ML25317A545
| ML25317A545 | |
| Person / Time | |
|---|---|
| Site: | 99902035 |
| Issue date: | 12/12/2025 |
| From: | Licensing Processes Branch |
| To: | |
| References | |
| EPID L-2025-TOP-0002 SSP-14-P01/028-TR-S1, Rev. 0 | |
| Download: ML25317A545 (0) | |
Text
Enclosure REGULATORY AUDIT REPORT FOR STUDSVIK TOPICAL REPORT SSP-14-P01/028-TR-S1, REVISION 0, SUPPLEMENT 1, GENERIC APPLICATION OF THE STUDSVIK SCANDPOWER CORE MANAGEMENT SYSTEM TO PRESSURIZED WATER REACTORS; SUPPLEMENT FOR EXTENDED ENRICHMENT, BURNUP, AND SMRS DOCKET NO. 99902035 EPID: L-2025-TOP-0002
1.0 BACKGROUND
By audit plan dated May 20, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25136A017), the U.S. Nuclear Regulatory Commission (NRC) staff conducted an audit for understanding of Topical Report (TR) SSP-14-P01/028-TR-S1, Generic Application of the Studsvik Scandpower Core Management System to Pressurized Water Reactors: Supplement for Extended Enrichment, Burnup, and SMRs [small modular reactors].
Studsvik Scandpower, Inc. (Studsvik) submitted SSP-14-P01/028-TR-S1 to the NRC for review and approval (ML25028A255). The audit was held on June 24 through 26, 2025, at the Studsvik facility in Idaho Falls, Idaho.
The NRC staff performed an audit for understanding to support timely completion of a safety evaluation and minimizing the potential need for a request for additional information (RAI) in accordance with Office of Nuclear Reactor Regulation Office Instructions LIC-111, Regulatory Audits, and LIC-500, Topical Report Review Process.
2.0 REGULATORY AUDIT OBJECTIVES The objective of the audit was to increase review process efficiency through interaction with Studsviks technical experts. During the audit, the NRC staff reviewed pertinent documentation made available by Studsvik. A more detailed narrative on the topics covered is included below in Section 4.0, Discussion, of this audit report. A list of all the documents that the NRC staff reviewed is included in Section 5.0, Documents Reviewed.
The NRC audit team was composed of the following members:
Alex Collier, Technical Reviewer, Nuclear Methods and Fuel Analysis Branch (SFNB),
Division of Safety Systems (DSS), Office of Nuclear Reactor Regulation (NRR)
Kevin Heller, Technical Reviewer, SFNB, DSS, NRR Devin Bradshaw, Technical Reviewer, Containment and Plant Systems Branch, DSS, NRR Lois James, Senior Project Manager, Licensing Projects Branch, Division of Operating Reactor Licensing, NRR The following Studsvik personnel represented or supported Studsvik during the audit:
Phil Sharpe, VP Applications Gerardo Grandi, Senior Nuclear Engineer Joel Rhodes, Chief Technology Officer Joshua Hykes, Senior Nuclear Engineer Petri Guimaraes, Senior Nuclear Engineer Rodolfo Ferrer, Manager Methods Tamer Bahadir, Engineering Fellow William Dawn, Senior Nuclear Engineer Jeff Borkowski, Chief Operations and Engineering Officer 3.0 REGULATORY AUDIT BASES While there are no directly applicable regulatory requirements for a generic application of a core simulator, a core simulator must adequately model core designs with an acceptable amount of uncertainty, as the core design establishes a basis for downstream safety analyses, which are used to demonstrate compliance with General Design Criteria 10, 11, 20, and 26 as defined in Appendix A, General Design Criteria for Nuclear Power Plants, to Part 50, Domestic licensing of production and utilization facilities, of Title 10 of the Code of Federal Registration and as discussed in Section 4.3, Revision 3, Nuclear Design, of NUREG 0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (ML070740003).
4.0 DISCUSSION The audit began with a discussion of the topics referenced in the audit plan (ML25136A025).
The NRC staff also reviewed documents that Studsvik uploaded to the electronic reading room.
A detailed discussion of these topics, potential RAIs, and other discussions is provided below.
