Calculations of neutron transport through a simulated waste matrix

Abstract

The assay of nuclear waste characterized as Remote-Handled Transuranic (RH-TRU) waste presents a formidable problem due to the waste's high neutron and γ-ray backgrounds. The differential die-away technique (DDT) that is currently employed as a method for the assay requires the incidence of a large fluence of thermal neutrons within the volume element that contains fissile material. To study the neutron moderation and transport through nuclear waste matrices of different densities, the simulation of neutron transport was initiated using the Monte Carlo N-Particle (MCNP) code. A series of calculations using the MCNP 4A computer code have taken place. The calculations were performed to examine neutron transport across a wide range of waste matrix densities and interrogating neutron source energies. A composite matrix based on elemental analysis of RH-TRU legacy waste at ORNL was simulated and the density of this composite was varied between 200 and 2000 kg/m3. An isotropic interrogating neutron source was varied from 0.01 to 15 MeV to examine the transport of neutrons across these various densities. Fast (0.01–15 MeV), epithermal (10−7–0.01 MeV), and thermal (10−8–10−7 MeV) neutron transport was examined. The moderation of fast neutrons into thermal neutrons and their subsequent transport to the center of the drum was studied. A table with predictions of lower limits of detection of fissile material in a RH-TRU waste drum under certain conditions has been prepared.