ML25105A284

Enclosure 1: LOCA Analysis of FFRD for Westinghouse 2-Loop, 3-Loop and 4-Loop Plants Audit Report Nonproprietary (1)

ML25105A284
Person / Time
Site:	Electric Power Research Institute
Issue date:	04/29/2025
From:	Delosreyes J Licensing Processes Branch
To:
References
EPID L-2024-NTR-0003 EPRI 3002028675 (NP), EPRI 3002028674 (P)
Download: ML25105A284 (1)

Text

OFFICIAL USE ONLYPROPRIETARY INFORMATION OFFICIAL USE ONLYPROPRIETARY INFORMATION REGULATORY AUDIT REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION IN SUPPORT OF THE REVIEW OF EPRI REPORT 3002028675 (NP)/3002028674 (P), LOCA

[LOSS-OF-COOLANT-ACCIDENT] ANALYSIS OF FUEL FRAGMENTATION, RELOCATION, AND DISPERSAL FOR WESTINGHOUSE 2-LOOP, 3-LOOP, AND 4-LOOP PLANTS -

PROPRIETARY, EVALUATION OF CLADDING RUPTURE IN HIGH BURNUP FUEL RODS SUSCEPTIBLE TO FINE FRAGMENTATION, REVISION 0 ELECTRIC POWER RESEARCH INSTITUTE 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 [Loss-of-Coolant-Accident]-Induced Fuel Fragmentation, Relocation, and Dispersal [(FFRD)] with Leak-Before-Break Credit -

Alternative Licensing Strategy [ALS], (2) EPRI Reports 3002028675 (NP)/3002028674 (P),

LOCA Analysis of Fuel Fragmentation, Relocation, and Dispersal 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 [Pressurized Water Reactor] 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 (TR) demonstrates an alternative performance for LOCA-induced FFRD phenomena in high burnup PWR fuel.

By email dated June 25, 2024 (ADAMS Accession No. ML24170A812), the NRC accepted the EPRI TRs for review and approval. This audit report deals with EPRI Reports 3002028675 (NP)/

3002028674 (P), LOCA Analysis of FFRD for Westinghouse 2-Loop, 3-Loop, and 4-Loop plants.

The NRC staff determined that a regulatory audit was needed to increase the efficiency, facilitate discussion, and clarify issues identified during the NRC staffs initial review and conducted a virtual regulatory audit beginning with the entrance meeting on November 26, 2024, until the exit meeting on January 31, 2025, based on the audit plan issued on November 21, 2024 and revised on February 3, 2025 (ADAMS Package Accession No. ML25014A309). The audit was held in accordance with the NRC Office of Nuclear Reactor

OFFICIAL USE ONLYPROPRIETARY INFORMATION OFFICIAL USE ONLYPROPRIETARY INFORMATION Regulation procedure as described in LIC-111, Regulatory Audits, and under the guidance provided in LIC-500, Revision 9, Topical Report Process. The audit was closed due to the proprietary nature of the information discussed. The information discussed during the audit was determined to be proprietary by the NRC staff. Based on the results of the audit, the NRC issued its request for additional information (RAI) via email dated February 26, 2025 (ADAMS Accession No. ML25049A049).

2.0 REGULATORY AUDIT BASES The current NRC regulations at Title 10 of the Code of Federal Regulation (10 CFR)

Section 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, limits peak cladding temperature and maximum cladding oxidation to satisfy General Design Criterion (GDC) 35, Emergency core cooling, of Appendix A, General Design Criteria for Nuclear Power Plants, to Part 50, Domestic licensing of production and utilization facilities, which requires, in part, that The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such thatclad metal-water reaction is limited to negligible amounts.

In other words, FFRD can challenge core coolability, so the simplest approach to demonstrate that GDC 35 is met is to demonstrate that no fuel dispersal occurs.

Nuclear power plant (NPP) licensees have informed the NRC that NPPs are considering the use of higher burnup fuel designs to meet operational objectives. However, FFRD is a technical challenge to the use of higher burnup fuel designs. EPRI has developed an alternative licensing approach for utilization of higher burnup fuel rods in PWRs which has been submitted to the NRC for review and approval (ADAMS Accession No. ML24121A203).