4.1 Fuel Thermomechanical Module The NRC staff reviewed documentation containing a description of the fuel thermomechanical module, including inputs, outputs, models, or correlations implemented within the module, and documentation regarding validation of SIMULATE5 fuel temperature predictions. The NRC staff discussed the range of applicability of the fuel thermomechanical module and any validation or verification at increased enrichments and higher burnups. In particular, the NRC staff focused the discussion on the burnup dependency of the thermal conductivity degradation (TCD) model.
The Studsvik representatives noted that the model contained within SIMULATE5 is not a fuel performance model. The Studsvik representatives further stated that the TCD model used in SIMULATE5 is the same model used in Fuel Analysis under Steady-state and Transients (FAST) ML20099A090). The Nuclear Fuel Industries correlation used in the FAST and SIMULATE5 models is valid for the burnup application range requested by Studsvik, and the NRC staff confirmed from FAST documentation (ML20099A089) that the correlation has been validated to the burnup application range requested by Studsvik. The Studsvik representatives also stated that the model contains a pellet rim effect model, and the NRC staff reviewed the effects of the pellet rim model in Studsvik s response to RAI 6 of SSP-14-P01/028-TR-S1 (ML17279A986).
4.2 Predicted Concentrations of Isotopes of High Importance at Higher Burnups The NRC staff requested to see predicted concentrations of isotopes of high importance at higher burnups and comparisons to either experimental data or code-to-code comparisons. The Studsvik representatives provided results of code-to-code comparisons between CASMO5 and SERPENT2 for number densities of uranium-235 (U-235), plutonium-239 (Pu-239), and plutonium-241 (Pu-241) in lattices that were burned up to 90 gigawatt-days per metric ton of uranium (GWd/MTU) in the documentation. The NRC staff reviewed the results and developed an RAI requesting comparisons of number densities of major actinides and fission products in lattices with higher enrichment and burnup.
4.3 Validation of Nuclear Uncertainty Factors (NUFs) and Nuclear Reliability Factors (NRFs) Methodologies The NRC staff asked the Studsvik representatives about the validation of the NUFs and NRFs methodologies at increased enrichment and higher burnup. The Studsvik representatives stated that because no methodology changes were made since the original approval and the methods for deriving NUFs are independent of enrichment and burnup, that no additional validation is required for the methodology used to develop NUFs. For NRFs, the Studsvik representatives also stated that because no significant biases were found in the code-to-code comparisons in the supplement, the NRFs derived in SSP-14-P01/028-TR-S1 remain applicable to fuels with increased enrichment and higher burnup.
4.4 Critical Benchmarks The NRC staff discussed the critical benchmarks in Section 2.3 of the supplement with the Studsvik representatives. In particular, the NRC staff asked about the keff value of the experiments, the experimental uncertainty, and the results of the benchmarks. The Studsvik representatives stated that the benchmarks had a keff of unity, and that the experimental uncertainty was difficult to determine. The Studsvik representatives further stated that the axial buckling component could be a contributor to the larger differences between CASMO5 and MCNP6 predictions. After reviewing the benchmarks and confirming that the critical benchmarks have a keff of unity and that CASMO5 and MCNP6 differences were at times outside of experimental uncertainty, the NRC staff developed an RAI further discussing the validation of the NRF given the apparent increase in uncertainty in the eigenvalue results.
4.5 Nuclear Design of Lattices The NRC staff reviewed the nuclear design of the lattices used for the analyses in Sections 2.2, 3.2, and 3.3 of the supplement in the documentation provided by Studsvik. The NRC staff discussed the design philosophy of the lattices and the results of the code-to-code comparisons with the Studsvik representatives to determine if the lattices were representative of potential core designs utilizing increased enrichment and higher burned fuels.
4.6 Generic Applicability to Small Modular Pressurized Light Water Reactors The NRC staff asked for clarification of Studsviks request for SSP-14-P01/028-TR-S1 and its supplement to be generically applicable to small modular pressurized light water reactors (PWR SMRs). The Studsvik representatives clarified that any generic application to PWR SMRs would be subject to the limitations and conditions of SSP-14-P01/028-TR-S1 and its supplement (i.e., methodology application range limits).
