RADIATION SHIELDING ANALYSIS OF A SPENT FUEL TRANSPORT CASK WITH AN ACTUAL CONFIGURATION MODEL USING THE MONTE CARLO METHOD - COMPARISON WITH THE DISCRETE ORDINATES Sn METHOD -

In order to demonstrate the features of Monte Carlo method, in comparison with the two-dimensional discrete ordinates Sn method, detailed modeling of the canister containing the fuel basket with 14 spent fuel assemblies, supplement shields located around the lower nozzles of the fuels, and the cooling fins attached on the cask body of the NFT-14P cask are performed using the Monte Carlo code MCNP 4C. Furthermore, the water level in the canister is assimilated into the present MCNP 4C calculations. For more precise modeling of the canister, the generating points of gamma rays and neutrons are simulated accurately from the fuel assemblies installed in it. The supplement shields located around the lower nozzles of the fuels are designed to be effective especially for the activation 60Co gamma rays, and the cooling fins for gamma rays in particular. As predicated, compared with the DOT 3.5 calculations, the total dose-equivalent rates with the actual configurations are reduced to approximately 30 % at 1m from the upper side surface and 85 % at 1m from the lower side surface, respectively. Accordingly, the employment of detailed models for the Monte Carlo calculations is essential to accomplish more reasonable shielding design of a spent fuel transport cask and an interim storage cask. Quality of the actual configuration model of the canister containing the fuel basket with 12 spent fuel assemblies has already been demonstrated by the Monte Carlo analysis with MCNP 4B, in comparison with the measured dose-equivalent rates around the TN-12A cask.