SCWR堆芯稳态物理-热工水力耦合计算程序系统CASIR的开发  

Development of Coupled Neutronics/Thermal-Hydraulics CASIR Code System for SCWR Core Steady State

作　　者：马永强[1] 柴晓明[1] 王育威[1] 潘俊杰[1] 安萍[1] 

机构地区：[1]中国核动力研究设计院核反应堆系统设计技术国家级重点实验室,成都610041

出　　处：《核动力工程》2013年第1期87-91,共5页Nuclear Power Engineering

摘　　要：针对超临界水冷反应堆(SCWR)堆芯冷却剂密度沿轴向变化剧烈的特点,开发用于SCWR堆芯稳态物理-热工水力耦合计算的程序系统CASIR。CASIR由改进的压水堆堆芯中子学计算程序和适用于SCWR燃料组件计算的子通道热工-水力程序组成,具备调整堆芯下腔室入口流量分配的功能。针对CSR1000双流程的SCWR首循环堆芯,通过与蒙特卡罗程序对比寿期初时刻计算结果的方式,初步验证CASIR计算SCWR堆芯中子学问题的准确性;通过SCWR堆芯燃耗模拟,以及调整堆芯流量分布使得最大包壳表面温度(MCST)满足设计限值的测试,表明CASIR满足SCWR堆芯设计的要求,可应用于方形燃料组件的SCWR堆芯概念设计。A coupled neutronic and thermal-hydraulic code system CASIR （Core Analysis System for Innovation Reactor） for SCWR core steady state is developed due to the rapid variation of coolant density along the axial section in SCWR （supercritical water cooled reactor） core. The code system consists of the improved neutronic codes which are used for PWR （pressurized-water reactor） neutronic calculation and the sub-channel thermal-hydraulic code which is applicable lbr SCWR fuel assembly calculation, including the function of adjusting the flow rate distribution in the lower plenum of core. The coupled code system is tested on a two pass SCWR core of first cycle in the beginning of cycle by comparing the results with those calculated by Monte Carlo code, and it shows that the CASIR system is applicable to simulate the SCWR core preliminarily; main results of burn-up simulation are obtained by adjusting the core flow distribution in order to restrict the maximum cladding-surface temperature below the limit of design. All above show that the CASIR code system satisfies the requirements of SCWR core design and can be used for conceptual design of SCWR core with square assemblies.

关 键 词：超临界水冷反应堆 稳态物理-热工水力耦合 概念设计 CASIR 

分 类 号：TL364[核科学技术—核技术及应用]