Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors

被引：9
|
作者：
Mohammadi, A. [1 ]
Hassanzadeh, M. [2 ]
Gharib, M. [3 ]
机构：
[1] Iran Radioact Waste Management Co, Tehran, Iran
[2] Nucl Sci & Technol Res Inst, Tehran, Iran
[3] Islamic Azad Univ, Sci & Res Branch, Tehran, Iran
来源：
APPLIED RADIATION AND ISOTOPES | 2016年 / 108卷
关键词：
Shielding; Criticality safety; Spent fuel; Transportation cask; Research Reactors; MCNP5; code; ORIGEN2.1;
D O I：
10.1016/j.apradiso.2015.12.045
中图分类号：
O61 [无机化学];
学科分类号：
070301 ; 081704 ;
摘要：
In this study, shielding calculation and criticality safety analysis were carried out for general material testing reactor (MTR) research reactors interim storage and relevant transportation cask. During these processes, three major terms were considered: source term, shielding, and criticality calculations. The Monte Carlo transport code MCNP5 was used for shielding calculation and criticality safety analysis and ORIGEN2.1 code for source term calculation. According to the results obtained, a cylindrical cask with body, top, and bottom thicknesses of 18, 13, and 13 cm, respectively, was accepted as the dual-purpose cask. Furthermore, it is shown that the total dose rates are below the normal transport criteria that meet the standards specified. (C) 2015 Elsevier Ltd. All rights reserved.
引用
收藏
页码：129 / 132
页数：4
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