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

摘要

Knowledge of the physical parameters of irradiated nuclear fuel is going to be a key issue for the continued and future use of nuclear energy. One of the major characteristics of spent fuel, which plays an important role in international nuclear materials safeguards is the quantity of plutonium (Pu) in wastes. It can be determined through the use of various techniques, one of which is the non destructive assay (NDA) method of slowing-down time spec trometry in lead where the energy spectrum of neutrons can be represented as being monoenergetic with minor deviation from the peak value in each time moment after a fast neutron pulse. This fact was successfully used in developing several methods of Pu mass determination and confirmed the potential of the Lead Slowing Down Spectrometer (LSDS) to get detailed information about spent fuel. A method, presented earlier at the Interna tional Conference on Current Problems in Nuclear Physics and Atomic Energy, was based on a matrix of time intervals where large differences in the number of fissions of ~(235)U and ~(239)Pu are observed. This technique allows increasing precision in the Pu evaluation by decreasing the self-shielding effect significantly. As opposed to homogeneous-volume approximations used in our previous research, we describe the detailed Monte Carlo models of real fuel assemblies, as well as the effects of the influence of the scintillation detector to the system in question. Although the pro posed method for characterizing spent fuel assemblies has only been studied using Monte Carlo simulations, it was possible to demonstrate the determination of ~(239)Pu using a DT pulsed neu tron source, a lead slowing down spectrometer, and fast timing scintillator that is sensitive to both photons and neutrons. Ad ditional information about the system can be obtained from n-y pulse shape discrimination.