ML25043A335
| ML25043A335 | |
| Person / Time | |
|---|---|
| Issue date: | 11/17/2025 |
| From: | Hayden T NRC/NRR/DNRL/NRLB |
| To: | |
| Shared Package | |
| ML25157A026 | List: |
| References | |
| DSS-ISG-2025-XX | |
| Download: ML25043A335 (8) | |
Text
DSS-ISG-2025-XX Treatment of Certain Loss-of-Coolant Accident Locations as Beyond-Design-Basis Accidents Draft Interim Staff Guidance November 2025
DSS-ISG-2025-XX Treatment of Certain Loss-of-Coolant Accident Locations as Beyond-Design-Basis Accidents Draft Interim Staff Guidance ADAMS Accession Nos.: Package - ML25157A026; ISG - ML25043A335; FRN - ML25157A032 OFFICE NRR/DSS NRR/DEX NRR/DEX/EIMB NRR/DNRL/NRLB NAME TClark KManoly SBailey MJardaneh DATE 2/13/2025 2/13/2025 2/26/2025 2/25/2025 OFFICE NRR/DNRL NRR/DEX NRR/DNRL OGC/LRAA/RASFP NAME DRudland TMartinez-Navedo MSampson SClark for JBielecki DATE 2/26/2025 2/27/2025 2/28/2025 6/26/2025 OFFICE OGC/LRAA/RASFP NRR/DSS NAME SClark VCusumano DATE 6/26/2025 6/30/2025 OFFICIAL RECORDS COPY
DRAFT INTERIM STAFF GUIDANCE TREATMENT OF CERTAIN LOSS-OF-COOLANT ACCIDENT LOCATIONS AS BEYOND-DESIGN-BASIS ACCIDENTS DSS-ISG-2025-XX PURPOSE The U.S. Nuclear Regulatory Commission (NRC, or Commission) staff is providing this interim staff guidance (ISG) to communicate the key safety principles that would enable the NRC staff to determine that certain break locations that would normally be analyzed as design-basis loss-of-coolant accidents (LOCAs) for light-water reactors can be treated as beyond-design-basis accidents. More detailed guidance will be developed as the need arises, and in consideration of public comment on this ISG and a related rulemaking.
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. Mechanistic (i.e., based on physical processes or phenomena) rationales for determining that certain LOCAs are unlikely to occur have generally not been accepted.
The NRC, however, has accepted mechanistic rationales for dispositioning certain phenomena for limited purposes. For example, the dynamic effects of pipe ruptures can be excluded from consideration in the design bases under 10 CFR Part 50, Appendix A, General Design Criteria
[(GDC)] for Nuclear Power Plants, GDC 4 if certain conditions are met. Specifically, the NRC needs to review and approve analyses that demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.
The determination that the probability of pipe ruptures is extremely low under GDC 4 is only for the analysis of dynamic effects and does not apply to the design-basis LOCA spectrum used to calculate ECCS or containment performance, among other aspects of structures, systems and components design. The NRC has nonetheless begun considering other aspects of reactor design for which engineering analysis methods have developed to a point that mechanistic considerations may be employed to exclude some LOCAs from the design-basis while continuing to maintain high level of probability that the emergency core cooling function will be accomplished. Other design-basis analyses that depend on the results of ECCS analyses may also be affected by this approach. Further, the NRC has begun rulemaking efforts to apply relaxed analytical methods to certain classes of LOCAs.
The NRC is currently considering circumstances under which an alternative interpretation of the design-basis LOCA spectrum may be found to be acceptable. For some applications now under review and anticipated to be submitted in the near to medium term, designers have sought to holistically reduce LOCA risks (e.g., reduced numbers of penetrations, larger volumes of water above the core, extended coping times, passive cooling systems). In consideration of design-
DSS-ISG-2025-XX Page 2 of 6 specific information, the NRC can review justifications that design-basis LOCAs need not be postulated at all conceivable locations.
This draft guidance describes the mechanistic considerations that the NRC staff may consider in determining whether an applicant has proposed an adequately protective design-basis LOCA spectrum.
RATIONALE The staff intends to use mechanistic information on the likelihood and realistic consequences of LOCAs at specific locations to determine whether to make a safety finding on a basis other than the calculated design-basis ECCS cooling performance under 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. This approach would be approved only when a beyond-design-basis accident analysis demonstrates acceptable defense in depth and safety margins.
