ACR-700核电厂小破口失水事故分析  

Analysis of Small Break LOCA for ACR-700 NPP

作　　者：郑利民[1] 申森[1] 

机构地区：[1]上海核工程研究设计院CANDU工程中心,200233

出　　处：《核动力工程》2006年第z1期5-8,72,共5页Nuclear Power Engineering

摘　　要：针对加拿大AECL最新推出的ACR-700先进重水堆核电厂设计,建立CATHENA MOD3.5d重水堆热工水力系统分析程序的分析模型,并用该程序进行小破口失水事故下热传输系统和反应堆热工水力瞬态特性分析。主要分析重水堆核电厂对应反应堆入口集管、热传输泵吸入段及反应堆出口集管3种不同破口位置的典型的最不利事故工况,确定了导致最不利事故后果的破口面积,并给出主要的计算分析结果。Analysis of a small break loss of coolant accident(SB-LOCA) for the latest design of ACR-700 PHWR nuclear power plant(NPP) developed by Atomic Energy of Canada Limited has been per-formed with CATHENA MOD 3.5d,a PHWR system thermal-hydraulic analysis code based on the analysismodel established.The limiting break size has been found by performing the sensitivity study for three dif-ferent break locations [i.e.reactor inlet head(RIH),HTS pump suction(PS) pipe and reactor outlet head(ROH)] under the limiting case(i.e.SB-LOCA with subsequent loss of class Ⅵ power with all safety sys-tem available),and main analysis results were also provided.

关 键 词：ACR-700 重水堆 核电厂 小破口失水事故 

分 类 号：TL364+.4[核科学技术—核技术及应用]