Simulation of typical accidents in fuel test loop of CMRR

Hu Bo; Guo Simao; Wang Guanbo; Qian Dazhi; Guo Yuchuan; Yu Heng

doi:10.11884/HPLPB201931.190023

Volume 31 Issue 9

Sep. 2019

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Hu Bo, Guo Simao, Wang Guanbo, et al. Simulation of typical accidents in fuel test loop of CMRR[J]. High Power Laser and Particle Beams, 2019, 31: 096001. doi: 10.11884/HPLPB201931.190023

Citation:

Hu Bo, Guo Simao, Wang Guanbo, et al. Simulation of typical accidents in fuel test loop of CMRR[J]. High Power Laser and Particle Beams, 2019, 31: 096001. doi: 10.11884/HPLPB201931.190023

Hu Bo, Guo Simao, Wang Guanbo, et al. Simulation of typical accidents in fuel test loop of CMRR[J]. High Power Laser and Particle Beams, 2019, 31: 096001. doi: 10.11884/HPLPB201931.190023

Citation:

Hu Bo, Guo Simao, Wang Guanbo, et al. Simulation of typical accidents in fuel test loop of CMRR[J]. High Power Laser and Particle Beams, 2019, 31: 096001. doi: 10.11884/HPLPB201931.190023

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Simulation of typical accidents in fuel test loop of CMRR

doi: 10.11884/HPLPB201931.190023

Institute of Nuclear Physics and Chemistry, CAEP, Mianyang 621900, China

Received Date: 2019-01-24
Rev Recd Date: 2019-05-24
Publish Date: 2019-09-15

Abstract

Abstract

In order to reflect the behavior of the nuclear materials in the operating reactor veritably and roundly, and to supply reliable data for the safety review of the nuclear materials, the most effective solution is building a fuel test loop (FTL) based on the real pressurized water reactor in the research reactor. To ensure the safety of the fuel test loop and the research reactor, it is necessary to analyze the safety of the fuel test loop during accident transient. This article simulates and analyzes two typical kinds of accidents—small break loss-of-coolant accidents (LOCA) and loss-of-flow accidents (LOFA)—based on the origin design of the fuel test loop in China's Mianyang Research Reactor (CMRR). The peak cladding temperature in cold leg SBLOCA and hot leg small break LOCA is 880.6 ℃ and 367.6 ℃ respectively, which is under the critical value (1204 ℃). During the total LOFA, the MDNBR is greater than 1.5, which means there is no departure from nucleate boiling. In the clamp shaft accident, the cladding temperature is 734.1 ℃, which is less than 1482 ℃. The results obtained are consistent with the safety criteria.
- fuel test loop,
- loss-of-coolant accident,
- loss-of-flow accident,
- China's Mianyang Research Reactor

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References(10)

References

[1]	程作用, 张劲松, 李忠宪, 等. 500 kW考验回路水中F-和Cl-的低压离子色谱分析[J]. 核动力工程, 1996, 17(5): 477-480. https://www.cnki.com.cn/Article/CJFDTOTAL-HDLG605.016.htm Cheng Zuoyong, Zhang Jingsong, Li Zhongxian, et al. Analysis of F- and Cl- ions in the primary circuit water of 500 kW test loop by low pressure ion chromatography. Nuclear Power Engineering, 1996, 17(5): 477-480 https://www.cnki.com.cn/Article/CJFDTOTAL-HDLG605.016.htm
[2]	Hadjam A, Souidi F, Loubar A, et al. Simulation of a LBLOCA in the CALLISTO test facility using the best estimate computer code RELAP5/SCDAP3.2[J]. Nuclear Engineering and Design, 2013, 262: 153-167. doi: 10.1016/j.nucengdes.2013.03.052
[3]	Choo K N, Cho M S, Kim B G, et al. Material irradiation at Hanaro, Korea[C]//Research Reactor application for Materials under High Neutron Fluence. 2011: 50-51.
[4]	张毅. CARR高温高压试验回路事故分析[D]. 北京: 中国原子能科学研究院, 2007: 23-28. Zhang Yi. Accidents analysis of high temperature and high pressure test loop of CARR. Beijing: China Institute of Atomic Energy, 2007: 23-28
[5]	Bae H, Kim D E, Yi S J, et al. Comparison of three SBLOCA tests with different break locations using the SMART-ITL facility to estimate the safety of the SMART design[J]. Nuclear Engineering and Technology, 2017, 49(5): 968-978. doi: 10.1016/j.net.2017.04.006
[6]	于平安, 朱瑞安, 喻真烷, 等. 核反应堆热工分析[M]. 西安: 西安交通大学出版社, 1979: 268-269. Yu Ping'an, Zhu Ruian, Yu Zhenwan, et al. Thermal analysis of nuclear reactors. Xi'an: Xi'an Jiaotong University Press, 1979: 268-269
[7]	张学学. 热工基础[M]. 北京: 高等教育出版社, 2006. Zhang Xuexue. Thermal engineering foundation. Beijing: Higher Education Press, 2006
[8]	The RELAP5-3D Code Development Team. RELAP5-3D Code Manual Volume1-5[M]. Idaho: INEEL, 2005.
[9]	朱继洲, 奚树人, 单建强, 等. 核反应堆安全分析[M]. 西安: 西安交通大学出版社, 2004: 96-102. Zhu Jizhou, Xi Shuren, Shan Jianqiang, et al. Safety analysis of nuclear reactors. Xi'an: Xi'an Jiaotong University Press, 2004: 96-102
[10]	余冀阳, 余尔俊. 核电厂事故分析[M]. 北京: 清华大学出版社, 2012: 23-24. Yu Jiyang, Yu Erjun. Accidents analysis of nuclear plants. Beijing: Tsinghua University Press, 2012: 23-24