Radiation Shielding Analysis for Pressurized Heavy Water Reactors (CANDU) Using MCNPX Code

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摘要 MCNPX(Monte Carlo N-Particle Transport Code)computer code is used to design a model to CANDU(Canada Deuterium Uranium)reactor core and its shielding system.It is assumed that reactor core is fueled with natural uranium.The core radiation sources are calculated which consider prompt neutrons,neutron induced gamma and prompt gamma radiations.The total neutron flux and dose rate are calculated along the shield and at outer shield points.The results indicated that the major dose rate component at outer shield points is due to neutron induced gamma dose rate(μSv/h).

作者 Afrah El-Khawlani Moustafa Aziz Ali Ellithi

机构地区 Faculty of Science Nuclear and Radiological Regulatory Authority College of Science

出处《材料科学与工程（中英文B版）》 2022年第2期50-57,共8页 Journal of Materials Science and Engineering B

关键词 CANDU reactor MCNPX code reactor shielding natural uranium radiation source

分类号 O57 [理学—粒子物理与原子核物理]

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材料科学与工程（中英文B版）

2022年第2期

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Radiation Shielding Analysis for Pressurized Heavy Water Reactors (CANDU) Using MCNPX Code

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