ML24326A329
| ML24326A329 | |
| Person / Time | |
|---|---|
| Site: | 05200050, 99902078 |
| Issue date: | 11/21/2024 |
| From: | NuScale |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML24326A327 | List: |
| References | |
| LO-176070 | |
| Download: ML24326A329 (1) | |
Text
Response to NuScale Topical Report Audit Question Question Number: A-LOCA.LTR-35 Receipt Date: 03/11/2024 Question:
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Response
The original audit response is unchanged and the supplemental information that addressed feedback from May 6, 2024 begins with Response Supplement - June 2024. Additional feedback received on July 10, 2024 is addressed in the section titled Response Supplement - July 2024 Items 1 and 2 In compliance with 10 CFR 50 Appendix K C.7.a, the hot region in the NRELAP5 model is no larger than that of one fuel assembly. The lumping of the limiting assembly and performing a critical heat flux (CHF) evaluation using a local condition-based model is a common practice that is not unique to NuScales Loss-of-Coolant Accident (LOCA) evaluation methodology. For example, the US-APWR small break LOCA methodology (Reference 1, Section 4.6.6) limiting hot assembly was lumped into a single axial channel with CHF evaluated using the 1986 NuScale Nonproprietary NuScale Nonproprietary
Groeneveld look-up tables. Similar to NSPN-1, the Groeneveld look-up tables are a local condition-based CHF model. The Groeneveld look-up tables used as described in Reference 1 are henceforth referred to as the Groeneveld model. ((
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Item 3 The VIPRE-01 Safety Evaluation (SE) (Reference 2) identifies that the use of a specific CHF correlation and its safety limit is to be provided by an applicant for a specific license application. The focus of the VIPRE-01 SE, in terms of CHF, is to address whether a CHF correlation developed based on steady-state CHF test data can be applied to the calculation of CHF for a transient scenario. The topic of interest in the VIPRE-01 SE is the improvement and deviation from a legacy method that was based on a snap-shot approach. For this legacy method, CHF was evaluated using the core boundary conditions at any instant during a transient solution to perform a steady-state analysis that provides local fluid conditions to evaluate CHF. The VIPRE-01 SE evaluates the appropriateness of an improved approach using a steady-state CHF correlation coupled with the actual local conditions from a transient core thermal-hydraulic solution. The discussion and conclusions of the VIPRE SE related to CHF are not unique to VIPRE-01 or specific to a single steady state CHF correlation. As identified within the SE, the intent is to address a broader issue of the use of any steady-state CHF model used to evaluate a transient CHF response. System codes that evaluate CHF commonly use steady-state CHF models in combination with system code transient core thermal-hydraulic solution. The mechanics of CHF NuScale Nonproprietary NuScale Nonproprietary
evaluation are synonymous whether a subchannel or system code is utilized. The difference is in the applicability of the model geometry and treatment of fluid conditions (subchannel or bundle-averaged) to which a CHF correlation is developed and applied. The VIPRE-01 SE provides a thorough summary and conclusions of various assessments completed utilizing transient CHF data. References in the VIPRE-01 SE generally conclude that the use of steady-state CHF correlations is consistent or conservative when evaluating the performance of steady-state CHF correlations relative to transient CHF tests. Steady-state CHF correlations are considered conservative to their predictability of transient data if the steady-state correlations predict a lower allowed CHF value or predict CHF onset earlier. The VIPRE-01 SE identifies that there are mixed performance results using steady-state CHF correlations for depressurization events such that the steady-state CHF correlations may not predict the CHF onset value or time of onset conservatively. The one test identified as providing mixed results in the VIPRE-01 SE is related to a test campaign performed by Westinghouse and EPRI on a 5x5 bundle (Reference 3). Inspection of the test matrix in Reference 3 reveals the tests where steady-state CHF correlations are deemed inappropriate are for conditions that result in flow reversal in the core. The VIPRE-01 SE also noted investigations by Leung and Gallivan (Reference 4) and summarizes that the transient data was predicted reasonably well. Further investigation into the work in Reference 4 reveals that tests that were not predicted well, similar to the EPRI campaign, are related to flow reversal conditions. Tests with positive up flow demonstrate that steady state CHF correlations are appropriate for transient evaluation and only when flow reversal occurs steady state correlations may not be appropriate. ((
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(( }}2(a)(c) (( }}2(a)(c) The following documents are provided in the eRR to support this response:
EC-152283, Revision 0, NRELAP5 Core Crossflow Model Development and Benchmarking
EC-103521, Revision 3, VIPRE-01 Sensitivity Analysis for the Core-250B Design References 1. NRC Safety Evaluation for Small Break LOCA Methodology for US-APWR, MUAP-07013-NP-A, Revision 3 (ML14094A115). 2. Safety Evaluation Report on EPRI NP-2511-CCM VIPRE-01, Nuclear Regulatory Commission, May 1986. 3. Full-Scale Controlled Transient Heat Transfer Tests Data Analysis Report, Electric Power Research Institute, NP-2547, Final Report, August 1982. 4. Leung, J.C., and Gallivan K. A., Prediction of Critical Heat Flux During Transients, Proceedings of the American Nuclear Society/European Nuclear Society Topical Meeting on Thermal Reactor Safety, CONF-800403, Vol. 2, pg. 1229 - 1239, 1980. NuScale Nonproprietary NuScale Nonproprietary
Response Supplement - June 2024 Items 1 and 2 Additional analysis was performed to compare the MCHFR results using the approved VIPRE-01 subchannel methods to the MCHFR results from the NRELAP5 model with crossflow. (( }}2(a)(c) Additionally, the timing of MCHFR between NRELAP5 and VIPRE-01 is consistent. (( }}2(a)(c) The updated analysis is documented in Section 4.4.2 of EC-152283, Revision 1, NRELAP5 Core Crossflow Model Development and Benchmarking. EC-152283 Revision 1 is available in the eRR as part of this response. NuScale Nonproprietary NuScale Nonproprietary
Item 3 Table 1 outlines the sources necessary to conclude that the mixed results summarized in the VIPRE-01 Safety Evaluation Report (Reference 1.1) occur when there is flow reversal in rapid depressurization events. NuScale Nonproprietary NuScale Nonproprietary
Table 1. Summarized Steady-State Critical Heat Flux Data by Transient Including Model Performance Source Transient Steady-State Critical Heat Flux Model Steady-State Model Performance Redfield 1967 (Reference 1.2) Depressurization (instantaneous decrease in pressure from 1400 to 700 psia) Unknown steady-state model Conservative for breaks near the core exit Cermak 1970 (Reference 1.3) Depressurization with no flow reversal (75 - 630 psi/sec) W-3 Conservative Shiralkar 1972 (Reference 1.4) Depressurization (43 - 136 psi/sec) Facility specific empirical steady-state model Conservative Cumo / ANL 1972 (Reference 1.5) Depressurization with inlet flow stoppage (Freon-11) CISE Consistent Guerrero / ANL 1976 (Reference 1.5) Depressurization with flow reversal Bowring Conservative Biasi Consistent CISE Conservative EPRI 1982 (Reference 1.6) Depressurization with flow reversal W-3, MacBeth,B&W-2, Biasi, CISE Mixed Cermak 1972 (Reference 1.7) Depressurization with no flow reversal (200 - 10,000 psi/sec) Tong (assessed against facility specific data) Conservative Semiscale S 09 (Reference 1.8, Reference 1.9) Depressurization with flow reversal (instantaneous decrease in pressure from 2,253 to 1,000 psia) Local-parameter based steady state models: MacBeth, Biasi, B&W-2, GE, LOFT Consistent (LOFT was non-conservative but noted to be applied outside its applicability domain.) Global-parameter (inlet enthalpy) based steady-state model: W-3 Non-Conservative Barnett Conservative NuScale Nonproprietary NuScale Nonproprietary
References 1.1. Safety Evaluation Report on EPRI NP-2511-CCM VIPRE-01, Nuclear Regulatory Commission, May 1986. 1.2. Redfield, J. A., Margolis, S. G., Murphy, J. H., & Snyder, G. A, Loss of Coolant from a Simulated Reactor Loop, Nuclear Applications, 3(4), 206-212, 1967. 1.3. Cermak, J.O., Farman, R.F., Tong, L.S., Casterline, J.E., Kokolis, S., Matzner, B., The Departure from Nucleate Boiling in Rod Bundles during Pressure Blowdown, ASME J Heat Transfer, 92(4), 621-627, 1970. 1.4. Shiralkar, B.S., Polomik, E.E., Lahey, R.T., et al., Transient Critical Heat Flux -- Experimental Results, GEAP-13295, 1972. 1.5. Leung, J.C., and Gallivan K. A., Prediction of Critical Heat Flux During Transients, Proceedings of the American Nuclear Society/European Nuclear Society Topical Meeting on Thermal Reactor Safety, CONF-800403, Vol. 2, 1229 - 1239, 1980. 1.6. Electric Power Research Institute, Full-Scale Controlled Transient Heat Transfer Tests Data Analysis Report, NP-2547, Final Report, August 1982. 1.7. Cermak, J.O., Farman, R.F, Post DNB Heat Transfer During Blowdown, WCAP-7837 (non-proprietary), Westinghouse Electric Corporation, January 1972. 1.8. Snider, Dale M., Analysis of the Thermal-Hydrualic Behavior Resulting in Early Critical Heat Flux and Evaluation of CHF Correlations for the Semiscale Core, TREE-NUREG-1073, March 1977. 1.9. Idaho National Engineering Laboratory, Experimental Data Report for Semiscale Mod-1 Tests S-02-9 and S-02-9A (Blowdown and Heat Transfer Tests), ANCR-1236, January 1976. NuScale Nonproprietary NuScale Nonproprietary
Response Supplement - July 2024 The differences between the results in TR-0516-49416, Revision 4, Non-Loss-of-Coolant Accident Analysis Methodology, Figure 4-3 and Figure 1 above is attributed to the type of model used with each analysis. (( }}2(a)(c) The models in Non-LOCA topical report are used for trending and screening purposes only and not for margin assessment. No changes to the SDAA are necessary. NuScale Nonproprietary NuScale Nonproprietary}}