The NRC staff plans to review applications in accordance with the NRC interpretation in this draft guidance (once it becomes effective) 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.1 The staff plans to employ the framework outlined in this ISG (once it becomes effective after public comment) until the Commission approves a long-term resolution of the associated technical issues, potentially through rulemaking. This ISG only applies to specific locations for which applicants have provided the technical justifications necessary to provide reasonable assurance of adequate protection of public health and safety. It is not intended for generic or permanent applicability to all parts of the spectrum of LOCAs (e.g., it does not apply to breaks larger than a certain size or with a specified low frequency.) This phased approach provides a rapid path toward regulatory certainty that does not disadvantage early adopters of these approaches.
This draft ISG outlines mechanistic considerations that are important in justifying that certain break locations can be treated as beyond-design-basis LOCAs, as well as some examples that could help illustrate how those considerations might be employed for a specific application. It does not prescribe a necessary and sufficient set of considerations that an applicant should specify, because the staffs decisions on these matters are highly design-specific. The staff would withdraw this guidance and its examples once more specific requirements and guidance are established.
APPLICABILITY All holders of and applicants for a construction permit or operating license for a light-water nuclear power reactor under 10 CFR Part 50, Domestic Licensing of Production and Utilization 1 If excluded from the design basis for the purposes of calculating ECCS cooling performance under 10 CFR 50.46, a LOCA would also be excluded from the design-basis for the purposes of all other requirements that consider the consequences of LOCAs. Examples of these requirements include 10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants, and GDCs 19, 35, 38, 41, 44, and 50. None of these NRC regulations replicate the prescriptive requirements in 10 CFR 50.46 that define the spectrum of LOCAs that must analyzed.
DSS-ISG-2025-XX Page 3 of 6 Facilities, except those who have permanently ceased operations and have certified that fuel has been permanently removed from the reactor vessel.
All holders of and applicants for a light-water power reactor combined license, standard design approval, or manufacturing license under 10CFR Part52, Licenses, Certifications, and Approvals for Nuclear Power Plants. All applicants for a light-water reactor standard design certification, including such applicants after initial issuance of a design certification rule.
GUIDANCE This ISG describes a framework for the mechanistic considerations that the staff could consider in determining whether an application has justified certain break locations as beyond-design basis-LOCAs. Additional detailed guidance, including acceptance criteria, will be developed as needed. As discussed above, the NRC does not intend to apply this guidance broadly to consideration of mechanistic factors in the treatment of LOCAs in general, including partitioning breaks in a generic sense by size.
To apply this approach to a given location, the staff would consider whether the application includes sufficient information to form a basis for the NRC to make a safety determination that certain break locations in the reactor coolant system can be analyzed as beyond-design-basis accidents rather than as part of the design-basis LOCA spectrum under 10 CFR 50.46. The reviewer should consider the safety significance of the break location in an integrated manner, addressing likelihood, consequences, margins, defense in depth, and performance monitoring.
The reviewer should determine whether the application demonstrates through realistic analyses that there is defense-in-depth and safety margin for breaks at these locations sufficient to provide reasonable assurance of adequate protection. To be acceptable under this guidance (and not require an exemption from LOCA evaluation model requirements), the reviewer should determine whether this demonstration sufficiently addresses mechanistic considerations, as outlined below, regarding the likelihood and consequences of the break of interest.
DSS-ISG-2025-XX Page 4 of 6
- 1.
The design implements a holistic safety approach that reduces LOCA risk through both prevention and mitigation.
This provision is intended to focus application of this criterion while general requirements for beyond-design-basis LOCAs are established. The reviewer should determine whether the application provides information sufficient to justify that the proposed design reduces the overall risk of LOCAs through specific features or design considerations.
While these considerations are highly design-specific, examples include:
reduced numbers of penetrations elimination of penetrations at lower elevations in the reactor vessel larger volumes of water above the core low pressure passive injection sources slower accident progression that provides more time (e.g., hours) for operators to respond effectively focused inspection accessibility for risk-significant locations, using qualified techniques
- 2.
Design and operational programs provide assurance that failures at the location of interest are highly unlikely.
