ML24039A149
| ML24039A149 | |
| Person / Time | |
|---|---|
| Issue date: | 02/06/2024 |
| From: | Joseph S NRC/NRR/DNRL/NRLB |
| To: | |
| References | |
| TR-108601-P, Rev 4, TR-0716-50350-P, Rev 3 | |
| Download: ML24039A149 (1) | |
Text
Presentation to the ACRS Subcommittee Staff Review of NuScale Topical Reports
TR- 108601- P, REV 4, STATISTICAL SUBCHANNEL ANALYSIS METHODOLOGY, SUPPLEMENT 1 TO TR- 0915- 17564- P-A, REVISION 2, SUBCHANNEL ANALYSIS METHODOLOGY TR- 0716- 50350- P, REV 3, ROD EJECTION ACCIDENT METHODOLOGY
February 6, 2024 (Open Session)
Non-Proprietary 1 NRC Technical Review Areas/Contributors
Statistical Subchannel Analysis Methodology Rebecca Patton (BC), Reactor Systems NRR/DSS/SRNB Antonio Barrett, NRR/DSS/SRNB Joshua Kaizer, NRR/DSS/SFNB Peter Lien, RES/DSA/CRAB II Rod Ejection Accident Methodology Rebecca Patton (BC), Reactor Systems NRR/DSS/SRNB Zhian Li, NRR/DSS/SRNB Ryan Nolan, NRR/DSS/SRNB Adam Rau, NRR/DSS/SNSB Andrew Bielen, RES/DSA/FSCB Project Managers Stacy Joseph, TR PM Getachew Tesfaye, Lead PM
2 Non-Proprietary SSAM Staff Review Timeline
NuScale submitted its Topical Report (TR) TR-108601- P, Rev 0 on December 30, 2021 (ML21364A133) as supplemented by letters dated April 25, 2022 (ML22115A222) and December 13, 2022 (ML22347A314).
Staff performed an audit between July 13, 2022 and September 27, 2023 (ML23295A001).
Following the audit, NuScale submitted Revisions 3 and 4 on October 12, 2023 (ML23285A341) and November 6, 2023 (ML23285A341) of the TR.
Staff issued the Advanced Safety Evaluation Report (SER) on November 6, 2023 (ML23277A007)
3 Non-Proprietary SSAM Regulatory Basis
- General Design Criterion 10, Reactor design, of Appendix A
The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
- Standard Review Plan, Section 4.4, Thermal and Hydraulic Design .
..there should be a 95-percent probability at the 95-percent confidence level that the hot
[fuel] rod in the core does not experience a DNB [departure from nucleate boiling] or boiling transition condition during normal operation or AOOs.
4 Non-Proprietary SSAM Staff SER Conclusions
- The SSAM is an acceptable methodology to calculate the margin to fuel thermal limits such as the critical heat flux ratio through a statistical combination of the uncertainties.
- There were two limitations and conditions:
- 1. An applicant referencing [the SSAM] in the safety analysis must also reference an approved CHF correlation which has been demonstrated to be applicable for use with [the NSAM]. (Carry over from NSAM)
- 2. The SSAM relies on multiple submodels to calculate the statistical critical heat flux analysis limit. While some of these submodels have been reviewed and approved as part of the NRC staffs review and approval of the SSAM, the submodels listed in the SER would need to be reviewed and approved before the application of this methodology for a licensing analysis.
5 Non-Proprietary Staff Review Timeline TR-0716-50350-P, Rev 3 Rod Ejection Accident Methodology
NuScale submitted its Topical Report (TR) TR-0716-50350 -P, Rev 2 on December 21, 2021 (ML21351A400).
NuScale supplemented its submittal by letter dated, September 14, 2022 in response to requests for additional information (RAI), RAI No. 9936 from the NRC staff.
Staff performed a limited scope audit between April 19, 2023 and September 27, 2023 (ML23295A001).
Following the audit, NuScale submitted Revision 3 of the TR on October 20, 2023 (ML23293A292)
Staff issued the Advanced SER on January 4, 2024 (ML23310A166)
6 Non-Proprietary Regulatory Basis
- General Design Criterion 28, Reactivity Limits, of Appendix A
Criterion 28Reactivity limits. The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition.
- Standard Review Plan Sections 4.2 and 15.4.8 and Regulatory Guide 1.236, Pressurized -Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Rod Drop Accidents for reactivity-initiated accidents.
7 Non-Proprietary Staff SER Conclusions
- TR-0716-50350 P, Revision 3 provides a systematic methodology for performing rod ejection accident (REA) analysis subject to the following limitations and conditions:
- 1. An applicant or licensee referencing this report is required to demonstrate the applicability of the REA methodology to the specific NPM design. The use of this methodology for a specific NPM design requires the NRC staff review and approval of the applicant or licensee determination of applicability.
- 2. The REA methodology is limited to evaluation of REAs for fuel that has not experienced significant depletion with control rods inserted, such as from non-baseload operation.
- 3. The staffs approval is limited to the use of the rod ejection methodology with TR-0616-48793-P-A, Revision 1 (Reference 14), Nuclear Analysis Codes and Methods Qualification, and TR-108601-P, Revision 4 (Reference 13),
Statistical Subchannel Analysis Methodology, Supplement 1 to TR-0915-17564-P-A, Revision 2, Subchannel Analysis Methodology.
8 Non-Proprietary Questions/comments from members of the public before the closed session starts?
9 Non-Proprietary