ML25142A151
| ML25142A151 | |
| Person / Time | |
|---|---|
| Site: | Electric Power Research Institute |
| Issue date: | 05/23/2025 |
| From: | Delosreyes J Licensing Processes Branch |
| To: | Kucuk A Electric Power Research Institute |
| Shared Package | |
| ML25142A157 | List: |
| References | |
| Download: ML25142A151 (5) | |
Text
1 REQUEST FOR ADDITIONAL INFORMATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REVIEW OF (3002023895) EPRI MRP-480, XLPR ESTIMATION OF PWR LOCA FREQUENCIES (NON-BILLABLE)
EPRI SDA DOCKET NO. 99902021 ISSUE DATE: 05/23/2025
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Background===
By application dated April 26, 2024 (Agencywide Documents Access and Management System(ADAMS) Accession No. ML24121A204), Electric Power Research Institute, Inc. (EPRI),
submitted technical report EPRI Report 3002023895, Materials Reliability Program: xLPR
[Extremely Low Probability of Rupture] Estimation of PWR Loss-of-Coolant Accident Frequencies (MRP-480), (ML24123A223) for U.S. Nuclear Regulatory Commission (NRC) review and approval. One of the objectives of MRP-480 is to demonstrate that analytically determined loss-of-coolant accident (LOCA) frequencies in pressurized water reactors (PWRs) using the probabilistic fracture mechanics (PFM) code, Extremely Low Probability of Rupture (xLPR), are similar to those presented in NUREG-1829, Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process, Vol. 1 (ML080630013). To complete its review, the NRC staff requests additional information.
Regulatory Basis Title 10 of the Code of Federal Regulations Part50 (10 CFR 50), Domestic Licensing of Production and Utilization Facilities, AppendixA, General Design Criteria for Nuclear Power Plants, Criterion 4, Environmental and dynamic effects design bases, specifies, in part, that when analyses approved by the NRC demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping, dynamic effects associated with postulated pipe ruptures may be excluded from the design basis.
The alternative licensing strategy (ALS) under which MRP-480 is implemented leverages the approved leak-before-break (LBB) analyses for the reactor coolant system main loop piping of pressurized water reactors (PWRs) that were performed to meet 10 CFR 50, Appendix A, Criterion 4. T he PFM analyses in MRP-480 are used to confirm and/or supplement the approved LBB analyses to further demonstrate that the likelihood of rupture under a LOCA is extremely low. The PFM analyses in MRP-480 are performed in the context of LBB behavior in Alloy 82/182 dissimilar metal welds (DMWs), in which an active degradation mechanism of primary water stress corrosion cracking (PWSCC) is present. Therefore, the regulatory basis for the following requests for additional information (RAIs) is related to demonstrating that the 10 CFR 50, Appendix A, Criterion 4 of piping rupture is extremely low under conditions consistent with the design basis for the piping is met for the main reactor coolant loop piping of PWRs with an active degradation mechanism.
RAI 1
Issue Sections 3.1 and 3.2 of MRP-480 discuss seismic effects analyzed in the xLPR Piping System Analysis report (Reference 8 of MRP-480, ML21217A088) and xLPR Generalization Study
2 report (Reference 9 of MRP-480, ML22088A006). Appendix B of these two reports shows the seismic stresses used in the analyses of the piping relevant to ALS (i.e., the main reactor coolant loop piping). During the audit, EPRI clarified the sources of these seismic stresses, and the NRC staff noted these sources in the audit report. Since the intent for MRP-480 is for generic use by a plant to justify the extremely low probability of a LOCA, the NRC staff noted that bounding seismic stresses are needed for the PFM analyses in MRP-480 of the main reactor coolant primary loop piping. However, it is not clear, based on Sections 3.1 and 3.2 of MRP-480 and the sources of the seismic stresses, whether bounding seismic stresses were used in the PFM analyses in MRP-480. It is also not clear whether the seismic stresses reflect the latest seismic hazard curves, as documented, for example, in NUREG/KM-0017, Seismic Hazard Evaluations for U.S. Nuclear Power Plants: Near-Term Task Force Recommendation 2.1 Results, December 2021 (ML21344A126).
The NRC staff also noted that MRP-480 relies on the PFM analyses in the xLPR Piping System Analysis report and xLPR Generalization Study report. These PFM analyses are used to demonstrate that the probability of rupture in piping systems with an active degradation mechanism remains extremely low, consistent with the requirement in 10 CFR 50, Appendix A, General Design Criteria 4. In the context of risk-informed decision making (RIDM), PFM analyses work in tandem with other RIDM principles, particularly the performance monitoring principle, when used as the basis in regulatory and safety decisions. For the reactor coolant loop piping, inservice inspection is one of the primary means of performance monitoring. The NRC staff noted that consideration of an adequate inservice inspection sample of the welds in the main reactor coolant primary loop piping within the scope of MRP-480 could address some of the concerns on seismic effects discussed above. However, MRP-480 does not include a discussion of the ongoing inservice inspection programs for the main reactor coolant primary loop piping within the scope of MRP-480.
Request
- a. Clarify whether bounding seismic stresses were used in the PFM analyses in MRP-480 of the main reactor coolant primary loop piping, and whether they reflect the latest seismic hazard curves.
