Analysis of 235U, 239Pu, and 241Pu Content in a Spent Fuel Assembly Using Lead Slowing Down Spectrometer and Time Intervals Matrix

Knowledge of the physical parameters of irradiated nuclear fuel isgoing to be a key issue for the continued and future use of nuclearenergy. One of the major characteristics of spent fuel, which playsan important role in international nuclear materials safeguards isthe quantity of plutonium (Pu) in wastes. It can be determinedthrough the use of various techniques, one of which is the nondestructiveassay (NDA) method of slowing-down time spectrometryin lead where the energy spectrum of neutrons can berepresented as being monoenergetic with minor deviation fromthe peak value in each time moment after a fast neutron pulse.This fact was successfully used in developing several methods ofPu mass determination and confirmed the potential of the LeadSlowing Down Spectrometer (LSDS) to get detailed informationabout spent fuel.1,2 A method, presented earlier at the InternationalConference on Current Problems in Nuclear Physics andAtomic Energy,3 was based on a matrix of time intervals wherelarge differences in the number of fissions of 235U and 239Pu areobserved. This technique allows increasing precision in the Puevaluation by decreasing the self-shielding effect significantly. Asopposed to homogeneous-volume approximations used in ourprevious research, we describe the detailed Monte Carlo modelsof real fuel assemblies, as well as the effects of the influence of thescintillation detector to the system in question. Although the proposedmethod for characterizing spent fuel assemblies has onlybeen studied using Monte Carlo simulations, it was possible todemonstrate the determination of 239Pu using a DT pulsed neutronsource, a lead slowing down spectrometer, and fast timingscintillator that is sensitive to both photons and neutrons. Additionalinformation about the system can be obtained from n-γpulse shape discrimination.