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

作　　者：Afrah El-Khawlani Moustafa Aziz Ali Ellithi 

机构地区：[1]Faculty of Science,Sana’a University,Sana’a 72738,Yemen [2]Nuclear and Radiological Regulatory Authority,Nasr City 11762,Cairo [3]College of Science,Cairo University,Cairo 12613,Egypt

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

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).

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]