ML25065A056

Non-Proprietary Enclosure 1 Report for NRC Regulatory Audit of Westinghouse Topical Report WCAP-18850-P/NP, Revision 0, Adaptation of the Full Spectrum LOCA (Fsloca) Evaluation Methodology to Perform Analysis of Cladding Rupture for High

ML25065A056
Person / Time
Site:	Westinghouse
Issue date:	03/13/2025
From:	Ekaterina Lenning Licensing Processes Branch
To:	Zozula C Westinghouse
References
EPID L-2023-TOP-0054
Download: ML25065A056 (1)

Text

Enclosure 1 U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION REGULATORY AUDIT REPORT FOR WESTINGHOUSE ELECTRIC COMPANY TOPICAL REPORT WCAP-18850-P/NP, REVISION 0, ADAPTATION OF THE FULL SPECTRUMTM LOCA (FSLOCATM) EVALUATION METHODOLOGY TO PERFORM ANALYSIS OF CLADDING RUPTURE FOR HIGH BURNUP FUEL DOCKET NO. 99902038; EPID: L-2023-TOP-0054

1.0 BACKGROUND

By letter dated February 29, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24060A161), Westinghouse Electric Company (Westinghouse) submitted Topical Report (TR) WCAP-18850-P/NP, Revision 0, Adaptation of the FULL SPECTRUM' LOCA (FSLOCA') Evaluation Methodology to Perform Analysis of Cladding Rupture for High Burnup Fuel (ADAMS Package Accession No. ML24060A160) to the U.S. Nuclear Regulatory Commission (NRC) for review and approval. WCAP-18850-P/NP, Revision 0, presents a rationale and a modified FSLOCA application methodology to predict cladding rupture using existing Westinghouse codes for analyses of pressurized water reactor (PWR) loss-of-coolant accident (LOCA) events identified in Chapter 15 of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (SRP). The proposed modified methodology is to be used for burnups up to [ ] rod and assembly average.

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 hybrid regulatory audit on October 22 through October 23, 2024, based on the audit plan issued on September 19, 2024 (ADAMS Package Accession No. ML24256A181).

The audit was held in accordance with the NRC Office of Nuclear Reactor 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 November 27, 2024 (ADAMS Accession No. ML24332A124).

2.0 REGULATORY AUDIT OBJECTIVES The objective of this audit was to increase review process efficiency through direct interaction with Westinghouses technical experts. More specifically, in preparation for the audit, Westinghouse made available, through their online document portal, a draft supplement document comprising additional information addressing concerns identified by the NRC staff during the acceptance review. The audit allowed the NRC staff to examine this supplemental document and obtain clarification on its contents, have extended discussions about differences in technical opinion, examine supportive documentation for the supplemental information and for the TR as submitted, and identify those areas of the review that need additional focus.

1. The applicability of fuel fragmentation and relocation models [

] considering the full set of experimental test data presently available and higher fuel burnups being proposed in WCAP-18850-P/NP, Revision 0.

2. Application of the evaluation model (EM) across the complete break spectrum,

[

]

3. Westinghouse-proposed limitations and conditions applicable to WCAP-18850-P/NP, focused particularly on areas where proposed treatments the proposed limitations and conditions differ from similar TR methodologies described in WCAP-18446-P-A and WCAP-16996-P-A.

4. The proposed fuel rod nodalization approach for WCAP-18850-P/NP, Revision 0 (e.g., as described on page 3-6 of the TR), including the basis and objectives for modeling [

]

a. Of particular consideration is the modeling of [ ] Unlike the baseline FSLOCA evaluation model, a fuel rod located at the edge or corner of a high-burnup assembly may have higher-power adjacent fuel rods on one or more sides (e.g., if adjacent to a fresh or once-burnt assembly) than a similarly conditioned rod in the interior of a high-burnup assembly.

