ACME整体试验台架堆芯模拟体设计
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摘要
为完成CAP1400非能动堆芯冷却系统试验验证,设计并建造了1/3高度比例的ACME整体试验台架.针对整体试验中堆芯处的关键热工水力过程进行比例分析,确定堆芯模拟体设计所需满足的相似准则,并在综合考虑加热棒制造工艺参数等因素后,通过不同方案比较确定了ACME堆芯模拟体加热棒参数及整体几何排布方案.为在试验运行过程中有效支撑加热棒,采用了具有梅花孔支撑形式的支撑板,并通过阻力计算确定支撑板流通孔开孔结构尺寸,以匹配台架堆芯阻力设计要求.结果表明:ACME试验台架堆芯模拟体的设计通过了加工制造、组装、阻力测试和试验运行各环节的验证,其设计满足整体试验要求.
To experimentally test the passive core cooling system of CAP1400 nuclear power unit,an integral test facility,called ACME,was designed and constructed as a 1/3height scale model of CAP1400,based on which thermal hydrualic scaling analysis was conducted on the reactor core,so as to determine the similarity criteria for design of the core simulator matching the integral test facility,and to find the optimal design specifications and geometric arrangement of the heater rods contained in the ACME core simulator,by comparing different design schemes and comprehensively considering relevant manufacturing processes and engineering constraints.To support the heater rods against dynamic load induced by turbulent flow during test operation,support plates with quatrefoil supporting holes were designed and installed,where drain holes were also punched,and the dimension of drain holes was determined through resistance calculation,so as to meet the requirements of flow resistance similarity.Results show that the design of ACME core simulator has been successfully verified in processes of manufacturing,assembling,pressure drop calibration and test operation,indicating that the design of reactor core simulator can meet the requirements of integral effect test.
To experimentally test the passive core cooling system of CAP1400 nuclear power unit,an integral test facility,called ACME,was designed and constructed as a 1/3height scale model of CAP1400,based on which thermal hydrualic scaling analysis was conducted on the reactor core,so as to determine the similarity criteria for design of the core simulator matching the integral test facility,and to find the optimal design specifications and geometric arrangement of the heater rods contained in the ACME core simulator,by comparing different design schemes and comprehensively considering relevant manufacturing processes and engineering constraints.To support the heater rods against dynamic load induced by turbulent flow during test operation,support plates with quatrefoil supporting holes were designed and installed,where drain holes were also punched,and the dimension of drain holes was determined through resistance calculation,so as to meet the requirements of flow resistance similarity.Results show that the design of ACME core simulator has been successfully verified in processes of manufacturing,assembling,pressure drop calibration and test operation,indicating that the design of reactor core simulator can meet the requirements of integral effect test.
引文
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[2]FRIEND M T,WRIGHT R F,HUNDAL R,et al.Simulated AP600 response to small-break loss-ofcoolant-accident and non-loss-of-coolant-accident events:anlaysis of SPES-2integral test results[J].Nuclear Technology,1998,122(1):19-42.
[3]SHOTKIN L M,KUKITA Y.Implications of the ROSA/AP600high-and intermediate-pressure test results[J].Nuclear Technology,1997,119(3):217-234.
[4]PURHONEN H,PUUSTINEN M,RIIKONEN V,et al.PACTEL integral test facility-description of versatile applications[J].Annals of Nuclear Energy,2006,33(11/12):994-1009.
[5]KIM Y S,CHOI K Y,PARK H S,et al.Commissioning of the ATLAS thermal-hydraulic integral test facility[J].Annals of Nuclear Energy,2008,35(10):1791-1799.
[6]郑开云,葛磊,陈功铭,等.AP1000设计基准事故试验热冲击过程数值模拟[J].动力工程学报,2014,34(1):25-31.ZHENG Kaiyun,GE Lei,CHEN Gongming,et al.Numerical simulation on thermal shock process in AP1000DBA tesing[J].Journal of Chinese Society of Power Engineering,2014,34(1):25-31.
[7]ISHII M,KATAOKA I.Scaling laws for thermal-hydraulic system under single phase and two-phase natural circulation[J].Nuclear Engineer and Design,1984,81(3):411-425.
[8]AGBODEMEGBE V Y,XU Cheng,AKAHO E H K,et al.Correlation for cross-flow resistance coefficient using STAR-CCM+simulation data for flow of water through rod bundle supported by spacer grid with split-type mixing vane[J].Nuclear Engineering and Design,2015,285(15):134-149.
[9]华绍曾,杨学宁.实用流体阻力手册[M].北京:国防工业出版社,1985.
[10]WULFF W,ROHATGI U S.System scaling for the Westinghouse AP600 pressurized water reactor and related test facilities[R].Washington,US:US Nuclear Regulatory Commission,1998.