While the specific assurances will depend on the location of interest (e.g., welded location, bolted location, component), example means of providing this assurance include:
confirmation that the component(s) of interest are designed to American Society of Mechanical Engineers (ASME)Section III, with identification of design features to demonstrate absence of known degradation mechanisms (e.g., ASME Section III Nonmandatory Appendix W) that may contribute to the components function during long term operation.
local2 inservice inspection consistent with the methods in the ASME BPV Code and local leakage detection that would demonstrate that, should a mode of degradation develop, it will be identified and addressed as early as possible mitigation strategies for identified flaws, as well as sufficient description of how the applicant would assess its prior conclusion of highly unlikely failures at the location demonstration of low probability of component(s) failure, through either qualitative or quantitative methods. If quantitative analyses are used, fracture mechanics analyses (deterministic or probabilistic) and appropriate acceptance criteria for the welds can be used to demonstrate adequate hypothetical margin to failure should a flaw exist during service. These analyses should also account for non-piping impacts and indirect piping failures (see NUREG-18293 and NUREG-19034 for NRC experience with analyses that demonstrate failures of certain size pipes have a low probability of occurrence) 2 In this context, local refers to inspection and leakage detection at the component(s) of interest location.
3 NUREG-1829, Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process, dated April 2008.
4 NUREG-1903, Seismic Considerations For The Transition Break Size, dated February 2008.
DSS-ISG-2025-XX Page 5 of 6
- 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 reviewer should consider the acceptability of analysis methods that deviate from those specified in 10 CFR 50.46. Specifically, staff conclusions may rest, in part, on realistic, best-estimate analyses without conservative biases and uncertainties, consistent with evaluation of other beyond-design-basis accidents. In addition, these analyses may credit the operation of non-safety-related systems and operator actions, where appropriately justified.
The reviewer should determine whether the application demonstrates that the fuel remains within acceptable design limits for all LOCAs, regardless of their status as DBAs or beyond-design-basis accidents. 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 beyond-design-basis (realistic) evaluation methodology results may also be compared to the 10 CFR 50.46(b) acceptance criteria.
In addition, the reviewer should determine whether the application demonstrates that applicable dose criteria are met.
The reviewer should determine whether the safety analysis report describes:
The method that was used to assess the consequences of breaks at beyond-design-basis LOCA locations, including a demonstration that the analytical techniques realistically account for the neutronic and thermal-hydraulic response (including appropriate conservatisms the applicant may elect to apply)
The detailed sequence of events of the accident progression for each location, including system response and any credited operator actions to mitigate the consequences, recover and restore core inventory, and remove long-term decay heat (some of which may have been developed as part of the design probabilistic risk assessment)
Assumptions and criteria necessary for the analysis to be valid, including operational attributes, processes, and operator actions Design and operational assumptions on which the analysis is based, which the staff may use to ensure the mitigation capabilities are available and associated analytical conclusions remain valid 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 Discussion and analysis of potential dose consequences of the event IMPLEMENTATION Based upon the considerations above, the staff could consider information in the application sufficient to demonstrate adequate defense-in-depth and safety margins for the breaks in the RCPB for LOCAs that would be considered to be beyond the design-basis. The staff may base its findings on this information for new reactor license applications with novel design-specific circumstances. For operating reactors, the NRC views requirements now under development in regard to disposition of LOCAs based on size as more suitable to address beyond-design-basis LOCAs for some designs.
DSS-ISG-2025-XX Page 6 of 6 BACKFITTING AND ISSUE FINALITY DISCUSSION The NRC staff may use the ISG as a reference in its regulatory processes, such as licensing, inspection, or enforcement. However, the NRC staff does not intend to use the guidance in this ISG to support NRC staff actions in a manner that would constitute backfitting as that term is defined in 10 CFR 50.109, Backfitting, and as described in NRC Management Directive 8.4, Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests (ML18093B087), nor does the NRC staff intend to use the guidance to affect the issue finality of an approval under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants. The staff also does not intend to use the guidance to support NRC staff actions in a manner that constitutes forward fitting as that term is defined and described in Management Directive 8.4. If a licensee believes that the NRC is using this ISG in a manner inconsistent with the discussion in this paragraph, then the licensee may file a backfitting or forward fitting appeal with the NRC in accordance with the process in Management Directive 8.4.
CONGRESSIONAL REVIEW ACT Discussion to be provided in final ISG.
FINAL RESOLUTION This guidance, once it becomes effective after public comment, would be implemented until the Commission establishes more specific requirements and guidance. The staff plans to withdraw this guidance, and its examples, once more specific requirements and guidance are established. Backfitting implications of the withdrawal of this interim position will be assessed, as appropriate, in conjunction with the establishment of more specific requirements and guidance.