摘要
随着反应堆数值计算的精度要求不断提高,物理-热工耦合计算在核反应堆分析中成为研究热点,作为工程领域中的主流计算方法之一,研究适用于确定论两步法的物理-热工耦合计算方法具有明确的工程应用价值。选取两步法程序DRAGON/DONJON和子通道程序COBRA-EN,开发了基于两步法的物理-热工耦合计算系统,并采用美国CASL项目提出的VERA系列基准题中VERA#6和VERA#7验证了耦合系统的正确性。结果表明:VERA#6的keff的误差在100×10^(-5)以内,组件径向裂变率的相对误差在±1%范围内,燃料温度和冷却剂温度的分布趋势与参考值吻合良好;VERA#7的临界硼浓度的计算误差在20×10^(-6)以内,径向功率分布的均方根误差为0.86%,堆芯出口处冷却剂温度与参考值的误差在±5 K以内,验证了耦合系统的可靠性和准确性。
With the increasing requirements for the accuracy of numerical calculations in reactors,Omore and more attention has been paid to multi-physics coupling calculation in nuclear reactor analysis.As one of the main calculation methods in industry,the study of the neutronics and thermal-hydraulics coupling calculation method which is suitable for the deterministic two-step method has a clear value for industry.Based on the two-step codes DRAGON/DONJON and the subchannel code COBRA-EN,a neutronics and thermal-hydraulics coupling code system based on a unified framework has been developed,and the coupling code system has been validated around the Virtual Environment for Reactor Applications(VERA)series of benchmark problems 6 and 7,which were proposed by the Consortium for Advanced Simulation of LWRs(CASL),in the United States.The results show that the error of ker for Problem 6 is within 100×10^(-5),and the radial fission rate of the component is within±1%,the distribution trends of fuel temperature and coolant temperature agree well with reference values;While the computational error of the critical boron concentration for Problem 7 is within 20×10^(-6),and the root-mean-square error of the power distribution is 0.86%,the coolant temperature at the core outlet differs from the reference value within±5 K.The computational capability of the coupling code system is verified.
作者
袁显宝
陈浩铭
刘曾豪
张彬航
张永红
唐海波
YUAN Xianbao;CHEN Haoming;LIU Zenghao;ZHANG Binhang;ZHANG Yonghong;TANG Haibo(College of Science,China There Gorges University,Yichang of Hubei Prov.443002,China;College of Mechanical and Power Engineering,China There Gorges University,Yichang of Hubei Prov.443002,China)
出处
《核科学与工程》
北大核心
2025年第4期632-641,共10页
Nuclear Science and Engineering
基金
国家自然科学基金资助项目(No.12175116,No.12275153,No.12305187)。
作者简介
袁显宝(1974-),男,湖北兴山人,教授,博士,主要从事核反应堆物理;通讯作者:张彬航,E-mail:evanustc@mail.ustc.edu.cn。