5. Basis for adjustments to [ ] discussed in Section 3.3.2 of WCAP-18850-P/NP, Revision 0, considering the following factors:

a. The experimental conditions for some datasets cited in this Section of WCAP-18850-P/NP, Revision 0, appear to have involved [

] Significant cladding surface temperature differentials have long been known to reduce rupture strains, and prototypical temperature differentials under LOCA conditions may be challenging to ascertain.

b. Discussion of how cladding rupture strains measured in the SATS facility compare to those of other facilities that were capable of maintaining more uniform cladding surface temperatures for any types of zirconium-alloy claddings.

c. Discussion of the [

]

d. Discussion of the [

]

e. Clarification of how Figure 3.3-5 in WCAP-18850-P/NP, Revision 0, was generated, and whether consideration of data at [

]

6. Fragmentation susceptibility threshold, including the following factors:

a. [

]

b. Whether any physical rationale exists to deem a [

] when determining the fragmentation threshold.

c. The selection of the fuel fragment size at which fragmentation is deemed to have occurred and its impacts on the burnup threshold.

d. For a plant with a large number of fuel rods susceptible near the threshold for fuel dispersal, the selection of the fragmentation and dispersal threshold may be quite significant. Since each analyzed fuel rod in the analytical model may represent thousands of fuel rods in an actual reactor core, even relatively small amounts of dispersed fuel on a per rod basis could be significant for the core as a whole.

7. Discussion of the technical basis for the proposed transient fission gas release model described in Section 3.6 of WCAP-18850-P/NP, Revision 0.

8. Discussion of the modeling of packing fraction and its behavior [

] as described in Section 3.7 of WCAP-18850-P/NP, Revision 0.

9. Discussion of inputs to the [

] according to WCAP-18850-P/NP, Revision 0, TR methodology.

10. Clarification on normalized fission interaction frequency and the behavior in Figure 4.3-1, as compared to the discussion in Section 9.4 of WCAP-16996-P-A, Revision 1, which discusses [

]

11. The basis for assuming a reactor coolant pump trip time of 5 minutes for intermediate breaks. While overestimating pump trip timing may be conservative for small breaks with times of peak cladding temperature well exceeding 5 minutes, an assumed trip time of 5 minutes may be neither realistic nor conservative for intermediate breaks with times of peak cladding temperature within 5 minutes. Reference a previous review of this issue for a different fuel vendor (e.g., Section 3.2.2.4 of the NRC staffs safety evaluation in ADAMS Package Accession No. ML20325A088).

a. The modeling of the operator action to trip reactor coolant pumps can be a critical issue for the intermediate break range and may affect a number of conclusions made in the sensitivity studies included in WCAP-18850-P/NP, Revision 0 (e.g., with respect to availability of offsite power, limiting break size, etc.).

12. Discussion of the basis for the uncertainty/modeling approaches to the baseline FSLOCA modeling approaches proposed for the WCAP-18850-P/NP, Revision 0, TR methodology, including those related to [ ]

a. Particular focus is intended upon the validation of the modeling approaches and the completion of appropriate validation analyses on both a separate effects and integral basis for the full suite of validation comparisons.

13. Discussion of which evaluation model is being used for the intermediate break region, and its appropriateness. Both the [

] and the rationale for the complete set of modeling approaches applied to the intermediate break range is not clear.

14. Clarification on applicability and conclusions to be drawn from sensitivity studies. For example, on page 5-17, WCAP-18850-P/NP, Revision 0, appears to make conclusions concerning 2-Loop pressurized-water reactors (PWRs). While the actual context and analysis appears to be indicative of Westinghouse 2-Loop PWRs, in a literal sense, other PWRs within the proposed scope of WCAP-18850-P/NP, Revision 0, could arguably be categorized as having two reactor coolant system loops (i.e., Combustion Engineering plants).

15. Clarification of what conclusions may be drawn from sensitivity studies in WCAP-18850-P/NP, Revision 0. Although some sensitivity studies appear to provide clear indications, not all appear definitive or fully comprehensive. Potential questions concerning the sensitivity studies include the following:

a. Impact of reactor coolant pump assumed trip timing.

b. Randomness in single-comparison analyses that could affect resolution or power of the simulations to distinguish the limiting condition.

c. Applicability of sensitivities to all plants - for some plants/parameters, there are no completed sensitivities.

16. Discussion of the lack of approval for the PAD5 code and its impacts on the WCAP-18850-P/NP, Revision 0, review, for example, for some parameters such as stored heat (Section 3.1 of WCAP-18850-P/NP, Revision 0) [

]

17. To what extent do the [

]

The NRC staff discussed each topic with Westinghouse, and the following is a summary of the important points and outcomes for each topic:

1. This broad topic was discussed in more detail under items 5, 6, 7, and 8. Nonetheless, Westinghouse stated that the new data was taken into account when assessing and developing the models related to fuel fragmentation, relocation, and dispersal.

