Monte Carlo Simulation of Neutronic Characteristics of the Yalina-Thermal Subcritical Assembly Using the ENDF/B-VII and JENDL-3.2 Evaluated Neutron Data Libraries

Abstract

The results of the numerical simulation of an experiment on the Yalina-Thermal subcritical assembly with 10% enriched fuel conducted for the determination of the fast neutron spectrum in the core volume of the subcritical assembly are presented. The work was carried out using the MCNP4B radiation transport calculation code, the ENDF/B-VII and JENDL-3.2 evaluated nuclear data libraries, and the SAND-II neutron spectrum unfolding code. The neutron spectra calculated using MCNP4B and the neutron spectra unfolded using SAND-II agree well, which proves that the SAND-II code is correct and can be used to unfold spectra from the experimental data of reaction rates with use of the Monte Carlo spectrum as a reference.