ML25107A076
| ML25107A076 | |
| Person / Time | |
|---|---|
| Site: | Electric Power Research Institute |
| Issue date: | 04/22/2025 |
| From: | Licensing Processes Branch |
| To: | |
| References | |
| EPID L-2024-TOP-0018 pre-fee, EPID L-2024-NTR-0002 post-fee, 3002028673 | |
| Download: ML25107A076 (6) | |
Text
REGULATORY AUDIT PLAN BY THE OFFICE OF NUCLEAR REACTOR REGULATION TO SUPPORT OF THE REVIEW OF ELECTRIC POWER RESEARCH INSTITUTES TECHNICAL REPORT 3002028673, REVISION 0 LOSS-OF-COOLANT-ACCIDENT (LOCA)-INDUCED FUEL FRAGMENTATION, RELOCATION AND DISPERSAL (FFRD) WITH LEAK-BEFORE-BREAK (LBB) CREDIT -
ALTERNATIVE LICENSING STRATEGY (ALS)
DOCKET NO. 99902021
1.0 BACKGROUND
By letter dated April 26, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24121A203), Electric Power Research Institute (EPRI) submitted (1) EPRI Report 3002028673, Loss-of-Coolant-Accident-Induced Fuel Fragmentation, Relocation, and Dispersal with Leak-Before-Break Credit - Alternative Licensing Strategy
[ALS], (2) EPRI Report 3002028675 (NP)/3002028674 (P), LOCA[Loss of Coolant Accident]
Analysis of Fuel Fragmentation, Relocation, and Dispersal [FFRD] for Westinghouse 2-Loop, 3-Loop and 4-Loop Plants - Proprietary, Evaluation of Cladding Rupture in High Burnup Fuel Rods Susceptible to Fine Fragmentation (collectively called LOCA Analysis of FFRD), and (3) EPRI Report 3002023895, Materials Reliability Program: xLPR [Extremely Low Probability of Rupture] Estimation of PWR Loss-of-Coolant Accident Frequencies (MRP-480) to the U.S.
Nuclear Regulatory Commission (NRC) for review and approval. The ALS for LOCA Analysis for FFRD topical report demonstrates acceptable performance for LOCA induced FFRD phenomena in high burnup pressurized water reactor (PWR) fuel.
By email dated June 25, 2024 (ADAMS Accession No. ML24170A812), the NRC accepted the EPRI topical reports (TRs) for review and approval. The NRC acceptance package includes completion forms and schedules for three TRs, and the withholding determination for the proprietary TR.
The NRC staff has determined that a regulatory audit of the EPRI TR 3002028673 should be conducted in accordance with the Office of Nuclear Regulatory Reactor Office Instruction (OI)
LIC-111, Regulatory Audits, Revision 2, dated December 2024 (ADAMS Accession No. ML24309A281), for the staff to gain a better understanding of EPRIs proposed ALS for LOCA Analysis for FFRD with LBB Credit. Regulatory review activities for EPRI Reports 3002028675 (NP)/3002028674 (P) and 3002023895 are being handled as separate activities.
A regulatory audit is a planned license or regulation-related activity that includes the examination and evaluation of primarily non-docketed information. The audit is conducted with the intent to gain understanding, to verify information, and to identify information that will require docketing to support the basis of a licensing or regulatory decision. Performing a regulatory audit is expected to assist the NRC staff in efficiently conducting its review and gaining insights to the vendors processes and procedures. Information that the NRC staff relies upon to make the safety determination must be submitted on the docket.
2.0 REGULATORY AUDIT BASES The NRC promulgated regulations to permit power reactor licensees and license applicants to implement an alternative regulatory framework with respect to special treatment, where special treatment refers to those requirements that provide increased assurance beyond normal industrial practices that structures, systems, and components perform their design-basis functions. Title 10 of the Code of Federal Regulations Part 50 Section 46 (10 CFR 50.46) provides the regulatory requirements for Emergency Core Cooling System (ECCS) performance for light water reactors. Additional relevant regulatory criteria are found in Appendix A to 10 CFR, General Design Criteria (GDC) for Nuclear Power Plants, including Criterion 4, Environmental and dynamic effects design bases, Criterion 14, Reactor coolant pressure boundary, Criterion 30, Quality of reactor coolant pressure boundary, Criterion 31, Fracture prevention of reactor coolant pressure boundary, Criterion 32, Inspection of reactor coolant pressure boundary, and Criterion 35, Emergency core cooling.
Section 10 CFR 50.46 and related GDCs requirements ensure that a licensed nuclear power plant has a sufficiently capable reactor coolant makeup system to safely handle 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.
