CAP1400核电站非能动堆芯冷却系统性能试验验证
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摘要
CAP1400是我国在引进消化AP1000的基础上自主研发的大型非能动压水堆核电站,其安全系统采用非能动设计理念,由自然力/过程来驱动,在事故工况下执行安全功能时无需外部动力电源。相对于现有商用压水堆核电站采用的能动安全系统,非能动安全系统的设计原理发生了根本变化,两者的事故进程与物理现象也存在很大差异。此外,CAP1400较AP1000具有更高的堆芯功率,因此,CAP1400安全评审要求对非能动安全系统开展试验验证。本文对CAP1400非能动堆芯冷却系统性能验证试验台架ACME进行介绍。相比原有AP600/1000开展的试验,ACME台架的试验参数范围更广,测量更加精细。通过试验结果及分析,研究了非能动堆芯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重要的试验结果支撑。
CAP1400 has been developed as a new passive safety power generation III+ reactor based on AP600/1000 technology. The safety system of CAP1400 is driven by natural force or process. Compared with existing commercial PWR, the design of passive safety system has changed a lot as well as the accident process and physical phenomenon. The core power of CAP1400 is much bigger than AP1000. Therefore, verification experiments of passive safety system are required by CAP1400 safety review. A large-scale integrated test facility ACME was designed and built to achieve the PXS test verification of CAP1400. The parameter range and simulation similarity are improved compared to previous passive system tests of AP600/1000. Based on test results and analysis, accident response and critical physical phenomenon are studied and relevant relations are improved,providing support of test results to CAP1400 safety review and related safety analysis code validation.
CAP1400 has been developed as a new passive safety power generation III+ reactor based on AP600/1000 technology. The safety system of CAP1400 is driven by natural force or process. Compared with existing commercial PWR, the design of passive safety system has changed a lot as well as the accident process and physical phenomenon. The core power of CAP1400 is much bigger than AP1000. Therefore, verification experiments of passive safety system are required by CAP1400 safety review. A large-scale integrated test facility ACME was designed and built to achieve the PXS test verification of CAP1400. The parameter range and simulation similarity are improved compared to previous passive system tests of AP600/1000. Based on test results and analysis, accident response and critical physical phenomenon are studied and relevant relations are improved,providing support of test results to CAP1400 safety review and related safety analysis code validation.
引文
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[3] Wright R F, Hundal R, Hochreiter L E, et al. Analysis and evaluation of the AP600 SPES-2 integral systems tests[C].International Conference on Nuclear Engineering,1996, 2:417-425.
[4] Kukita, Yutaka, Yonomoto, et al. ROSA/AP600 testing:facility modifications and initial test results[J].Nuclear Science and Technology, 1996, 33:259-265.
[5] Reyes J N, Hochreiter L, Scaling analysis for the OSU AP600 test facility(APEX)[J]. Nuclear Engineering and Design, 1998, 186:53-109.
[6] Yu Quan Li, Hua Jian Chang, Zi Shen Ye, et al. Analyses of ACME integral test results on CAP1400 smallbreak loss-of-coolant-accident transient[J]. Progress in Nuclear Energy, 2016, 88:375-397.
[7] 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.
[8]陈炼,常华健,李玉全,等.ACME整体性能试验设施工作压力选取方案分析[J].原子能科学技术,2011,45(10):1215-1220.
[2]林诚格,郁祖盛,欧阳予,等.非能动安全先进压水堆核电技术(上)[M],北京,原子能出版社,2010.
[3] Wright R F, Hundal R, Hochreiter L E, et al. Analysis and evaluation of the AP600 SPES-2 integral systems tests[C].International Conference on Nuclear Engineering,1996, 2:417-425.
[4] Kukita, Yutaka, Yonomoto, et al. ROSA/AP600 testing:facility modifications and initial test results[J].Nuclear Science and Technology, 1996, 33:259-265.
[5] Reyes J N, Hochreiter L, Scaling analysis for the OSU AP600 test facility(APEX)[J]. Nuclear Engineering and Design, 1998, 186:53-109.
[6] Yu Quan Li, Hua Jian Chang, Zi Shen Ye, et al. Analyses of ACME integral test results on CAP1400 smallbreak loss-of-coolant-accident transient[J]. Progress in Nuclear Energy, 2016, 88:375-397.
[7] 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.
[8]陈炼,常华健,李玉全,等.ACME整体性能试验设施工作压力选取方案分析[J].原子能科学技术,2011,45(10):1215-1220.