CALCULATIONS OF NEUTRON-TRANSPORT THROUGH A SIMULATED WASTE MATRIX

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 gamma-ray backgrounds, The differential die-away technique (DDT) that is currently employed as a method for the assay requires t he incidence of a large fluence of thermal neutrons within the volume element that contains fissile material. To study the neutron moderatio n and transport through nuclear waste matrices of different densities, the simulation of neutron transport was initiated using the Monte Car lo N-Particle (MCNP) code. A series of calculations using the MCNP 4A computer code have taken place. The calculations were performed to exa mine neutron transport across a wide range of waste matrix densities a nd interrogating neutron source energies. A composite matrix based on elemental analysis of RH-TRU legacy waste at ORNL was simulated and th e density of this composite was varied between 200 and 2000 kg/m(3). A n isotropic interrogating neutron source was varied from 0.01 to 15 Me V 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 ne utrons into thermal neutrons and their subsequent transport to the cen ter 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. (C) 1997 Elsevier Science B.V.