Research on Inter-unit Common Cause Failure of Multi-unit PSA
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摘要: 合理评估同一厂址内不同机组间共因失效(CCF)对电厂安全风险的贡献是多堆概率安全分析(PSA)建模中需要解决的重要技术问题。本研究对设备机组间CCF组的选取、CCF建模和参数估计的方法进行梳理,并以具有4台机组的某压水堆核电厂丧失厂外电(LOOP)事件为分析案例,定量评估在考虑设备机组间CCF前后的多堆PSA模型堆芯损坏频率(CDF)变化情况。研究结果表明,在考虑设备机组间CCF后,多堆PSA结果中仅有一台机组发生堆芯损坏(CD)的频率和多台机组同时发生CD的频率均会有所增加。由此可见,机组间CCF对多堆PSA结果有一定影响。
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关键词:
- 多堆 /
- 概率安全分析(PSA) /
- 共因失效(CCF)分析 /
- 堆芯损坏频率(CDF)
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. -
表 1 EDG的CCF因子
Table 1. CCF Factors for EDG
CCF参数 启动失效 运行失效 二阶CCF β因子 8.33×10−3 1.41×10−2 三阶CCF β因子 6.48×10−3 1.25×10−2 三阶CCF γ因子 3.98×10−1 3.17×10−1 表 2 EDG的机组间CCF参数计算过程
Table 2. Parameters Calculation of Inter-unit CCF for EDG
共因组编号 描述 k值 r值 Rx 共因启动失效概率 共因运行失效概率 U12_EDG 1、2号机组间EDG的CCF 2.50×10−1 1.00×100 2.35×10−1 5.84×10−5 2.76×10−5 U13_EDG 1、 3号机组间EDG的CCF 1.25×10−1 5.00×10−1 4.79×10−2 1.19×10−5 1.77×10−5 U14_EDG 1、4号机组间EDG的CCF 1.25×10−1 5.00×10−1 4.79×10−2 1.19×10−5 1.77×10−5 U23_EDG 2、3号机组间EDG的CCF 1.25×10−1 5.00×10−1 4.79×10−2 1.19×10−5 1.77×10−5 U24_EDG 2、4号机组间EDG的CCF 1.25×10−1 5.00×10−1 4.79×10−2 1.19×10−5 1.77×10−5 U34_EDG 3、4号机组间EDG的CCF 2.50×10−1 1.00×100 2.35×10−1 5.84×10−5 8.69×10−5 U123_EDG 1、2、3号机组间EDG的CCF 1.25×10−1 5.00×10−1 6.84×10−3 5.25×10−7 7.09×10−8 U124_EDG 1、2、4号机组间EDG的CCF 1.25×10−1 5.00×10−1 6.84×10−3 5.25×10−7 7.09×10−8 U134_EDG 1、3、4号机组间EDG的CCF 1.25×10−1 5.00×10−1 6.84×10−3 5.25×10−7 7.09×10−8 U234_EDG 2、3、4号机组间EDG的CCF 1.25×10−1 5.00×10−1 6.84×10−3 5.25×10−7 7.09×10−8 U1234_EDG 4台机组间EDG的CCF 1.25×10−1 5.00×10−1 9.77×10−4 7.51×10−8 1.01×10−8 表 3 多堆CDF对比
Table 3. Comparison of Multi-unit CDF
同时发生CD的
机组数量/台CDF/(堆·年)−1 考虑设备机组间CCF 不考虑设备机组间CCF 4 5.80×10−15 1.00×10−16 3 6.10×10−10 4.36×10−14 2 2.61×10−6 7.84×10−8 1 7.52×10−5 7.32×10−5 合计 CDF值 7.78×10−5 7.33×10−5 -
[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.