Research on Inter-unit Common Cause Failure of Multi-unit PSA

Yang Chunju; Wang Ming; Lin Modi; Zhang Bing; Wang Jinkai

doi:10.13832/j.jnpe.2024.02.0171

Volume 45 Issue 2

Apr. 2024

Turn off MathJax

Article Contents

Article Navigation > Nuclear Power Engineering > 2024 > 45(2): 171-177

Yang Chunju, Wang Ming, Lin Modi, Zhang Bing, Wang Jinkai. Research on Inter-unit Common Cause Failure of Multi-unit PSA[J]. Nuclear Power Engineering, 2024, 45(2): 171-177. doi: 10.13832/j.jnpe.2024.02.0171

Citation:

Yang Chunju, Wang Ming, Lin Modi, Zhang Bing, Wang Jinkai. Research on Inter-unit Common Cause Failure of Multi-unit PSA[J]. Nuclear Power Engineering, 2024, 45(2): 171-177. doi: 10.13832/j.jnpe.2024.02.0171

Citation:

PDF( 2307 KB)

Research on Inter-unit Common Cause Failure of Multi-unit PSA

doi: 10.13832/j.jnpe.2024.02.0171

State Key Laboratory of Nuclear Power Safety Technology and Equipment, China Nuclear Power Engineering Co., Ltd., Shenzhen, Guangdong, 518172, China

Received Date: 2023-05-23
Rev Recd Date: 2024-01-15
Publish Date: 2024-04-12

Abstract

Abstract

How to reasonably estimate the contribution of inter-unit common cause failure (CCF) to the safety risk of power plant is an important technical problem that needs to be solved in multi-unit probabilistic safety analysis (PSA) modeling. In this study, analysis methods of inter-unit CCF group selection, modeling and parameter estimation are presented, and the loss of offsite power (LOOP) event of a four-unit pressurized water reactor (PWR) nuclear power plant is taken as an analysis case to quantitatively evaluate the change of core damage frequency (CDF) of multi-unit PSA model before and after considering CCF. The results show that after considering the inter-unit CCF, the frequency of core damage (CD) occurrence in only single unit and the frequency of CD occurrence in multiple units will increase. It can be seen that inter-unit CCF has a certain effect on multi-unit PSA results.
- Multi-unit,
- Probabilistic safety analysis (PSA),
- Common cause failure (CCF) analysis,
- Core damage frequency (CDF)

FullText(HTML)

References(14)

References

[1]	Pickard-Lowe and Garrick, Inc. Seabrook station probabilistic safety assessment: PLG-0300[R]. Irvine: Pickard-Lowe and Garrick, Inc., 1983.
[2]	LE DUY T D, VASSEUR D. A practical methodology for modeling and estimation of common cause failure parameters in multi-unit nuclear PSA model[J]. Reliability Engineering & System Safety, 2018, 170: 159-174.
[3]	KIM D S, HAN S H, PARK J H, et al. Multi-unit level 1 probabilistic safety assessment: approaches and their application to a six-unit nuclear power plant site[J]. Nuclear Engineering and Technology, 2018, 50(8): 1217-1233. doi: 10.1016/j.net.2018.01.006
[4]	International Atomic Energy Agency. Multi-unit probabilistic safety assessment: Safety Reports Series No. 110[R]. IAEA: Vienna, 2023.
[5]	KIM D S, JIN H P, LIM H G. A pragmatic approach to modeling common cause failures in multi-unit PSA for nuclear power plant sites with a large number of units[J]. Reliability Engineering & System Safety, 2020, 195: 106739.
[6]	ZHANG S, TONG J J, ZHAO J. An integrated modeling approach for event sequence development in multi-unit probabilistic risk assessment[J]. Reliability Engineering & System Safety, 2016, 155: 147-159.
[7]	ZHANG S, TONG J J. Treating common-cause failures in multi-unit PRAs[C]//The 2017 International Topical Meeting on Probabilistic Safety Assessment and Analysis (PSA 2017). Pittsburgh: ANS, 2017.
[8]	何劼,刘涛,张忞隽,等. 多机组核电厂总体风险的一级PSA方法研究[J]. 原子能科学技术,2014, 48(5): 867-871.
[9]	MOSLEH A, FLEMING K N, PARRY G W, et al. Procedures for treating common cause failures in safety and reliability studies: NUREG/CR-4780, EPRI-NP-5613[R]. Washington: Nuclear Regulatory Commission, 1989.
[10]	ASME/ANS. Addenda to ASME/ANS RA-S-2008, Standard for level 1/large early release frequency probabilistic risk assessment for nuclear power plant applications: ASME/ANS RA-Sb-2013[S]. New York, U.S.: The American Society of Mechanical Engineers & American Nuclear Society, 2013: 19.
[11]	Nuclear Energy Institute. 10 CFR 50.69 SSC categorization guideline: NEI 00-04[R]. Nuclear Energy Institute: Washington, D.C., 2005.
[12]	US Nuclear Regulatory Commission. Common-cause failure database and analysis systems: event data collection, classification, and coding: NUREG/CR-6268[R]. Nuclear Regulatory Commission: Washington, D.C., 2007.
[13]	US Nuclear Regulatory Commission. CCF parameter estimations (2015 Update): NUREG/CR-5497[R]. Nuclear Regulatory Commission, Washington: D.C., 2016.
[14]	国家核安全局. 中国核电厂设备可靠性数据报告[R]. 北京: 国家核安全局,2016.