Development of an Open Source Spontaneous Fission Source File Generation for Multiplicity Counting Monte Carlo Simulations

This work introduces an open-source tool for generating spontaneous fission source files to facilitate neutron multiplicity counting simulation using Monte Carlo radiation transport codes such as OpenMC and Serpent 2. The tool incorporates the options to use the CGMF and FREYA fission event generators, leveraging their fission models and nuclear data to more accurately model spontaneous fission events, specifically including energy and angle correlations of neutrons emitted in fission. These source files enable simulations essential for nondestructive nuclear material assay, with applications in safeguards and nuclear security. Specifically, the tool allows researchers to optimize detector designs and conduct neutron multiplicity studies, contributing to advancements in material accountability and IAEA-driven security efforts. Results include multiple simulations of different plutonium-240 masses and neutron multiplicity counting mass estimates based on a helium-3 detector response. This method achieved errors within 15% of the true mass. This work represents the first steps towards a fully open source multiplicity counting simulation suite following the recent trends of open data of the IAEA and open science overall.