摘要
采用通过编写连接MCNP程序和ORIGEN2程序的接口处理程序的方法进行快中子系统的燃耗计算。由MCNP、ORIGEN2、接口处理程序和截面文件组成的软件系统可用于燃料或堆芯非均匀布置快中子系统的燃料同位素成分和燃耗反应性损失计算,在燃耗反应性损失计算中采用了伪裂变产物的方法。介绍程序系统的研制情况,并给出用该软件系统计算中国实验快堆首炉堆芯和OECD/NEAMOX燃料快堆基准题的燃耗计算结果。
To calculate burnup for the cores with fast neutron spectrum, we have written the linking codes to process the input and output of MCNP code and ORIGEN2 code. The code package which includes MCNP, ORIGEN2, the linking codes and cross sections files can be used to calculate the fuel isotopes compositions and burnup reactivity loss for a fast reactor core with heterogeneous configuration. The pseudo fission products are adopted to calculate burnup reactivity loss. The burnup results of the first core of China Experimental Fast Reactor and the benchmark on MOX fuel fast reactor of OECD/NEA, which are calculated with the code package, are presented in this paper.
出处
《核科学与工程》
CSCD
北大核心
2005年第1期24-29,共6页
Nuclear Science and Engineering
基金
核工业科学基金资助项目的一部分