Westinghouse provided several clarification points in response to discussion and questions from the NRC staff:

a. Previously, no fragmentation or relocation was modeled until fuel rod burst was predicted, but the new model accounts for these phenomena prior to rod burst.

b. [

]

2. Westinghouse described the treatment of the intermediate break region (i.e., Region IB) evaluation model and how it relates to Region I and Region II. Westinghouse noted that the model is [

] The NRC staff determined that RAI would be necessary to docket additional information and justification.

3. Westinghouse and the NRC staff discussed the limitations and conditions and noted some possible changes to the language in the Westinghouse-proposed limitations and conditions.

4. Westinghouse described the conservative energy redistribution methodology in the FSLOCA EM that includes consideration of the example described in this question, such that the limiting rods should be identified and correctly modeled. The NRC staff determined that additional details provided by Westinghouse should be documented in a response to an RAI.

5. Westinghouse made [

] The NRC staff stated that an RAI would likely be sent on this topic to formally document some of the discussion points made during the audit.

6. The NRC staff discussed the database for fragmentation susceptibility threshold with Westinghouse. [

] The NRC staff determined that an RAI should be asked for further justification if Westinghouse decides to use the [ ]

7. Westinghouse provide more detailed descriptions of its transient fission gas release model and responded to the NRC staff comments and questions. The NRC staff determined that some of the additional details provided by Westinghouse should be documented in a response to an RAI.

8. The NRC staff questioned Westinghouse on how the [ ] bounds were chosen, and about a potential burnup dependence of these bounds. Westinghouse agreed with the NRC staff observation that the proposed bounds did not encompass all the available data. The NRC staff determined that an RAI should be asked on this topic.

9. Westinghouse described some sensitivities associated with the [ ]

such as changes in isotopic composition for higher enriched and higher burnup fuel with respect to decay heat. Westinghouse noted that the [

]

10. The NRC staff identified a discrepancy between the old and new models described in WCAP-16696-P-A and WCAP-18850-P/NP, Revision 0. Westinghouse clarified that the description of the new model is correct and reflective of the updates.

11. Westinghouse described its sensitivity studies related to reactor coolant pump (RCP) trip timing. The NRC staff determined that an RAI should be asked for further justification and docketing of information.

12. Westinghouse described the treatment of [ ] The NRC staff determined that an RAI is necessary to docket information related to the new treatment of [ ]

13. Westinghouse provided a detailed list of modeling differences between Region I and Region IB. The NRC staff determined that an RAI would be necessary to document a complete description of the calculational procedure for Region IB calculations.

14. The NRC asked Westinghouse about the conclusions that could be reasonably drawn from the sensitivity studies provided in the TR. The NRC staff determined that an RAI would be needed to determine if the sensitivity studies are sufficiently representative, such that there is increased confidence in the conclusion drawn in WCAP-18850-P/NP, Revision 0, TR regarding the sensitivity studies.

15. The NRC staff asked Westinghouse about the sensitivity of some parameters discussed in the sensitivity studies included in TR WCAP-18850-P/NP, Revision 0, with particular focus on the RCP trip timing. The NRC staff determined that an RAI would be necessary to docket information regarding the treatment of RCP trip timing.

16. Westinghouse described the expectations and acceptance criteria for WCOBRA-TRAC/TF2 and an approved version of the PAD5 code. In particular, [

] This will be documented in a Limitation and Condition.

17. Westinghouse described the corrosion characteristics of AXIOM, ZIRLO, and Optimized ZIRLO', with particular focus on the hydrogen pickup and its relation to oxide thickness. The NRC staff did not identify a need for an RAI.

5.0 EXAMINED AUDIT DOCUMENTS

1. [ ]

2. [ ]

3. [ ]

4. [ ]

5. [ ]

6. [ ]

7. [ ]

8. [ ]

6.0 CONCLUSION

The regulatory audit accomplished the objectives listed in Section 2.0 by allowing direct interaction with Westinghouses technical experts. The NRC staff obtained clarification on the contents of the supplement, examined calculation notes supporting the supplement and the TR as-submitted, and discussed at-length differences in technical opinion. The clarifications and examined calculation notes helped the NRC staffs review. The discussions on the topics of concern allowed the NRC staff and Westinghouse to reassess positions and facilitate full resolution of these concerns during the review process via RAIs.