Experimental Studies on the Self-Shielding Effect in Fissile Fuel Breeding Measurement in Thorium Oxide Pellets Irradiated with 14 MeV Neutrons  

作　　者：Mitul ABHANGI Nupur JAIN Rajnikant MAKWANA Sudhirsinh VALA Shrichand JAKHAR T. K. BASU C. V. S. RAO 

机构地区：[1]Fusion Neutronics Laboratory, Institute for Plasma Research,Bhat, Gandhinagar-382428,Gujarat, India [2]School of Nuclear Energy, Pandit Deendayal Petroleum University,Gandhinagar,Gujarat 382007, India

出　　处：《Plasma Science and Technology》2013年第2期166-170,共5页等离子体科学和技术（英文版）

摘　　要：The 14 MeV neutrons produced in the D-T fusion reactions have the potential of breeding Uranium-233 fissile fuel from fertile material Thorium-232. In order to estimate the amount of U-233 produced, experiments are carried out by irradiating thorium dioxide pellets with neutrons produced from a 14 MeV neutron generator. The objective of the present work is to measure the reaction rates of 232Th ＋ in --＊ 233Th --＊ 2a^Pa --＊ 2a3U in different pellet thicknesses to study the self-shielding effects and adopt a procedure for correction. An appropriate assembly consisting of high-density polyethylene is designed and fabricated to slow down the high-energy neutrons, in which Thorium pellets are irradiated. The amount of fissile fuel （~3~U） produced is estimated by measuring the 312 keV gammas emitted by Protactinium-233 （half-life of 27 days）. A calibrated High Purity Germanium （HPGe） detector is used to measure the gamma ray spectrum. The amount of 233U produced by Th232 （n, ~） is calculated using MCNP code. The self-shielding effect is evaluated by calculating the reaction rates for different foil thickness. MCNP calculation results are compared with the experimental values and appropriate correction factors are estimated for self-shielding of neutrons and absorption of gamma rays.The 14 MeV neutrons produced in the D-T fusion reactions have the potential of breeding Uranium-233 fissile fuel from fertile material Thorium-232. In order to estimate the amount of U-233 produced, experiments are carried out by irradiating thorium dioxide pellets with neutrons produced from a 14 MeV neutron generator. The objective of the present work is to measure the reaction rates of 232Th ＋ in --＊ 233Th --＊ 2a^Pa --＊ 2a3U in different pellet thicknesses to study the self-shielding effects and adopt a procedure for correction. An appropriate assembly consisting of high-density polyethylene is designed and fabricated to slow down the high-energy neutrons, in which Thorium pellets are irradiated. The amount of fissile fuel （~3~U） produced is estimated by measuring the 312 keV gammas emitted by Protactinium-233 （half-life of 27 days）. A calibrated High Purity Germanium （HPGe） detector is used to measure the gamma ray spectrum. The amount of 233U produced by Th232 （n, ~） is calculated using MCNP code. The self-shielding effect is evaluated by calculating the reaction rates for different foil thickness. MCNP calculation results are compared with the experimental values and appropriate correction factors are estimated for self-shielding of neutrons and absorption of gamma rays.

关 键 词：233U breeding fissile fuel MCNP 14 MeV neutron source 

分 类 号：TL613[核科学技术—核技术及应用]