Development of the SCK CEN reference datasets for spent fuel safeguards research and development

A set of reference spent fuel libraries has been organized in datasets containing 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 fuel assembly geometries for Pressurized Water Reactors (PWR 17 × 17) and Boiling Water Reactors (BWR 8 × 8) are included in the datasets, and for each geometry both uranium oxide (UO(2)) and mixed oxide (MOX) fuel materials are considered. The datasets contain the information for spent fuel with a broad range of initial enrichment (UO(2) fuel) or initial Pu content (MOX fuel), discharge burnup, and cooling time. 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. The datasets are publicly available and are in a format that facilitates the further data extraction and processing for applications in the safeguards research & development related to spent fuel measurement and verification.