摘要
用新研制的三维多群P3 中子输运蒙特卡罗程序MCMG ,通过与栅元均匀化WIMS程序耦合 ,计算反应堆临界 燃耗问题。高通量工程试验堆 (HFETR)临界计算取得了与MCNP程序和实验一致的结果 ,且在相同计算精度下 ,MCGM计算时间较MCNP计算时间少数倍。
A 3 D multigroup P 3 neutron transport Monte Carlo code MCMG was introducel in this paper.It is used to simulate the reactor critical burnup problems by coupling the lattice homogenization WIMS code.The critical calculation results of High Flux Engineering Test Reactors(HFETR) were consistent with MCNP results and experiments.With the same precision,the computational time of the MCMG code is several times shorter than that of the MCNP code.
出处
《核科学与工程》
CAS
CSCD
北大核心
2002年第1期27-31,共5页
Nuclear Science and Engineering
