MCNP simulation of neutron energy spectra for a TN-32 dry shielded container

Abstract

This paper presents a numerical analysis of neutron energy spectra for a TN-32 spent fuel dry storage cask using Monte Carlo simulation. The analysis results were compared with experimental measurements to determine the suitability of using such codes for neutron flux calculations in soft-spectrum neutron environments. Complete spent fuel compositions were generated using Scale 4.4a. Variations in source definition and geometry determined that geometric and source simplifications in the computational model have negligible effect on final neutron energy distribution. Variations between experimental and computed spectra at energies above 1 MeV and below 100 keV demonstrated the shortfalls of various detection instruments used to collect the experimental neutron energy spectra data principally because these instruments were calibrated based on high neutron energy spectra. The MCNP calculations were generally in agree with the experimental data, but predicted that the detectors would over-respond to the neutron spectra around a spent fuel dry shielded container. Computed neutron energy spectra were always conservative when compared to experimental spectra.