ACME台架PRHR管线破口位置敏感性试验研究  

Experimental Study on Sensitivity of PRHR Pipeline Break Location on ACME Test Facility

作　　者：刘宇生 许超 吴鹏 王楠 李振啸 Liu Yusheng;Xu Chao;Wu Peng;Wang Nan;Li Zhenxiao(Fundamental Science on Nuclear Safety and Simulation Technology Laboratory,Harbin Engineering University,Harbin,150001,China;National Environmental Protection Key Laboratory for Simulation Analysis and Verification of Nuclear and Radiation Safety Review,Nuclear and Radiation Safety Center,MEE,Beijing,100082,China;State Nuclear Power Technology R&D Center,Beijing,102209,China)

机构地区：[1]哈尔滨工程大学核安全与仿真技术国防重点学科实验室,哈尔滨150001 [2]生态环境部核与辐射安全中心,国家环境保护核与辐射安全审评模拟分析与验证重点实验室,北京100082 [3]国核华清(北京)核电技术研发中心有限公司,北京102209

出　　处：《核动力工程》2021年第5期64-70,共7页Nuclear Power Engineering

基　　金：国家科技重大专项核动力厂安全分析用计算机软件评估基准题及共享平台开发(2019ZX06005001)。

摘　　要：为研究先进非能动(AP)型核电厂在非能动系统失效条件下的安全性能,利用我国先进堆芯冷却机理整体试验台架(ACME)开展了非能动余热排出(PRHR)管线破口失水试验研究,分析了主要的试验进程和破口位置对事故过程各阶段关键参数的影响。结果表明,ACME PRHR管线破口试验进程与冷管段小破口失水事故(SBLOCA)进程基本一致,再现了非能动核电厂自然循环阶段、自动卸压系统(ADS)喷放阶段和安全壳内置换料水箱(IRWST)安注阶段的安全特性;在不同破口位置的试验中,非能动堆芯冷却系统(PXS)均可保证堆芯得到补水,堆芯活性区始终处于混合液位以下;破口位置对ACME LOCA事故进程、反应堆冷却剂系统(RCS)初期降压速率、PRHR热交换器(HX)流量、喷放流量、堆芯液位、IRWST安注流量等参数具有显著影响,对堆芯补水箱(CMT)和蓄压安注箱(ACC)安注流量的影响较小。To study the safety performance of the advanced passive(AP)nuclear power plant(NPP)under passive system failure condition,the experimental study on the loss of coolant from the passive residual heat removal(PRHR)pipeline break is performed by the advanced core-cooling mechanism experiment(ACME)facility,during which the effect of main test sequences and break location on the key parameters in different phases of the accident is analyzed.As demonstrated by the study results,the ACME PRHR pipeline break test sequences,basically same as those for the small-break loss of coolant accident(SBLOCA)of the cold leg,reproduce the safety characteristics in the natural circulation phase of the passive NPP,blowout phase of the automatic depressurization system(ADS)and safety injection(SI)phase of the in-containment refueling water storage tank(IRWST);in the test at different break locations,the passive core cooling system(PXS)can always ensure the water makeup for core and the core active region remains below the mixed liquid level;and the break locations have notable effect on the ACME LOCA accident sequence,initial depressurization rate of reactor coolant system(RCS),PRHR heat exchanger(HX)flow,blowout flow,core level,IRWST SI flow and other parameters,and have little impact on the SI flow of the core makeup tank(CMT)and accumulator(ACC).

关 键 词：小破口失水事故(SBLOCA) 先进堆芯冷却机理整体试验台架(ACME)台架 整体效应试验 PRHR管线 

分 类 号：TL364.9[核科学技术—核技术及应用]