Investigation of evaluated nuclear data in the prediction of inherent neutron sources

Quantifying inherent neutron sources in matter, particularly (α,n) reactions and spontaneous fission, is important in nuclear engineering and other fields. The SOURCES code is a common tool for calculating the yield and spectrum of such neutrons. This paper critically examines all modeling assumptions and nuclear data in SOURCES and proposes alternative approaches where applicable. For (α,n) reactions, we show that the alpha emission lines for 235U should be updated. Furthermore, we compare four different stopping power data sets for alpha particles slowing down and propose measurements to constrain mixed oxide nuclear fuel data. We use the computer code PHITS to show that energy and angular straggling during the slowing down of alpha particles in the material of interest is unimportant. Then, we compare the cross section and emission spectrum of (α,n) reactions in SOURCES to recently evaluated data libraries. Importantly, the modeling of SOURCES for the emission spectrum seems too simple and may need to be updated. Finally, we compare data on spontaneous fission and show that while the neutron yield from SOURCES is reliable, some discrepancy is found with the neutron spectrum of evaluated data libraries. Complementing this work is an implementation of spontaneous fission in the Monte Carlo code OpenMC.