Monte Carlo transport and burnup calculation

被引：0
|
作者：
Deng, L
Xie, ZS
Li, S
机构：
[1] Inst Appl Phys & Computat Math, Lab Computat Phys, Beijing 100088, Peoples R China
[2] Xian Jiaotong Univ, Dept Nucl Engn, Xian 710049, Peoples R China
来源：
ANNALS OF NUCLEAR ENERGY | 2003年 / 30卷 / 01期
关键词：
D O I：
暂无
中图分类号：
TL [原子能技术]; O571 [原子核物理学];
学科分类号：
0827 ; 082701 ;
摘要：
A 3-D multigroup P-3 neutron transport Monte Carlo code MCMG-BURN is developed for coupling neutron transport with burnup. MCMG-BURN code is based on Monte Carlo code MCNP with the continuous energy cross-section and the reactor lattice code WIMS. It uses the up-front multigroup macroscopic cross-section library based on burnup for Monte Carlo calculations, The almost consistent results with the experiments have been achieved. (C) 2002 Elsevier Science Ltd. All rights reserved.
引用
收藏
页码：127 / 132
页数：6
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