MCNP源子程序在停机剂量率计算中的应用  

Application of MCNP SOURCE subroutine to shutdown dose rate calculation

作　　者：吴明昌[1,2] 吴亮亮[1,2] 郝丽娟[2] 郑华庆[2] 宋婧[1,2] FDS团队 

机构地区：[1]中国科学技术大学,安徽合肥230027 [2]中国科学院核能安全技术研究所,安徽合肥230031

出　　处：《核科学与工程》2012年第4期366-370,共5页Nuclear Science and Engineering

基　　金：中科院信息化专项项目(INFO-115-C01-SDB2-01)

摘　　要：在核聚变装置的停堆剂量率的计算中,通常采用MCNP程序来实现光子的输运计算,但由于聚变装置几何和材料的高度复杂性使得栅元数量庞大,同时由于标准MCNP在进行光子输运计算时,SDEF通用源卡只支持1 000个以下的栅元描述,因此直接采用SDEF通用源卡的方法无法实现聚变堆的停堆剂量率精确计算与分析。本论文采用MCNP内置源子程序方法直接对衰变光子源进行抽样,解决了SDEF通用源卡受限的问题。以国际热核聚变实验堆ITER最新发布的停机剂量率基准例题以及ITER-T426基准实验例题对源子程序进行了校验,结果表明了该方法的可用性与正确性。The photon transport calculation in shutdown dose rate calculation of fusion facility is carried out by MCNP usually. Huge amount of cells are needed to describe the fusion facility due to its high complexity of material and geometry. The SDEF card of MCNP limits that the number of source cells should be below 1 000. Thus, the photon source involved in shutdown dose rate calculation code cannot be specified by the card because of this restriction. To settle down the problem, the source is sampled directly in SOURCE subroutine in this paper. Tests by ITER benchmark for shutdown dose rate and ITER-T426 experimental benchmark show its validity and reliability.

关 键 词：MCNP 源子程序 停机剂量率 网格计数 

分 类 号：TL63[核科学技术—核技术及应用]