix K
- 7.
Rationale, 3rd paragraph Editorial correction - delete extra space between the dash and "specific"
Comment Number Section Comment/Basis Recommendation
- 8.
Rationale -
Footnote 1 See previous comment where this footnote is used in the text. The NRC technical staff has sometimes referenced other regulations as needing exemptions in addition to 10 CFR 50.46.
Include other regulations such as 10 CFR 50 Appendix A definition, 10 CFR 50 GDC 35, and 10 CFR 50 Appendix K for consistency with other comment above.
- 9.
Guidance -
generic The ISG is focused on recent precedent to establish the framework within the ISG.
Language should be added to ensure that other acceptable paths to implement a risk-informed or RIPB still remain a potential acceptable approach.
- 10.
Guidance -
- 2. Design and operational programs provide assurance that failures at the location of interest are highly unlikely.
The third bullet - this information does not seem necessary to specify in an application because it is under the jurisdiction of the Corrective Action Program (CAP), a 10 CFR 50 Appendix B requirement. If a flaw is identified, the CAP would be relied upon to take appropriate actions. Evaluations under CAP would include review of analyses or assumptions that may be impacted. Other existing regulations, such as 10 CFR 50.59, would then determine if the resolution strategy pursued under CAP would require NRC review and approval.
Remove the third bullet.
- 11.
Guidance -
- 2. Design and operational programs provide assurance that failures at the location of interest are highly unlikely.
4th bullet discusses demonstration of low probability of component(s) failure through either qualitative or quantitative methods.
The qualitative methods need to be clarified to characterize acceptable approaches.
Delete the parenthetical information in last sentence of Bullet 4 and replace it with: NUREG-1829 and NUREG-1903 can provide possible basis to demonstrate failures of certain size pipes have a low probability of occurrence.
- 12.
Guidance -
- 2. Design and operational programs provide assurance that failures at the location of interest are highly unlikely.
4th bullet indicates that the analyses should also account for non-piping impacts. However, the definition of LOCA is limited to piping breaks.
Remove sentence: These analyses should also account for non-piping impacts and indirect piping failures.
Comment Number Section Comment/Basis Recommendation
- 13.
Guidance -
- 3. Realistic, best-estimate analyses of LOCAs at the location of interest as beyond design-basis accidents demonstrate that the consequences would be acceptable.
The ISG should define the analyses as BDBA LOCA analyses rather than realistic, best estimate analysis.
Significant confusion exists regarding terminology such as realistic or best-estimate analysis. 10 CFR 50.46, regulatory guide 1.157, topical reports, and licensing amendment requests have used these terms inconsistently. For instance, realistic system codes may be used as part of high level of probability of Appendix K style EMs.
Change the title of Item 3 to Beyond Design Basis Accident LOCA analyses demonstrate that the consequences would be acceptable.
This would improve clarity in the communication.
- 14.
Guidance -
- 3. Realistic, best-estimate analyses of LOCAs at the location of interest as beyond-design-basis accidents demonstrate that the consequences would be acceptable.
Item 3 states: The reviewer should determine whether the application demonstrates that the fuel remains within acceptable design limits for all LOCAs, regardless This is confusing since it refers to 10 CFR 50.46 as a fuel performance rule. While the rule historically used cladding post-quench ductility as a conservative means to ensure core coolability, 10 CFR 50.46 is an ECCS performance rule.
The ISG should be revised to clarify the intention of the ECCS demonstration, coolability and dose, rather than fuel performance. The historical fuel-specific performance criteria in 10 CFR 50.46(b)(1)-(3) should only be used as an example.
The reviewer should determine whether the application demonstrates acceptable ECCS cooling performance following a LOCA:
(i)The ECCS provides sufficient coolant so that the fuel remains in a coolable geometry during and following any heatup and quench.
(ii)The ECCS provides sufficient coolant so that decay heat will be removed for the extended period of time required by the long-lived radioactivity remaining in the fuel.
- 15.
Guidance -
- 3. Realistic, best-estimate analyses of LOCAs at the location of interest as BDBA demonstrate that the consequences would be acceptable.
Demonstration of acceptable consequences for BDBA is not commensurate with the risk.
Once the failure at the location of interest is established as highly unlikely, these accidents should be treated as a severe accident and only dose consequences should be the measure, not fuel integrity or 10 CFR 50.46 acceptance criteria.
Add for beyond-design-basis accidents to the last sentence of the second paragraph of the section:
The reviewer should determine whether the application demonstrates that applicable dose criteria are met for beyond-design-basis accidents.
Comment Number Section Comment/Basis Recommendation
- 16.
Guidance -
- 3. Realistic, best-estimate analyses of LOCAs at the location of interest as BDBA demonstrate that the consequences would be acceptable.
Delete the word "detailed" in the second bullet. The meaning of the word is subjective and therefore does not add value. The word "detailed" is not used in the other bullets regarding the safety analysis report content, so it is not clear what the use means in this bullet.
Remove word detailed
- 17.
Guidance -
- 3. Realistic, best-estimate analyses of LOCAs at the location of interest as BDBA demonstrate that the consequences would be acceptable.
Item 3 states, in part:
Sensitivity studies to demonstrate that uncertainties are acceptably addressed and there are no cliff-edge effects; topics of interest may include the consequences of equipment failures or delayed operator action.
This requirement is contradictory to the allowance to use BDB analysis methods. Such methods do not address any uncertainties and are designed to produce a result which is consistent with the expected operational state. Conservative parameters may be used, but they are for the purpose of simplifying the calculation, not for the purpose of producing a more conservative result.
There are several other places within the ISG where design basis approaches are suggested.
Revise the sentence as follows:
Sensitivity studies to demonstrate that the design space is acceptably addressed; topics of interest may include cliff-edge effects, variability in operator action, and equipment availability.
Other areas where design-based approaches are suggested should be revised.