Application of UNIST Monte Carlo Code MCS for Nuclear Fuel Transport Cask Analysis

Abstract

In this study, the applicability of the MCS code for the safety analysis of fuel casks is demonstrated. The criticality safety analysis and shielding calculations are carried out using the continuous-energy Monte Carlo neutron- and photon-transport code MCS developed by the COmputational Reactor physics and Experiment laboratory (CORE) group of Ulsan National Institute of Science and Technology (UNIST). Several cask models are simulated, including the KN12, KN18, GBC32 and KORAD21 casks. The value of the neutron multiplication factor is obtained for both fresh and used fuel. The neutron and gamma fuel radiation sources provided by the deterministic code STREAM are simulated in MCS to calculate the neutron and gamma dose rates. Calculations for selected casks are repeated with the MCNP6 Monte Carlo code and the SCALE6 package for comparison purposes.