Method for calculating delayed gamma-ray response in the ACRR Central Cavity using MCNP

¹ Sandia National Laboratories, Albuquerque, NM, USA
² Sandia National Laboratories (Retired), Albuquerque, NM, USA

^* Corresponding author: mmoren@sandia.gov

Published online: 2 March 2023

Abstract

This document presents the process for a new method developed for the characterization of the delayed gamma-ray radiation fields in pulse reactors like the Annular Core Research Reactor (ACRR). The environments used to test this method in the ACRR Central Cavity were various standard irradiation configurations that used various amounts of moderators and shielding. All environment configurations used the same fission product gamma-ray source energy spectrum. This method required the use of fission sites recorded in the Monte Carlo N-Particle (MCNP) KCODE source tape. A FORTRAN script was written to translate and extract the coordinates for the fission sites. The fission sites were then input into an MCNP source mode input file. Using a MATLAB script, a parametric analysis was performed and determined that 10K fission sites are an appropriate number of coordinates to converge to the correct answer. The method gave excellent results as compared to previous methods. This method can be applied to other pulse research reactors as well.