Mohammadi A, Hassanzadeh M, Gharib M
Iran Radioactive Waste Management Company, Tehran, Iran.
Nuclear Science and Technology Research Institute, Tehran, Iran.
Appl Radiat Isot. 2016 Feb;108:129-132. doi: 10.1016/j.apradiso.2015.12.045. Epub 2015 Dec 18.
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.
在本研究中，对通用材料试验反应堆（MTR）研究堆的临时储存及相关运输容器进行了屏蔽计算和临界安全分析。在这些过程中，考虑了三个主要方面：源项、屏蔽和临界计算。蒙特卡罗输运代码MCNP5用于屏蔽计算和临界安全分析，ORIGEN2.1代码用于源项计算。根据所得结果，一个筒体、顶部和底部厚度分别为18厘米、13厘米和13厘米的圆柱形容器被认定为两用容器。此外，结果表明总剂量率低于符合规定标准的正常运输标准。
