Criticality Safety Analysis for Cask Design with High Discharged Fuel Burnup Using MCNPX Code

Document Type : Original Article


1 Nuclear Safety Center, Egyptian Atomic Energy Authority, Cairo, Egypt

2 department of radiation safety research center for nuclear safety atomic energy authority


MCNPX computer code is used to model the general PWR cask which contain 32 typical PWR spent fuel assemblies. For Safe storage and transportation of the cask , factors that affect the criticality were studied using the concept of burn up credit. Several parameters such as initial fuel enrichment , fuel burnup history , cooling time , and axial burnup profile were analysed. The analysis was performed in two different steps , first burn the fuel assembly at different burnup and storage (cooling ) conditions , secondly, incorporate the details of the assemblies into the cask (canister ) model and perform a criticality calculations for the cask. Several cases of unnormal storage conditions are considered in the case of UO2- PWR only]. In this research high discharged fuel assemblies burn up include Standard UO2 - PWR and next generation MOX fuel. The present results are compared with similar GBC-32 benchmark and satisfactory agreements were found..