PSA-Based Uncertainty Analysis of Pressurized Water Reactor LBLOCA

Deng Jian; Xiong Qingwen; Gou Junli; Liu Yu; Bao Hui; Shen Danhong; Zhou Jiayue

doi:10.13832/j.jnpe.2022.05.0188

Volume 43 Issue 5

Oct. 2022

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Article Navigation > Nuclear Power Engineering > 2022 > 43(5): 188-194

Deng Jian, Xiong Qingwen, Gou Junli, Liu Yu, Bao Hui, Shen Danhong, Zhou Jiayue. PSA-Based Uncertainty Analysis of Pressurized Water Reactor LBLOCA[J]. Nuclear Power Engineering, 2022, 43(5): 188-194. doi: 10.13832/j.jnpe.2022.05.0188

Citation:

Deng Jian, Xiong Qingwen, Gou Junli, Liu Yu, Bao Hui, Shen Danhong, Zhou Jiayue. PSA-Based Uncertainty Analysis of Pressurized Water Reactor LBLOCA[J]. Nuclear Power Engineering, 2022, 43(5): 188-194. doi: 10.13832/j.jnpe.2022.05.0188

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PSA-Based Uncertainty Analysis of Pressurized Water Reactor LBLOCA

doi: 10.13832/j.jnpe.2022.05.0188

1.
Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, 610213, China
2.
School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049, China

Received Date: 2021-09-07
Accepted Date: 2022-02-14
Rev Recd Date: 2022-01-18
Publish Date: 2022-10-12

Abstract

Abstract

In order to combine deterministic and probabilistic analysis to carry out more realistic safety analysis of nuclear reactor accident conditions, a method combining probabilistic safety analysis (PSA) and best estimate plus uncertainty (BEPU) analysis is proposed, the extreme accident of the double-ended fracture large-break loss of water accident (LBLOCA) in a typical three-loop pressurized water reactor cold pipe section is taken as the object. First, the accident failure analysis of the emergency core cooling system is carried out based on PSA. Then, combined with BEPU analysis, the cladding peak temperature (PCT) distribution and conditional core damage probability (CCDP) of each accident sequence in the event tree are evaluated, and the core damage frequency of the PWR in this accident condition is finally determined. The analysis results show that the emergency core cooling system of PWR can ensure the safety of the reactor under the condition of double-ended fracture of cold pipe section, and a row of low-pressure safety injection systems are sufficient to remove the residual heat of the core and ensure the safety of the reactor.
- Large-break loss of coolant accident (LBLOCA),
- Uncertainty analysis,
- Probabilistic safety assessment (PSA),
- Risk-informed

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References(18)

References

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