Stage Progress and Analysis of International Standard Problem 51 “Open Test”

Zhang Xueyan; Yang Jun; Wang Shiqi; Deng Chengcheng; Li Yuquan; Zhang Peng; He Dandan

doi:10.13832/j.jnpe.2022.06.0015

Volume 43 Issue 6

Dec. 2022

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Zhang Xueyan, Yang Jun, Wang Shiqi, Deng Chengcheng, Li Yuquan, Zhang Peng, He Dandan. Stage Progress and Analysis of International Standard Problem 51 “Open Test”[J]. Nuclear Power Engineering, 2022, 43(6): 15-23. doi: 10.13832/j.jnpe.2022.06.0015

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Zhang Xueyan, Yang Jun, Wang Shiqi, Deng Chengcheng, Li Yuquan, Zhang Peng, He Dandan. Stage Progress and Analysis of International Standard Problem 51 “Open Test”[J]. Nuclear Power Engineering, 2022, 43(6): 15-23. doi: 10.13832/j.jnpe.2022.06.0015

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Stage Progress and Analysis of International Standard Problem 51 “Open Test”

doi: 10.13832/j.jnpe.2022.06.0015

1.
Department of Nuclear Engeneering and Technology, School of Energy and Power Engineering, Huazhong University of Science and Technology, Wuhan, 430074, China
2.
State Power Investment Corporation Research Institute, Beijing, 100029, China

Received Date: 2021-12-31
Rev Recd Date: 2022-09-29
Publish Date: 2022-12-14

Abstract

Abstract

In order to verify the ability of the thermal-hydraulic code to simulate the behavior of passive nuclear reactor under the condition of loss of coolant accident, and to assess the ability of optimal estimation codes to predict specific experiments, based on China's large-scale global effect test bench-Advanced Core Cooling System Mechanism Experiment (ACME) facility, the Organization for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) organized the International Standard Problem No. 51 (ISP-51) project. A preliminary comparative analysis is made on the code calculation results submitted at the current open test stage. The results show that for the loss of coolant accident with a 2-inch small break in the same cold pipe section, the simulation results of the thermal-hydraulic optimal estimation code RELAP5 used by Huazhong University of Science and Technology and Polytechnic University of Catalonia are in good agreement with the experimental data in terms of triggering time and flow value of the passive safety system. In the simulation results of TRACE code used by Polytechnic University of Madrid and Spanish NFQ Company, the triggering time of each passive safety system is delayed, and the gas-liquid flow rate of ADS1~3 is significantly higher than the experimental value, which may be related to the selection of different critical flow models and the setting of valves and pipelines. This project has set a precedent for non-member countries of OECD to initiate and take charge of international standard problem projects. It also helps relevant scientific research teams in China to become more familiar with the operation, organization and management of international nuclear science and technology research cooperation projects, and to undertake more work in international nuclear science and technology cooperation.
- International standard problem,
- ACME facility,
- Passive safety,
- Code validation

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References(22)

