SCK-CEN spent fuel libraries
Description
This dataset is composed of a set of reference spent fuel libraries. The dataset contains the material composition and radiation emission (i.e. neutrons and gamma-rays) of spent fuel. The data was obtained with computer simulations using the ORIGEN-ARP code, which is part of the SCALE 6.1 package. Both PWR and BWR fuel geometries are included in the dataset, and for each geometry both UO2 and MOX fuel materials are considered. The dataset contains the information for spent fuel with a broad range of initial enrichment (UO2 fuel) or initial fissile content (MOX fuel), discharge burnup, and cooling times. For each simulation the neutron spectra, divided into contributions from (α,n) reactions, spontaneous fission, and total neutron emission, as well as total gamma-ray spectra are included. The neutron emission from selected isotopes is also reported, divided in contributions from (α,n) reactions and spontaneous fissions.