Implementation of a tree algorithm in MCNP code for nuclear well logging applications

被引：2
|
作者：
Li, Fusheng [1 ]
Han, Xiaogang [1 ]
机构：
[1] Baker Hughes Inc, Houston, TX 77073 USA
来源：
APPLIED RADIATION AND ISOTOPES | 2012年 / 70卷 / 07期
关键词：
Tree structure; MCNP; Monte Carlo simulation; Neutron interaction tally; Gamma-ray index tally; Nuclear parameters; RESPONSE FUNCTIONS;
D O I：
10.1016/j.apradiso.2011.11.005
中图分类号：
O61 [无机化学];
学科分类号：
070301 ; 081704 ;
摘要：
The goal of this paper is to develop some modeling capabilities that are missing in the current MCNP code. Those missing capabilities can greatly help for some certain nuclear tools designs, such as a nuclear lithology/mineralogy spectroscopy tool. The new capabilities to be developed in this paper include the following: zone tally, neutron interaction tally, gamma rays index tally and enhanced pulse-height tally. The patched MCNP code also can be used to compute neutron slowing-down length and thermal neutron diffusion length. (C) 2011 Elsevier Ltd. All rights reserved.
引用
收藏
页码：1063 / 1069
页数：7
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