SCWR-FQT回路的热工物理耦合分析  

Coupled thermal-physics analysis of SCWR-FQT loop

作　　者：汪子迪 曹臻[1] 刘晓晶[1] 程旭[1] 

机构地区：[1]上海交通大学核科学与工程学院,上海200240

出　　处：《核科学与工程》2016年第2期178-184,共7页Nuclear Science and Engineering

基　　金：国家自然科学基金(51106097)

摘　　要：中欧核能合作研究项目超临界水堆燃料验证实验(SCWR-FQT)的主要研究内容为在超临界水环境下对一个小型燃料组件进行堆内性能分析和验证。本文应用修过后的系统程序ATHLET-SC对该实验回路进行建模,同时结合堆芯中子物理的计算结果,对由于压力管进口管破裂形成的失水事故进行热工水力和中子物理的耦合分析,并讨论了物理耦合中停堆棒的负反应性、冷却剂温度系数等参数对结果的影响。计算结果表明,进行了中子物理耦合的结果得到的最高包壳温度比未进行中子耦合的结果要低15℃,同时停堆棒引入的负反应性是该事故过程中影响燃料棒最高包壳温度的一个主要因素。The main purpose of the Sino-Euro corporation proj ect Supercritical Water Reactor Fuel Qualification Test （SCWR-FQT）is to analyze and verify a supercritical water-cooled experiment loop containing a small scale fuel assembly. The modified system code ATHLET-SC was applied to model this loop and perform the coupled thermal-hydraulics and neutron-physics calculation analysis of the loss of coolant accident induced by the coolant inlet pipe break with neutron data. The effects of some important parameters such as external reactivity, reactivity coefficient for fluid temperature are also investigated in this paper. The results indicate that the peaking cladding temperature of the coupled calculation is 1 5 ℃ lower than the uncoupled calculation. And the external reactivity caused by the shut-down rods is a main factor affecting the peaking cladding temperature during the accident.

关 键 词：超临界水堆 中子耦合 LOCA事故 事故分析 

分 类 号：TL333[核科学技术—核技术及应用]