Application of DMRM on LOCA Integrated Effect Experiment

Huang Tao; Li Zhongchun; Sun Wei; Deng Jian; Ding Shuhua; Liu Yu; Wu Dan; Qian Libo; Shen Yaou; Du peng; Wu Zenghui

doi:10.13832/j.jnpe.2021.S2.0109

Volume 42 Issue S2

Dec. 2021

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Article Navigation > Nuclear Power Engineering > 2021 > 42(S2): 109-112

Huang Tao, Li Zhongchun, Sun Wei, Deng Jian, Ding Shuhua, Liu Yu, Wu Dan, Qian Libo, Shen Yaou, Du peng, Wu Zenghui. Application of DMRM on LOCA Integrated Effect Experiment[J]. Nuclear Power Engineering, 2021, 42(S2): 109-112. doi: 10.13832/j.jnpe.2021.S2.0109

Citation:

Huang Tao, Li Zhongchun, Sun Wei, Deng Jian, Ding Shuhua, Liu Yu, Wu Dan, Qian Libo, Shen Yaou, Du peng, Wu Zenghui. Application of DMRM on LOCA Integrated Effect Experiment[J]. Nuclear Power Engineering, 2021, 42(S2): 109-112. doi: 10.13832/j.jnpe.2021.S2.0109

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Application of DMRM on LOCA Integrated Effect Experiment

doi: 10.13832/j.jnpe.2021.S2.0109

Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, 610213, China

Received Date: 2021-07-19
Accepted Date: 2021-12-07
Rev Recd Date: 2021-11-17
Publish Date: 2021-12-29

Abstract

Abstract

China Nuclear Power Research and Design Institute independently developed the loss of coolant accident (LOCA) evaluation model of Deterministic Model-Realistic Method (DMRM), and developed the advanced reactor system analysis code ARSAC based on the model. In this paper, this method is applied to large-break LOCA conditions to verify the correctness and rationality of the method. In the verification process, firstly, a conservative bench system model is established through sensitivity analysis; On this basis, the input conditions (system pressure, safety injection tank water temperature, etc.) that have a great impact on the peak cladding temperature (PCT) are randomly sampled to form 59 groups of working conditions for calculation and analysis; These working conditions are analyzed by ARSAC to obtain the final results, and the final results are compared with the experimental data. The results show that the pressure and other parameters calculated by the code are consistent with the experimental data, which proves the correctness of the ARSAC calculation; the PCT temperature obtained by DMRM is significantly higher than the experimental value, which proves the conservativeness of DMRM.

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References

[1]	国家核安全局. 核动力厂安全分析用计算机软件开发与应用(试行): 国核安发[2017]323号[A]. 2017-12-20.
[2]	NRC. Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, Appendix K, ECCS evaluation models[S]. US: NRC, 1992.
[3]	BURTT J D. International standard problem 13 (loft experiment L2-5): EGG-NTAP-6276[R]. Idaho National Engineering Liboratory, 1983

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