4.7 Reflector Cross Sections The NRC staff asked the Studsvik representatives if the reflector cross sections were affected by higher enrichments and burnups. The Studsvik representatives stated that the generation of the reflector cross sections is not dependent on higher enrichments or burnups; rather, any neutron source will be effective in generating the cross sections because of the slowing down effect of the baffle and water prior to the reflector.
4.8 Miscellaneous Topics Other topics were briefly discussed for the NRC staffs understanding of the material, rather than for resolution of concerns or identification of RAIs. These topics include discontinuity factors, pin power calculations, downstream methods, radial pellet power profiles, and case matrices.
5.0 DOCUMENTS REVIEWED The following documents provided by Studsvik were reviewed by the NRC staff during the audit:
ETE-NAF-2024-0113, Revision 0, Dominion Energy Engineering Technical Evaluation, Summary of CMS5 Burnup Comparison for North Anna High Burnup Assembly FM3, dated January 2023.
SSP-2-P01/006-CN, Revision 0, Studsvik Calculation Notebook for CASMO5-MCNP6 Uniform Lattice Reactivity Comparison and Computational Benchmarks for Higher Enrichment, dated May 2025.
SSP-2-P01/009-CN, Revision 0, Studsvik Calculation Notebook for CASMO5-Serpent2 Fuel Pin Cell and Lattice Reactivity, Fission Rate and Number Density Comparison for Higher Burnups and Enrichments, dated June 2025.
SSP-07/431, Revision 8, Studsvik CASMO5 - A Fuel Assembly Burnup Program - User Manual, dated August 2014.
SSP-10/465, Revision 3, Studsvik SIMULATE5 - Methodology, dated January 2014.
SSP-14-P01/012-R, Revision 1, Studsvik CASMO5 PWR Methods and Validation Report, dated November 2015.
SSP-14-P01/016-C, Revision A, Studsvik Analysis Documentation Notebook for S5C PWR Case Matrix Assessment, dated December 2015.
SSP-14-P01/020-C, Revision 1, Studsvik Analysis Documentation Notebook for SIMULATE5 Pin-to-Box Uncertainties, dated August 2015.
SSP-14-P01/022-C, Revision 0, Studsvik Analysis Documentation Notebook for Benchmarking SIMULATE5 with Halden Fuel Temperature Measurements, dated October 2015.
SSP-14-P01-023-C, Revision 0, Studsvik Analysis Documentation Notebook for Benchmarking the SIMULATE5 BEAVRS Model with CASMO5 MxN, dated December 2015.
In addition to internal documents, Studsvik also supplied the following conference proceedings to the NRC staff during the audit:
Beavrs Benchmark Evaluation with CASMO5 and SIMULATE5, Bahadir, T.
PHYSOR 2020, March 29 - April 2, 2020.
CASMO-4E and CASMO-5 Analysis of the Isotopic Compositions of the LWR [light water reactor]-PROTEUS Phase II Burnt PWR [pressurized water reactor] UO2 [uranium dioxide] Fuel Samples, Grimm, P., Perret, G., and Ferroukhi, H., PHYSOR 2014.
September 28 - October 3, 2014.
CASMO-5 Analysis of Reactivity Worths of ANALYSIS OF REACTIVITY WORTHS OF Burnt PWR Fuel Samples Measured in LWR-PROTEUS Phase II, Grimm, P.,
Hursin, M., Perret, G., Siefman, D., and Ferroukhi, H., PHYSOR 2016. May 1-5, 2016.
Modeling of Base Irradiation Histories of LOCA [loss-of-coolant-accident] Tests Using CMS5 and ENIGMA, Grandi, G., Karlsson, J., and Hemlin, M., Top Fuel 2019.
September 22 - 27, 2019.
NEA/OECD TVA [Tennessee Valley Authority] Watts Bar Unit 1 Multi-Physics Multi-Cycle Depletion Benchmark with CASMO5 and SIMULATE5, Georgieva, E., and Bahadir, T. PHYSOR 2024. April 21 - 24, 2024.
6.0 CONCLUSION
The audit accomplished the objectives listed in Section 2.0 by allowing direct interaction with Studsviks technical experts. The NRC staff participants were able to obtain clarification on multiple questions and examine calculation notes and supporting documentation. The NRC staff will continue its review of the proposed TR and intends to issue RAIs to address any issues where further information is necessary to complete the review. No regulatory decisions were made in the audit.