3.0 REGULATORY AUDIT SCOPE The NRC staff plans to conduct a virtual audit which will include technical discussion pertaining to potential request for additional information questions as well as comprehensive questions on the TR itself.
4.0 INFORMATION NEEDS The areas of focus for the regulatory audit include information contained in the TR and supplements, the enclosed audit information needs, and all associated and relevant supporting documentation (e.g., methodology, process information, calculations, etc.). The relevant supporting documents are identified in this section.
4.1 Audit Topics 4.1.1 Document Needed for NRC review At the time of writing this audit plan, no specific documents have been identified for NRC review.
In the future, documents may be requested for review and might be necessary based on the results of audit discussions related to Section 4.1.2 of this audit plan.
4.1.2 Areas for Discussion
- 1) Section 1.7 states that the approach chosen for this TR was an LBB-based approach since it would be more expeditiously developed than a risk-informed, RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, approach. Therefore, these analyses are assumed to be deterministic, and the report does not further discuss use of RG 1.174.
However, in Section 2.3.2, the TR states, Although ALS is not a probabilistic risk assessment (PRA) methodology, the above discussion of screening is relevant, supports the use of mean frequency values from NUREG-1829, Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process, and supports assessment of consequences on a risk-informed, not bounding, basis. Therefore, it appears that the consequences are determined through a risk-informed approach, but the analyses are deterministic. Clarify and further explain this perceived contradiction.
- 2) Section 1.9 of EPRIs ALS Report 3002028673 states that the regulatory approach focuses on demonstrating that large break loss-of-coolant-accidents (LB-LOCAs) have an extremely low likelihood of occurrence; therefore, FFRD can be excluded from core cooling analyses of LB-LOCAs. The approach does this by leveraging past leak-before-break (LBB) approvals and recent probabilistic fracture mechanics analyses results for the time between leakage and rupture. NRC policy statement, 54 FR 18649, Policy Statement on Additional Applications of Leak-Before-Break Technology, states that the Commission determined not to modify the regulation to support the use of LBB beyond the requirements of GDC-4 and its application to the removal of dynamic effects from the design basis. Since this TR utilizes the LBB technology beyond that of GDC-4, discuss how this work conforms with that policy and/or if the exemption process, per 54 FR 18649 and 10 CFR 50.12, Specific exemptions, will be used by the licensees attempting to apply this TR to a licensing action.
- 3) Section 4.1.2 states that the LBB technology cannot be applied to pipe systems with active material degradation. However, primary water stress corrosion cracking (PWSCC) has been an active degradation mechanism in the piping systems applicable to ALS.
explain how PWSCC and its mitigation of any kind, including inspections, impact the results in this report.
- 4) Section 4.4.7 states that in some cases plant staff misdiagnosed abnormal leakage trends and returned the reactor to operation only to find the abnormal leakage persisted.
provide additional information on these incidents including how many similar events have occurred, and how the analyses would be impacted if these pipe systems with low leakage rates were near rupture.
- 5) Section 6.2.1 of EPRI TR 3002028673 discusses the transition break sizes (also called maximum break sizes) considered in the ALS and related LOCA-induced cladding rupture analysis. Specifically, Section 6.2.1 describes the maximum break sizes that correspond to the inner diameters of the pressurizer surge line (hot-leg side) and accumulator line (cold-leg side). clarify the following aspects of the maximum break sizes: (a) the nominal pipe sizes and pipe schedules associated with the maximum break sizes; (b) why the residual heat removal piping size is not included in the maximum break sizes; and (c) whether or not the piping connected to the cross-over leg of the main coolant loop, if any, is bounded by the pressurizer surge line and accumulator line in terms of the maximum break size and what is its impact on fuel rod burst.
- 6) Section 5.2 states that, For a power curve fit to the 40-year mean data in Table 5-1, a 27-inch inner diameter pipe yields a maximum mean frequency of about 9x10-8 per calendar year and includes the possibility of larger pipe breaks. From Table 5-1, the mean frequencies for a 14-inch and 31-inch pipe are 4.8x10-7 and 7.5x10-8,respectively.
Confirm how the 9x10-8 was calculated?
- 7) Section 6.1 states that Alternative approaches are employed to assess the vulnerability to a component rupture large enough to potentially cause FFRD. From the sections that follow, it appears that this approach is a qualitative assessment of the vulnerability of these components to failure. Discuss the rationale or basis that qualitative arguments for non-piping components are acceptable where quantitative arguments are given for piping.
- 8) Throughout the document, credit is taken for both piping and non-piping components for the American Society of Mechanical Engineers (ASME) code inservice inspections that typically occur each inspection interval. However, a recent effort by the industry and the ASME code is focused on the optimization of inspection that reduces the number of inspections that occur on these major components each inspection interval.
