Accuracy of Particle Production for Nuclear Transport with Geant4 and ENDF: Developing an IRT-M Neutron Flux Test Case

A study on particle production in nuclear reactions for various target isotopes was conducted. The Evaluated Nuclear Data File (ENDF) compresses all relevant cross-sections, playing a key role in nuclear transport simulations. Cross-sections indicate the probability of nuclear interactions between a projectile and a target and often includes particle production spectra resulting from these interactions. Utilizing a realistic reactor data, the likelihood of reaction products being emitted at different energies was assessed by comparing those findings to the evaluated ENDF files. To accurately evaluate the particle production, a Geant4 simulation test case for particle production accuracy was developed. It includes a mono-isotopic target and the realistic Baghdad Reactor neutron flux. The test case created can be generalized to various neutron flux distributions enabling a quick verification of particle production data within the ENDF. This neutron flux characterizes an IRT-M reactor and compares the simulated gamma-ray emissions outputs from Geant4/ENDF with the production data from The Baghdad Atlas. These gamma-ray productions, from inelastic reactions, are widely used in isotope identification applications and employed in modern applications such as nuclear forensics. By analyzing the outputs, inconsistencies were identified related to the Geant4 cascade generator and to the incomplete ENDF data entries. Our findings highlight the need for improvements in the accuracy of particle production models and evaluated nuclear data files.