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温度相关核截面数据库在MCNP计算中的必要性研究
引用本文:柴晓明,王侃,余纲林. 温度相关核截面数据库在MCNP计算中的必要性研究[J]. 原子核物理评论, 2006, 23(2): 111-114. DOI: 10.11804/NuclPhysRev.23.02.111
作者姓名:柴晓明  王侃  余纲林
作者单位:清华大学工程物理系,北京100084
摘    要:MCNP程序由于其几何模拟和核数据上的优越性,现在在反应堆的研究分析中已经得到较多应用。通过基准题的计算,定量地说明MCNP通过其自带的常温(294K)下的核素截面数据库不能够对反应堆进行非常准确的计算(由于反应堆内各种材料/位置的温度不同),而且,它也不能够计算反应堆中与温度相关的量,如反应性温度系数。选用了一个带有不同温度下核素截面数据的MCNP输入格式的数据库,使用MCNP-4C对基准题进行了计算,发现计算结果与基准值符合得非常好。这说明通过使用不同温度下的核素截面数据库,MCNP可以准确计算温度系数和增殖系数等,从而说明在反应堆设计计算中制作不同温度下的核素截面库的必要性。Due to the advantage of geometry simulation and nuclear data, the code MCNP is now widely used in the reactor analysis. Based on our calculation of the fuel temperature reactivity coefficient benchmark, it is quantificationally proved that MCNP with its own cross section library can' t be used to simulate the reactor accurately and to calculate the temperature reactivity coefficient. Furthermore, we use MCNP- 4C with a database that contains temperature dependent nuclear cross sections to calculate the benchmark. The results are well agreement with benchmark results. This means that, with the temperature dependent nuclear cross sections library, MCNP can calculate the temperature reactivity coefficient and reactor multiplication factor accurately. So the temperature dependent nuclear cross section library should be processed to meet the requirement of reactor calculation.

关 键 词:MCNP程序   核素截面数据库   增殖系数   反应性温度系数   基准题
文章编号:1007-4627(2006)02-0111-04
收稿时间:2005-11-20
修稿时间:2006-01-09

Requirement of Temperature Dependent Nuclear Cross Sections for MCNP Calculation
CHAI Xiao-ming,WANG Kan,YU Gang-lin. Requirement of Temperature Dependent Nuclear Cross Sections for MCNP Calculation[J]. Nuclear Physics Review, 2006, 23(2): 111-114. DOI: 10.11804/NuclPhysRev.23.02.111
Authors:CHAI Xiao-ming  WANG Kan  YU Gang-lin
Affiliation:Department of Engineering Physics, Tsinghua University, Beijing 100084, China
Abstract:Due to the advantage of geometry simulation and nuclear data,the code MCNP is now widely used in the reactor analysis.Based on our calculation of the fuel temperature reactivity coefficient benchmark,it is quantificationally proved that MCNP with its own cross section library can't be used to simulate the reactor accurately and to calculate the temperature reactivity coefficient.Furthermore,we use MCNP-4C with a database that contains temperature dependent nuclear cross sections to calculate the benchmark.The results are well agreement with benchmark results.This means that,with the temperature dependent nuclear cross sections library,MCNP can calculate the temperature reactivity coefficient and reactor multiplication factor accurately.So the temperature dependent nuclear cross section library should be processed to meet the requirement of reactor calculation.
Keywords:MCNP code  nuclear cross section library  multiplication factor  temperature reactivity coefficient  benchmark  
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