- a. For the piping systems applicable for ALS, the dissimilar metal welds are well maintained through the ASME Code Case N-770 process, however, the similar metal welds are typically within a risk-informed inspection (RI-ISI) program, where only about 10% of the welds in the same category are inspected (no known degradation). Besides the dissimilar metal welds, discuss the percentage of piping welds in the piping systems applicable to ALS that are currently receiving inspections each inspection interval, i.e., about how many of the similar metal welds applicable for ALS are inspected for each inspection interval.
- b. For the non-piping systems, expand on how the ongoing efforts to optimize the inspection (to about 25 percent of the required inspections) of the steam generator, pressurizer and reactor pressure vessel shell welds impact the conclusions of this effort.
- 9) TR Section 2.3.2 Level of Conservatism includes a brief discussion of the technical specifications (TS) and other criteria that a licensee needs to be considered suitable for the ALS approach. Additional related information is provided in Section 4.4 RCS
[reactor coolant system] Leak Detection and Response. Based on the TS listed, an unidentified leakage of upto 1 gallon per min (gpm) could theoretically exist for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> before being identified during a surveillance. This would equate to more than 4000 gallons. Additional leak detection programs are mentioned in TR Sections 2.3.2, 4.4, and Appendix A, which would greatly reduce the amount of possible leakage, but it is not clear if they are required by a licensee desiring to adopt ALS. Clarify the required leak detection mechanisms and the maximum possible unidentified leakage with consideration for additional required mass balance checks, TS, trip setpoints, etc.
5.0 TEAM AND R EVIEW ASSIGNMENTS The audit team will consist of the following NRC staff:
NAME ASSIGNMENT DIVISION BRANCH Lois James Senior Project Manager Division of Operating Reactor Licensing (DORL)
Licensing Projects Branch (LLPB)
Chris Van Wert Senior Technical Advisor for Reactor Fuels Division of Safety Systems (DSS) n/a Brandon Wise Reactor Systems Engineer DSS Nuclear Methods and Fuel Analysis (SFNB)
Seung Min Materials Engineer Division of New and Renewed Licenses (DNRL)
Piping and Head Penetrations Branch (PHPB)
David Rudland Senior Level Advisor for Nuclear Power Plant Materials DNRL n/a David Dijamco Materials Engineer DNRL Vessels and internals Branch (NVIB)
Eric Palmer Materials Engineer DNRL NVIB Gordon Curran Safety and Plant Systems Engineer DSS Containment and Plant Systems Branch (SCPB)
John Lehning Senior Nuclear Engineer DSS SFNB Matthew Leech Reliability and Risk Analysis Division of Risk Assessment (DRA)
PRA Oversight Branch (APOB) 6.0 LOGISTICS The audit will be conducted from May 2 through June 27, 2025, via an online portal (also known as electronic portal, ePortal, or electronic reading room) established by EPRI.
The audit team will conduct a Microsoft Teams-based entrance meeting with the vendor on May 2, 2025, for the purposes of introducing the team, discussing the scope of the audit, and describing the information to be made available on the portal. During the audit period between May 2 through June 27, 2025, the NRC staff will hold breakout sessions with representatives of EPRI to answer audit team questions and to have technical discussions. An exit meeting/call will be held at the conclusion of the audit on June 27, 2025.
At this time, the NRC staff does not foresee the need for an onsite visit or in-person discussions between the NRC and vendor staff to discuss information to be provided on the portal. However, if the need for a such a meeting is identified in the future, the audit plan will be revised, and the schedule for the audit will be adjusted accordingly. The NRC project manager (PM) will coordinate any changes to the audit schedule and the location with the vendor.
7.0 SPECIAL REQUESTS While the audit team currently has not identified documents for audit review, it is likely that documents will be requested as a part of the discussion items in Section 5.0 are addressed. The audit team would like access to these documents through an online portal when they are identified to allow the audit team to access the documents via the internet. The following conditions associated with the online portal must be maintained throughout the duration of the audit so that the audit team has access to the online portal:
The online portal will be password-protected, and separate passwords will be assigned to the NRC staff who are participating in the audit.
The online portal will be sufficiently secure to prevent the NRC staff from printing, saving, downloading, or collecting any information on the online portal.
Conditions of the use of the online portal will be displayed on the login screen and will require acknowledgement by each user.
Username and password information should be provided directly from the vendor to the NRC staff. The NRC PM will transmit the names and contact information of the NRC staff who will be participating in the audit directly to the vendor. All other communications should be coordinated through the NRC PM.
8.0 DELIVERABLES An audit summary report will be prepared approximately 90 days after the completion of the audit. If the NRC staff identifies information during the audit that is needed to support its regulatory decision, the NRC staff will issue RAI(s) to the vendor.