- b. Clarify or summarize the performance monitoring programs (e.g., ongoing inservice inspection programs) for the welds in the main reactor coolant primary loop piping within the scope of MRP-480, which include DMWs and similar metal welds (both stainless steel welds and ferritic steel welds). In this clarification or summary, discuss whether these programs are sufficient for performance monitoring to support the PFM analyses in MRP-480 and how these programs address seismic effects.
RAI 2
Issue Section 1.1 of MRP-480, last paragraph, states:
While the ALS is an immediate driver for the investigation into NUREG-1829 LOCA frequency results and time between detectable leakage and LOCA, it should be noted that the results herein are intended to be generic and of use to other projects.
The NRC staff noted that ALS leverages the approved LBB analyses for the main reactor coolant loop piping of PWRs, and that the PFM analyses in MRP-480 are used to confirm and/or
3 supplement the approved LBB analyses to show that the likelihood of a LOCA is extremely low, and to demonstrate the time between leakage and rupture is sufficient to safely shutdown the plant before rupture occurs. Since MRP-480 was submitted for review and approval under the overall umbrella of ALS, the NRC staffs review of MRP-480 is based on the reports use for supporting ALS. Generally speaking, any topical report and the NRC staffs review of it are based on a clear tie to a specific technical and/or regulatory issue that the report is addressing.
The specific issue addressed in ALS and MRP-480 is the potential for fuel fragmentation, relocation, and dispersal (FFRD) in high burnup fuel during design basis accidents. Thus, uses of MRP-480 outside the scope of ALS (i.e., outside the evaluation of potential for FFRD and its impact) would need to be reviewed and approved by the NRC staff on a case-by-case basis.
Request In the last paragraph of Section 1.1 of MRP-480, clarify that uses of MRP-480 outside the scope of ALS will be submitted to the NRC for review and approval.
RAI 3
Issue The NRC staff noted that Table 3-3 of MRP-480 includes three cases for Combustion Engineering (CE) and Babcock & Wilcox (B&W) designs, which have ferritic steel piping (base and weld metal) in the main reactor coolant loop. Section 5 of MRP-480 assesses degradation mechanisms in stainless steels and nickel-based alloys, but does not include an assessment of degradation mechanisms in ferritic steels.
Request Discuss how the results of the xLPR analyses in MRP-480, which only include the relevant degradation mechanism in stainless steels and nickel-based alloys, bound the results for ferritic steel (base and weld metal) piping systems without these materials.
RAI 4
Issue The NRC staff noted that the base cases and many of the sensitivity cases listed in Table 3-3 of MRP-480 use the Direct Model 1 (DM1) PWSCC initiation model in xLPR. The report does not clearly describe DM1, including conservatisms in the model, the datasets (i.e., field data or lab data) to which the model was calibrated, and the sensitivity studies that were focused on addressing model uncertainty in DM1. The NRC staff also noted that xLPR Technical Report, Primary Water Stress-Corrosion Cracking Initiation Model Parameter Development, Confirmatory Analyses, and Validation, 2017 (ML19337C202) contains information on DM1 that is not described in MRP-480.
Request Describe DM1 including conservatisms in the model, the datasets (i.e., field data or lab data) to which the model was calibrated, and the sensitivity studies that were focused on addressing the model uncertainty in DM1.
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RAI 5
Issue Section 3.1 of MRP-480 states:
A single loop analysis provides a more efficient way to reach statistical convergence of mean values. Approximately 100,000 realizations were executed per case explicitly modeling crack initiation and approximately 5,000 realizations were executed per case utilizing the initial flaw model, as these were the number of realizations respectively that were estimated to be necessary to guarantee that any undesirable event would not be missed in the analysis.
Also, Table 4-10 of MRP-480 states there was no direct acceptance criteria for convergence of the results of the PFM analyses in MRP-480. Other than these two places in MRP-480, there is no other discussion of convergence, and therefore, it is not clear to the NRC staff how it was determined that convergence of the PFM results of the cases analyzed in MRP-480 was reached. The NRC staff noted that Reference 16 of MRP-480, NUREG/CR-7278, Technical Basis for the use of Probabilistic Fracture Mechanics in Regulatory Applications, January 2022 (ML22014A406), contains guidance on convergence discussed during the audit and noted in the audit report that is not described in MRP-480.
Request Explain how it was determined that convergence of the PFM results of the cases analyzed in MRP-480 was reached with the single loop analysis approach cited above from Section 3.1 of MRP-480 and considering the guidance in NUREG/CR-7278 on convergence.
RAI 6
Issue Section 4.1.5 of MRP-480 discusses the use of rupture frequency as an analogue for large break LOCA frequency and explains two of the cases analyzed in MRP-480 to illustrate this approach. It is not clear how only two cases out of the many cases in MRP-480 are sufficient to demonstrate that rupture frequency is an approximate analogue for large break LOCA frequency. Also, the assumption in Section 4.1.5 of MRP-480 seems to be that every realization of rupture has a corresponding realization of large break LOCA. The NRC staff noted that not every realization of large break LOCA would necessarily lead to rupture, in which case the large break LOCA frequency could actually be higher than rupture frequency.
Request Further explain how the cases analyzed in MRP-480 demonstrate that the approach of using rupture frequency as an approximate analogue for large break LOCA frequency is adequate.
OFFICE NRR/DORL/LLPB/PM NRR/DORL/LLPB/BC NRR/DNRL/NVIB/BC NAME JDelosreyes JRankin ABuford
5 DATE 05/19/2025 05/19/2025 05/19/2025