EPRI Report 3002028675 (NP)/3002028674 (P) presents the results of cladding rupture calculations for small-break and intermediate-break LOCA, which indicate that cladding rupture would likely not occur in high burnup fuel rods susceptible to fine fragmentation for specific Westinghouse fuel designs utilized in 2-Loop, 3-Loop, and 4-Loop Westinghouse PWRs. If cladding rupture would not occur, finely fragmented fuel would not disperse from the high burnup fuel rods into the reactor coolant during a LOCA.

3.0 AUDIT DISCUSSION Topics discussed during the audit and in this report included proprietary information. Proprietary information is identified using a (( yellow highlighted, bolded, double bracketed )) text format.

Consistent with the Commission proprietary withholding criteria in 10 CFR 2.390, Public inspections, exemptions, requests for withholding, the staff is protecting the identified proprietary material from public disclosure. For the redacted publicly available version of the audit report, the protected material will be deleted from the bolded brackets (i.e., designated by a blank (( )) bracketed format).

Each of the items below correspond to the discussion topics described in the audit plan (1-14).

Additional discussion topics were identified during the audit and were not described in the audit plan. These topics will be described in additional detail below (15-16).

OFFICIAL USE ONLYPROPRIETARY INFORMATION OFFICIAL USE ONLYPROPRIETARY INFORMATION

1. The NRC staff discussed the limiting break size in the intermediate-break size region. It was decided that the response to this question is most appropriately provided through an existing RAI issued in the concurrent review WCAP-18850 as the question pertains to the base methodology.

2. The NRC staff discussed the threshold temperature for cladding metal-water reaction.

Westinghouse noted that ((

)) The NRC staff decided that this approach was consistent with the as-submitted WCAP-18850 methodology and decided that no RAI was necessary.

3. The NRC staff discussed the limiting treatment of steam generator tube plugging (SGTP). In particular, the staff was interested if the SGTP levels used in the submitted evaluations were limiting. Westinghouse noted that part of the justification for the treatment of SGTP is provided in response to the RAI responses in WCAP-18850. The NRC staff issued an additional RAI, as part of this review, to inquire about

((

))

4. The NRC staff discussed the base composite model that was generated before applying conservatisms related to high-ranked PIRT parameters from WCAP-16696-P-A and the treatment of medium or low-ranked parameters. Westinghouse noted that the composite models are based on existing plant designs, such that medium and low-ranked parameters are nominal values. The NRC staff issued an RAI for Westinghouse to provide some additional detail to discuss how the composite models were generated.

5. The NRC staff asked about the treatment of the hot leg downcomer gap flow treatment in WCAP-18850P-A. Westinghouse confirmed that the treatment is consistent with WCAP-16696-P-A and the limitations and conditions therein. The NRC staff determined that no RAI was necessary.

6. The NRC staff asked about several parameters where Westinghouse noted that the type of steam generator was important. Westinghouse provided details related to the relevant steam generator parameters and how they are conservatively modeled. The NRC staff determined that no RAI was necessary.

7. The NRC staff asked about the impact of certain geometric parameters that were not conservatively modeled but may have a negative impact on the evaluation. Specifically, the NRC staff discussed cross-over leg piping geometry. Westinghouse noted that there was little variation in this geometry such that a nominal modeling of certain piping dimensions would be appropriate. The NRC staff determined that no RAI was necessary.

8. The NRC staff discussed a parameter that was ranked as high in WCAP-16696-P-A but not conservatively dispositioned in this TR. Westinghouse noted that this particular parameter is not expected to be of any significance for small and intermediate-break sizes. The NRC staff decided to issue an RAI for Westinghouse to provide justification for omitting the identified parameter.

9. The NRC requested clarification regarding the treatment for fuel rod and assembly average burnup. Westinghouse confirmed that the treatment of burnup is consistent with WCAP-16696-P-A and the as-submitted WCAP-18850 methodologies. The NRC staff decided to issue an RAI to clarify some ambiguous language related to the treatment of fuel burnup.

10. The NRC staff asked about the relation between Tables 3.5-1 through 3.5-3 and the evaluations in Section 4.0 of this TR. Westinghouse clarified the two-step methodology for LOCA evaluations and why there is a discrepancy between these sections. The NRC staff decided that no RAI was necessary.

OFFICIAL USE ONLYPROPRIETARY INFORMATION OFFICIAL USE ONLYPROPRIETARY INFORMATION

11. The NRC staff discussed the impact to this TR due to the concurrent review of WCAP-18850 and noted that any changes to the as-submitted WCAP-18850 TR should be addressed in this TR. The NRC staff will issue an RAI related to this item at a later date.

