压力管式超临界水堆极限事故下“无堆芯熔化”概念评估  

Assessment of No-Core-Melt Concept for Pressure Tube Supercritical Water Cooled Reactors under Extreme Accidents

作　　者：吴攀[1] 任彦昊 单建强[1] 黄彦平[2] Wu Pan;Ren Yanhao;Shan Jianqiang;Huang Yanping(School of Energy and Power Engineering,Xi’an Jiaotong University,Xi’an,710049,China;CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology,Nuclear Power Institute of China,Chengdu,610213,China)

机构地区：[1]西安交通大学核科学与技术学院,西安710049 [2]中国核动力研究设计院中核核反应堆热工水力技术重点实验室,成都610213

出　　处：《核动力工程》2021年第5期156-161,共6页Nuclear Power Engineering

基　　金：国家重点研发计划(2018YFE0116100)。

摘　　要：在自主研发的事故分析程序SCTRAN的基础上,开发并验证了二维导热模型和辐射换热模型,并将改进后的SCTRAN应用于加拿大压力管式超临界水堆在失水事故(LOCA)叠加丧失紧急堆芯冷却系统(LOECC)事故中的堆芯安全评估,并对燃料棒到慢化剂之间的传热效率以及关键的影响因素进行了评估。计算结果表明,在LOCA叠加LOECC工况下,燃料棒到燃料通道的辐射换热和燃料棒到蒸汽的自然对流换热能够有效导出反应堆的衰变余热,最高功率的燃料组件内、外圈燃料棒的最高包壳温度分别为1278℃和1192℃,均低于不锈钢包壳的熔化温度,因此整个事故过程中不会发生堆芯熔化。This paper develops and verifies the two-dimensional(2D)heat conduction model and radiation heat transfer model based on the self-developed accident analysis code SCTRAN,applies the improved SCTRAN code to the core safety assessment of the Canadian pressure tube supercritical water cooled reactor(PT-SCWR)under the loss-of-coolant accident(LOCA)plus loss of emergency core cooling system(LOECC)accident,and assesses the heat transfer efficiency between the fuel rods and moderator and the key factors.The assessment results show that the residual decay heat of the reactor can be effectively removed by the radiation heat transfer from the fuel rods to fuel channels and the natural convection heat exchange from the fuel rods to steam and that the maximum cladding temperatures of the fuel rods in inner and outer rings of the fuel assembly at the maximum power are 1,278°C and 1,192°C,respectively,which are below the stainless steel cladding melting temperature.Therefore,no core meltdown occurs throughout the accident.

关 键 词：失水事故 压力管式超临界水堆(PT-SCWR) 无堆芯熔化 辐射换热 二维导热 

分 类 号：TL48[核科学技术—核技术及应用]