Safeguards Instrumentation Development for Measuring Pu Content in Spent Fuel

The direct measurement of total fissile content in spent fuel assemblies is needed to safeguard against the diversion of fissile materials to illegal weapons use. This is due to the fact that spent fuel contains a significant quantity of the world's weapons-useable fissile materials. The primary objective of this work is to develop and assess the sensitivity of using Self-Interrogation Neutron Resonance Densitometry (SINRD) for nuclear safeguards measurements. In order to achieve this objective, Monte Carlo Neutral Particle eXtended (MCNPX) transport code3 will be used to simulate SINRD for the following geometries: fissile metal plates, MOX fuel rods, and LWR fresh and spent fuel assemblies. The sensitivity of SINRD is based on using the same fissile materials in the sample and fission chamber because the effect of the resonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in fission chamber.