12. The NRC asked about the implementation of this TR to confirm the staffs understanding. EPRI confirmed that the NRC staffs understanding was correct and referenced the implementation requirements described in the TR. The NRC staff decided that no RAI was necessary.

13. The NRC staff asked some questions related to various parameters considered in the composite model design. These parameters included break size, fuel pellet design, burnable absorber loading, and fuel enrichment to ensure the limiting treatments were being applied. The NRC staff decided to issue an RAI for Westinghouse to describe how the fuel design parameters incorporated into the composite design are bounding.

14. This topic was discussed under other items. The NRC staff did not hold any discussions under this topic.

15. The NRC staff asked about the applicability of this TR to CE-designed PWRs.

Westinghouse noted that that this TR is not applicable to CE-designed PWRs, only Westinghouse-designed 2-, 3-, and 4-loop PWRS. Applicability to CE-designed PWRs would require a supplement or additional TR.

16. The NRC staff inquired about the conservative direction of parameters. The staff was particularly interested if any directions of conservatism for certain parameters would flip after applying conservative design changes related to high-ranked parameters.

Westinghouse noted that the directions of conservatism are unlikely to change for the most important parameters. Less important parameters or phenomena could flip in their conservative direction, but the overall impact to the calculated PCT would be small compared to the proposed conservatisms in the composite model. Overall, the directions of conservatism of first-order parameters are well understood. The NRC staff decided that no RAI was necessary.

17. The NRC staff asked about the basis of the maximum break sizes described in Table 1-1 of the TR (EPRI Report 3002028674/5). Westinghouse confirmed that the maximum break sizes considered in the TR represent the largest break sizes of the reactor coolant pressure boundary piping excluding the hot-leg, cold-leg and crossover-leg piping at the PWRs within the scope of the TR (i.e., Westinghouse-designed 2-loop, 3-loop and 4-loop PWRs). The NRC staff decided that no RAI was necessary.

The NRC staff reviewed the following documents provided by EPRI and Westinghouse during the audit.

List of Documents Review by the NRC Staff

((

))

((

))

((

))

((

))

((

))

((

))

((

))

((

))

((

))

((

))

((

))

OFFICIAL USE ONLYPROPRIETARY INFORMATION OFFICIAL USE ONLYPROPRIETARY INFORMATION List of Documents Review by the NRC Staff

((

))

((

))

((

))

((

))

((

))

4.0 TEAM AND REVIEW ASSIGNMENTS The audit team consisted of the following NRC staff:

NAME ASSIGNMENT DIVISION BRANCH Noushin Amini Nuclear Engineer Division of Safety Systems (DSS)

Nuclear Methods and Fuel Analysis Branch (SFNB)

James Delosreyes Project Manager Division of Operating Reactor Licensing (DORL)

Licensing Projects Branch (LPLB)

Gerond George Branch Chief DORL LLPB Lois James Senior Project Manager DORL LLPB Matthew Leech Reliability and Risk Analyst Division of Risk Assessment PRA Oversight Branch John Lehning Project Manager DORL LLPB Ekaterina Lenning Project Manager DORL LLPB Seung Min Materials Engineer Division of New and Renewed Licenses Piping and Head Penetrations Branch Demetrius Murray Project Manager DORL LLPB Brandon Wise Reactor Systems Engineer DSS SFNB The EPRI audit team consisted of the following staff:

NAME AFFILIATION Sean Martin EPRI Kurshad Muftuoglu EPRI Fred Smith EPRI Kevin Barber Westinghouse Electric Co. (WEC)

Aaron Everhard WEC Eric Husser WEC Brian Ising WEC Jeffrey Kobelak WEC Nathaniel Mackereth WEC

5.0 CONCLUSION

The audit accomplished the objectives and goals listed in the audit plan by allowing direct interaction with EPRI technical experts. The NRC staff were able to obtain clarification on

OFFICIAL USE ONLYPROPRIETARY INFORMATION OFFICIAL USE ONLYPROPRIETARY INFORMATION various topics, examine notes supporting EPRIs responses, and discuss differences in technical opinion. The clarifications and examination will allow the NRC staff to assess the need for RAIs and future RAI responses more efficiently.