ML24260A095
| ML24260A095 | |
| Person / Time | |
|---|---|
| Site: | Westinghouse |
| Issue date: | 09/19/2024 |
| From: | Gerond George Division of Operating Reactor Licensing |
| To: | Westinghouse |
| Shared Package | |
| ML24256A181 | List: |
| References | |
| EPID L-2024-NTR-0005, WCAP-18850-NP, Rev. 0, WCAP-18850-P, Rev. 0 | |
| Download: ML24260A095 (6) | |
Text
Enclosure 1 U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION REGULATORY AUDIT PLAN FOR THE UPCOMING CLOSED AUDIT WESTINGHOUSE ELECTRIC COMPANY 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 BURNUP FUEL DOCKET NO. 99902038 EPID: L-2024-NTR-0005
1.0 BACKGROUND
By letter dated February 29, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24060A160), 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 (Proprietary/Non-proprietary) for the U. S. Nuclear Regulatory Commission (NRC) review and approval.
The NRC staff has performed an initial review of the submitted TR and determined that a regulatory audit of the items in Section 4.0 Information Requests and Section 9.0 Audit Topics of this audit plan would facilitate the timely completion of our review. The NRC staff is continuing to review other aspects of the TR and may identify the need to audit additional subjects by separate correspondence or during audit discussions.
2.0 REGULATORY AUDIT BASES The NRC staff has determined that a hybrid regulatory audit is the most efficient approach toward a timely resolution of questions associated with this review. The audit will provide the NRC staff an opportunity to minimize the potential for requests for additional information by ensuring a clear mutual understanding of the WCAP-18850-P/NP, Revision 0, TR methodology and clarifying potential review questions with focused communication. As appropriate to facilitate efficient audit interactions, the NRC staff is requesting that Westinghouse provide access to the requested audit documentation through an online reference portal prior to the commencement of the hybrid audit.
Upon completion of this audit, the NRC staff is expected to achieve the following:
- 1.
Confirm internal Westinghouse information that supports statements made in the TR.
2
- 2.
Determine whether the information included in the audited documents is necessary to be submitted to support a safety conclusion.
The audit information the NRC staff determines to be necessary to support the development of its safety evaluation will be requested to be submitted on the docket.
3.0 REGULATORY AUDIT SCOPE AND METHODOLOGY The purpose of this audit is to gain a more detailed understanding of the information supporting Westinghouses proposed TR methodology.
The areas of focus for the regulatory audit are the information contained in the submitted TR, the enclosed audit information needs, and all associated and relevant supporting documentation (e.g., methodology, process information, calculations, etc.) identified below. The audit will be performed consistent with the NRC Office of Nuclear Reactor Regulation (NRR) Office Instruction LIC-111, Revision 1, Regulatory Audits, dated October 31, 2019 (ADAMS Accession No. ML19226A274).
4.0 INFORMATION REQUESTS Please make the following information available for the NRC staff to audit via an online reference portal by October 1, 2024, to support an efficient face-to-face audit:
- 1.
The analyses comparing models referenced in paragraph two of Section 4.2 of WCAP-18850-P/NP, Revision 0, specifically the analyses which may show potential non-conservatisms in the original reference model.
- 2.
The dependency analyses described in paragraph three of Section 4.4 of WCAP-18850-P/NP, Revision 0.
- 3.
Supporting documentation for the sensitivity analyses and demonstration calculations described in Section 6 of WCAP-18850-P/NP, Revision 0.
- 4.
[
]
- 5.
Calculations used to determine a [ ] axial gas communication, as described in Section 3.6.3 of WCAP-18850-P/NP, Revision 0.
In addition, to support the NRC staffs understanding of the TR and to help identify any additional information needed to support its review, the NRC staff requests that Westinghouse be prepared to discuss the audit topics identified in Section 9.0.
3 5.0 AUDIT TEAM Key Westinghouse personnel involved in the development of the TR should be made available for interactions on a mutually agreeable schedule to respond to any questions from the NRC staff.
Team Member Division Area of Responsibility Jeremy Dean NRR/DSS/SFNB1 Technical Lead Brandon Wise NRR/DSS/SFNB Technical Reviewer John Lehning NRR/DSS/SFNB Technical Reviewer Jack Vande Polder NRR/DSS/SFNB Technical Reviewer Patrick Raynaud NRR/DSS/SFNB Technical Reviewer Ekaterina Lenning NRR/DORL/LLPB2 Project Management 6.0 LOGISTICS The audit will involve both the review of documents on the electronic portal as well as an in-person face-to-face discussion. The online portion of the audit will be conducted from October 1, 2024, to October 31, 2024, through an online portal established by Westinghouse.
The NRC staff access to the online portal should be terminated on October 31, 2024.
The NRC staff requests a two-day hybrid (in-person and virtual) audit to be conducted at the Westinghouses Headquarters located in Cranberry, PA, to support technical discussions associated with the review and supporting audit information. The audit is planned to be conducted on October 22, 2024, to October 23, 2024.
7.0 SPECIAL REQUESTS The NRC staff would like access to the documents listed above in Section 4.0 through the online portal that allows the NRC staff to access documents via the internet. The following conditions associated with the online portal must be maintained throughout the duration that the NRC staff have 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 use of the online portal will be displayed on the login screen and will require acknowledgement by each user.
8.0 DELIVERABLES The NRC team will develop an audit summary report to convey the results. The report will be placed in ADAMS within 90 days of the completion of the final audit session. The audit information the NRC staff determines to be necessary to support the development of the NRC staffs safety evaluation will be requested to be submitted on the docket.
1 Division of Safety Systems (DSS)/Nuclear Methods and Fuels Branch (SFNB) 2 Division of Operating Licensees (DORL)/Licensing Projects Branch (LLPB)
4 9.0 AUDIT TOPICS
- 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.
- 2. Application of the evaluation model across the complete break spectrum, [
]
- 3. Westinghouse-proposed limitations and conditions applicable to WCAP-18850-P, 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 (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, considering the following factors:
- a. The experimental conditions for some datasets cited in this Section of WCAP-18850-P 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 was generated, and whether consideration of data at [
]
5
- 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.
- 8. Discussion of the modeling of packing fraction and its behavior [
] as described in Section 3.7 of WCAP-18850-P.
- 9. Discussion of inputs to the [
] according to WCAP-18850-P 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, 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 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 (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 methodology, including those related to [ ]
6
- 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 appears to make conclusions concerning 2-Loop PWRs. While the actual context and analysis appears to be indicative of Westinghouse 2-Loop pressurized-water reactors, in a literal sense, other pressurized-water reactors within the proposed scope of WCAP-18850-P 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.
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 review, for example, for some parameters such as stored heat (Section 3.1 of WCAP-18850-P), [
]
- 17. To what extent do the [
]