CANDU-SEU fuel bundle shielding analysis during spent fuel transport

The basic tasks accomplished by the shielding calculations in a nuclear safety analysis consist in radiation doses calculation, to prevent any risks for personnel protection and to reduce the impact on the environment during the spent fuel manipulation, transport and storage. In the last decade, a general trend to raise the discharge fuel burnup has been world wide registered for operating reactors a trend valid also in future reactor projects. For CANDU type reactors, the most attractive solution seems to be SEU and RU fuels utilization. The paper investigates the effects induced by fuel enrichment and bundle geometry modifications on the CANDU SEU spent fuel shielding analysis. In this study, the CANDU SEU 43 rods fuel bundle, developed in INR Pitesti, has been considered. The goal of the shielding analysis is to estimate the radiation doses to the shipping cask wall, by means of Monte Carlo MORSESGC code, included in the ORNL's SCALE 5 programs package. Before the transport, CANDU-SEU spent fuel was cooled up to 10 years in the NPP pools. All the geometrical and material data relating to the cask were considered according to the shipping cask type B model. (authors)