Application of Dose Evaluation of the MCNP Code for the Spent Fuel Transport Cask

Abstract

The spent fuel transport cask has structure based mostly on the multi-layer for the radiation shielding. However, the structure for neutron shielding is very weak around the trunnion. It is important to evaluate realistic dose-equivalent rate in shielding design of the spent fuel transport cask, therefore the three dimensional conti- nuous-energy Monte Carlo radiation transport code that exactly treating the complicated geometry was applied. In the fixed source problem such as a neutron deep penetration calculation with the Monte Carlo method, the application of the variance reduction method is the most important for a high figure of merit and the most reliable calculation. The concerned items are setting method for the variance reduction. The validation of dose-equivalent rate evaluation for the spent fuel transport cask by Monte Carlo code were performed by two kinds of neutron shielding benchmark ex- periments, deep penetration experiment for the side surface of the cask and streaming experiment around the trunnion. Dose-equivalent rate distributions at each benchmark were measured and compared with the calculated results. The comparison showed a good consistency between calculation and experiment results.