CAP1400核电站非能动安全壳冷却系统综合性能试验研究
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摘要
CAP1400非能动安全壳冷却系统(Passive Containment Cooling System,PCS)仍采用与AP600和AP1000相同的先进非能动设计理念。本研究针对CAP1400的非能动安全壳冷却系统,设计并建造了非能动安全壳冷却系统综合性能试验平台(Containment safety vErification via integRal Test facility,CERT),开展了试验研究。对CAP1400非能动安全壳冷却系统综合性能试验验证需求,试验装置的设计特点、研究内容及代表性的试验结果进行了介绍。通过PCS综合性能试验的开展,研究了非能动安全壳冷却系统的事故响应特性及关键物理现象,为CAP1400的安全评审及相关安全分析程序验证提供了试验结果支撑。
CAP1400 passive containment cooling system(PCS), which is still based on the design concept of the AP600/1000 passive safety technology. To verify the design of CAP1400 PCS, one large-scale integral test facility(CERT) was designed and built to carry out the PCS test verification of CAP1400. In this paper, the overall test verification requirements of CAP1400 passive containment cooling system, the design features of CERT facility, and the research contents and representative test results are introduced. Based on the test results and analysis, accident response and critical physical phenomenon are studied. The research results provide valuable support for CAP1400 safety review and related safety analysis code validation.
CAP1400 passive containment cooling system(PCS), which is still based on the design concept of the AP600/1000 passive safety technology. To verify the design of CAP1400 PCS, one large-scale integral test facility(CERT) was designed and built to carry out the PCS test verification of CAP1400. In this paper, the overall test verification requirements of CAP1400 passive containment cooling system, the design features of CERT facility, and the research contents and representative test results are introduced. Based on the test results and analysis, accident response and critical physical phenomenon are studied. The research results provide valuable support for CAP1400 safety review and related safety analysis code validation.
引文
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[2]Matzie R A. API000 will meet the challenges of near-term deployment[J]. Nuclear Engineering and Design, 2008,238:1856-1862.
[3]林诚格,郁祖盛,欧阳予,等,非能动安全先进压水堆核电技术[M],北京:原子能出版社,2010.
[4] Wang W. W, Su G H, Tian W X, et al. Research on thermal hydraulic behavior of small-break LOCAs in AP1000[J]. Nuclear Engineering and Design, 2013:380-394.
[5]孙汉虹,等.第三代核电技术AP1000[M].北京:中国电力出版社,2010.
[6] Zuber N,Wilson G E, Ishii M, et al. An integrated structure and scaling methodology for severe accident technical issue resolution:development of methodology[J].Nuclear Engineering and Design, 1998,186:1-21.
[7] Ruichang Zhao, Liuli Sun, Chengxin Cui,et al.Delayed-feedback artificial neural network applied in multiparameters thermal-hydraulic system performance assessing:An integral PCCS test case-study[J]. Progress in Nuclear Energy, 2016,90:27-36.