References

[1]	KARWAT H. CSNI: International Standard Problems (ISP), brief descriptions (1975−1999): NEA/CSNI/R(2000)5[R]. Paris: NEA/CSNI, 2000.
[2]	FIRNHABER M, FISCHER K, SCHWARZ S, et al. International standard problem ISP44 (KAEVER): experiments on the behavior of core-melt aerosols in a LWR containment-comparison report: NEA/CSNI/R(2003)5[R]. Paris: NEA/CSNI, 2002.
[3]	HERING W, HOMANN C, LAMY J S, et al. Comparison and interpretation report of the OECD international standard problem No. 45 exercise (QUENCH-06 experiment): NEA/CSNI/R(2002)23[R]. Paris: NEA/CSNI, 2002.
[4]	CLÉMENT B, HASTE T. Comparison report on international standard problem ISP-46 (PHEBUS FPT1): NT SEMAR 03/021 Revision 3[R]. Paris: IRSN, 2003.
[5]	ALLELEIN H J, FISCHER K, VENDEL J, et al. International standard problem ISP-47 on containment thermal-hydraulics final report: NEA/CSNI/R(2007)10[R]. Paris: NEA/CSNI, 2007.
[6]	DOUGHTY G, SHEPHERD D, PRINJA N, et al. International standard problem No. 48-containment capacity. Synthesis report: NEA/CSNI/R(2005)5[R]. Paris: NEA/CSNI, 2005.
[7]	HESSHEIMER M. International standard problem No. 48 analysis of a 1: 4-scale prestressed concrete containment vessel model under severe accident conditions: NEA/CSNI/R(2005)7[R]. Paris: NEA/CSNI, 2005.
[8]	KOTCHOURKO A, BENTAIB A, FISCHER K, et al. ISP-49 on hydrogen combustion: NEA/CSNI/R(2011)9[R]. Paris: NEA/CSNI, 2012.
[9]	KANG K H, MOON S K, PARK H S. ATLAS facility and instrumentation description report: KAERI/TR-3779/2009[R]. Seoul: KAERI, 2009.
[10]	CHOI K Y, BAEK W P, CHO S, et al. International standard problem No. 50 ATLAS test, SB-DVI-09: 50% (6-inch) break of DVI line of the APR1400-final integration report: NEA/CSNI/R(2012)6[R]. Paris: NEA/CSNI, 2012.
[11]	苏罡. 中国核能科技“三步走”发展战略的思考[J]. 科技导报,2016, 34(15): 33-41.
[12]	ZUBER N, WILSON G E, ISHII M, et al. An integrated structure and scaling methodology for severe accident technical issue resolution: development of methodology[J]. Nuclear Engineering and Design, 1998, 186(1-2): 1-21. doi: 10.1016/S0029-5493(98)00215-5
[13]	LI Y Q, HAO B T, ZHONG J, et al. Comparative experiments to assess the effects of accumulator nitrogen injection on passive core cooling during small break LOCA[J]. Nuclear Engineering and Technology, 2017, 49(1): 54-70. doi: 10.1016/j.net.2016.06.014
[14]	LI Y Q, CHANG H J, YE Z S, et al. Analyses of ACME integral test results on CAP1400 small-break loss-of-coolant-accident transient[J]. Progress in Nuclear Energy, 2016, 88: 375-397. doi: 10.1016/j.pnucene.2016.01.012
[15]	LI Y Q, YE Z S, ZHONG J, et al. Core makeup tank behavior investigation during ACME integral effect tests[J]. Nuclear Engineering and Design, 2020, 364: 110701. doi: 10.1016/j.nucengdes.2020.110701
[16]	FANG F F, YANG F M, HAO B T, et al. Experimental and numerical analyses of typical SBLOCAs on advanced core-cooling mechanism experiment[J]. Atomic Energy Science and Technology, 2017, 51(8): 1393-1399.
[17]	刘卓,常华健,阳祥,等. 非能动安全壳冷却系统空气流道自然循环的比例分析与设计[J]. 原子能科学技术,2016, 50(2): 254-260. doi: 10.7538/yzk.2016.50.02.0254
[18]	邓程程,常华健,秦本科,等. 比例试验台架中的储热分析[J]. 原子能科学技术,2013, 47(10): 1751-1759. doi: 10.7538/yzk.2013.47.10.1751
[19]	邓程程,常华健,秦本科,等. 非能动试验台架中CMT的比例分析及失真评价[J]. 原子能科学技术,2013, 47(11): 2026-2032. doi: 10.7538/yzk.2013.47.11.2026
[20]	DENG C C, CHANG H J, QIN B K, et al. Stored energy analysis in the scaled-down test facilities[J]. Annals of Nuclear Energy, 2016, 96: 19-25. doi: 10.1016/j.anucene.2016.05.018
[21]	ZHU S. Study on code modeling and test initialization condition method of ACME facility[J]. Atomic Energy Science and Technology, 2016, 50(7): 1179-1185.
[22]	USNRC. TRACE V5.0 theory manual field equations, solution methods, and physical models: NUREG/CR-5535/Rev P3-Vol I[R]. Washington, D.C: USNRC, 2010.