Extension of the SCK•CEN spent fuel inventory library

SCK•CEN is performing R&D work related to the development of Non Destructive Assay on spent fuel elements. Due to difficulty in accessing spent fuel, having an accurate and validated model of the measurement equipment and a characterization of the source term is important for this type of studies. To better understand and develop methods based on neutron counting and gamma spectroscopy, a significant effort was done to determine the source terms associated to spent fuel with depletion and evolution code calculations. This paper reports about the extension of the SCK•CEN spent fuel inventory library with the ORIGEN-ARP code. The cases considered a Low Enriched Uranium 17×17 PWR fuel with an initial enrichment of 2.0, 2.5 and 3.0 %. This set of enrichments complements the data published previously for enrichments between 3.5 % and 5 %. In addition, results obtained with Mixed Oxide (MOX) fuel are presented and discussed. The impact of the initial enrichment, for LEU fuel, and the Pu mass, for MOX fuel, on the neutron emission and production of the main neutron emitters, are also studied.