摘要
本文介绍了开发的蒙特卡罗燃耗计算程序MCNTRANS。MCNTRANS的中子学计算参数直接采用MCNP5程序的反应率计算值,燃耗计算方法采用图论算法跟踪燃耗链,同时,对实际燃耗过程进行详细分析以提高计算精度与程序适用性,并使用预估-校正方法以获取较大的燃耗计算步长。程序计算结果通过OECD/NEA与JAERI燃耗基准题实验结果进行验证,并与其他程序的计算结果进行比较。结果表明,MCNTRANS程序在不同燃耗深度下的计算结果和实验值与其他程序的计算值符合较好,部分锕系核素与裂变产物的计算精度更高。
A new nuclear fuel burnup calculation code MCNTRANS based on MCNP was introduced in this paper .The neutronics calculation parameter was extracted from the MCNP5 reaction rate tally result ,while a graph theory algorithm was implemented to track the burnup chain and the analytic solution of the Bateman equation was given . At the same time ,the detailed physical process was considered to improve the accuracy and serviceability of this code ,and prediction-correction method was used to allow a large burnup step .The OECD/NEA and JAERI pin cell benchmark problems were used to validate the code MCNTRANS while a reference result was given by other code .It can be concluded that the calculation results of MCNTRANS are generally consistent with the experimental result and that of the other burnup codes , and part of the actinides and fission products calculation result show better accuracy .
出处
《原子能科学技术》
EI
CAS
CSCD
北大核心
2013年第10期1824-1828,共5页
Atomic Energy Science and Technology
