[1]赵海歌,郭诚湛.微型中子源核反应堆绝对中子通量密度计算[J].深圳大学学报理工版,2004,21(2):147-150.
 ZHAO Hai-ge and GUO Cheng-zhan.Calculation of absolute thermal neutron flux density in the core of miniature neutron source reactor[J].Journal of Shenzhen University Science and Engineering,2004,21(2):147-150.
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微型中子源核反应堆绝对中子通量密度计算()
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《深圳大学学报理工版》[ISSN:1000-2618/CN:44-1401/N]

卷:
第21卷
期数:
2004年2期
页码:
147-150
栏目:
物理
出版日期:
2004-04-30

文章信息/Info

Title:
Calculation of absolute thermal neutron flux density in the core of miniature neutron source reactor
文章编号:
1000-2618(2004)02-0147-04
作者:
赵海歌 郭诚湛
深圳大学核技术应用研究所, 深圳 518060
Author(s):
ZHAO Hai-ge and GUO Cheng-zhan
The Joint Institute of Applied Nuclear Technology Shenzhen University, Shenzhen 518060, P. R. China
关键词:
微堆 绝对中子通量密度 数值算法
Keywords:
MNSR absolute neutron flux density numerical algorithm
分类号:
TL 362.1; YL 364.4
文献标志码:
A
摘要:
阐述微型中子源核反应堆绝对中子通量密度测量的数值算法机理, 并给出其计算公式.结合深圳大学微堆实例给出计算结果, 与实际中子通量密度对比, 说明应用该方法时的注意事项.
Abstract:
Numerical analysis about the measurement of absolute thermal neutron flux density in the core of Miniature Neutron Source  Reactor(MNSR) was illustrated in detail. The calculation formula was deduced, and the program frame was provided. The algorithm was applied to measure the flux density of Shenzhen University MNSR, and the result was compared with actual neutron flux density. At last, some points of attention were mentioned.
更新日期/Last Update: 2015-11-06