非能动堆芯冷却系统LOCA下冷却能力分析
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摘要
本文基于机理性分析程序建立了包括反应堆一回路冷却剂系统、专设安全设施及相关二次侧管道系统的先进压水堆分析模型,对典型的小破口失水事故和大破口失水事故开展了全面分析。针对不同破口尺寸、破口位置的失水事故,分析了非能动堆芯冷却系统(PXS)中非能动余热排出系统(PRHRS)、堆芯补水箱(CMT)、安注箱(ACC)、自动卸压系统(ADS)和安全壳内置换料水箱(IRWST)等关键系统的堆芯注水能力和冷却效果。研究表明,虽然破口尺寸、破口位置会影响事故进程发展,但所有事故过程中燃料包壳表面峰值温度不超过1 477K,且反应堆堆芯处于有效淹没状态。PXS能有效排出堆芯衰变热,将反应堆引导到安全停堆状态,防止事故向严重事故发展。
The analysis model for advanced pressurized water reactor was established by using mechanism analytical code.The model included reactor coolant system,engineering safety features and related secondary side pipes system.The typical small break loss of coolant accident and large break loss of coolant accident were selected to analyze the accident progression.The water injection capacity and cooling capacity of passive residual heat removal system(PRHRS),core makeup tank(CMT),accumulator(ACC),automatic depressurization system(ADS)and in-containment reactor water storage tank(IRWST)included in passive core cooling system(PXS)were focused on for LOCA with different sizes and locations.The results show that the size and location of break have an influence on the accident progression.But the peak cladding surface temperature does not exceed 1 477 Kand the reactor core is flooded underwater in all the accident conditions.The PXS can effectively remove reactor core decay heat and keep the reactorin the safe shutdown situation to prevent severe accident.
The analysis model for advanced pressurized water reactor was established by using mechanism analytical code.The model included reactor coolant system,engineering safety features and related secondary side pipes system.The typical small break loss of coolant accident and large break loss of coolant accident were selected to analyze the accident progression.The water injection capacity and cooling capacity of passive residual heat removal system(PRHRS),core makeup tank(CMT),accumulator(ACC),automatic depressurization system(ADS)and in-containment reactor water storage tank(IRWST)included in passive core cooling system(PXS)were focused on for LOCA with different sizes and locations.The results show that the size and location of break have an influence on the accident progression.But the peak cladding surface temperature does not exceed 1 477 Kand the reactor core is flooded underwater in all the accident conditions.The PXS can effectively remove reactor core decay heat and keep the reactorin the safe shutdown situation to prevent severe accident.
引文
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[11]Appendix K to 10CFR50-ECCS evaluation models[S].Washington D.C.:US NRC,1992.
[12]WINTERS J W,VIJUK R P,CUMMINS W E.AP1000 design control document,Version 19[R].Pittsburgh,US:Westinghouse Electric Company LLC,2004.
[2]SCHULZ T L.Westinghouse AP1000advanced passive plant[J].Nuclear Engineering and Design,2006,236(14-16):1 547-1 557.
[3]AKIRA O,HAJIME A,YOSHIO M.Applicability of REFLA/TRAC code to a small-beak LOCA of PWR[J].Journal of Nuclear Science and Technology,1995,32(3):245-256.
[4]FERNG Y M,LEE C H.Numerical simulation of natural circulation experiments conducted at the IIST facility[J].Nuclear Engineering and Design,1994,148(1):119-128.
[5]CHUNG M S.Simulation of SBLOCA based on an improved choked flow model for RELAP5/MOD3code[J].Annals of Nuclear Energy,2005,32(9):913-924.
[6]莫小锦,庄亚平,佟立丽,等.AP1000主给水管道断裂事故中PRHR系统冷却能力分析[J].原子能科学技术,2012,46(增刊):309-313.MO Xiaojin,ZHUANG Yaping,TONG Lili,et al.Analysis on PRHR system cooling capacity during feedwater system pipe break accident for AP1000[J].Atomic Energy Science and Technology,2012,46(Suppl.):309-313(in Chinese).
[7]莫小锦,佟立丽,曹学武.AP1000丧失正常给水事故PRHR冷却能力研究[J].科技导报,2012,30(21):26-29.MO Xiaojin,TONG Lili,CAO Xuewu.The PRHR cooling capacity during the fail of normal feedwater flow accident for AP1000[J].Science and Technology Review,2012,30(21):26-29(in Chinese).
[8]ZOU J,LI Q,TONG L L,et al.Assessment of passive residual heat removal system cooling capacity[J].Progress in Nuclear Energy,2014,70:159-166.
[9]ANALYTIS G T.Developmental assessment of RELAP5/MOD3.1with separate effect and integral test experiments:Model changes and options[J].Nuclear Engineering and Design,1996,163(1-2):125-148.
[10]10CFR50.46 Acceptance criteria for emergency core cooling systems for light water nuclear power reactors[S].Washington D.C.:US NRC,1992.
[11]Appendix K to 10CFR50-ECCS evaluation models[S].Washington D.C.:US NRC,1992.
[12]WINTERS J W,VIJUK R P,CUMMINS W E.AP1000 design control document,Version 19[R].Pittsburgh,US:Westinghouse Electric Company LLC,2004.