摘要: The 14 MeV neutrons produced in the D-T fusion reactions have the potential of breeding Uranium-233 ¯ssile 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 + ¹n→²³³Th→²³³Pa→²³³U in different pellet thicknesses to study the self-shielding e®ects 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 (²³³U) 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, γ) is calculated using MCNP code. The self-shielding e®ect 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.