Time intervals matrix analysis of 235U and 239Pu content in a spent fuel assembly using Lead Slowing down Spectrometer

A key motivation for developing the technology capable of quantifying plutonium (Pu) in spent fuel assemblies with nondestructive assay (NDA) techniques is knowledge of the physical parameters of irradiated nuclear fuel which is important both for nuclear safeguards and nuclear fuel management. One of the most attractive NDA approaches applied for the determination of the total amount of plutonium using a pulsed neutron source is the method of slowing-down time spectrometry 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 [1-2].A method which we presented earlier is based on a matrix of time intervals where large differences in the number of fissions of 235U and 239Pu are observed [3]. 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 in this work 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 proposed method for spent fuel assemblies characterization has only been studied using Monte Carlo simulations, it was possible to demonstrate the 239Pu determination using a DT pulsed neutron source, Lead Slowing Down Spectrometer and fast timing scintillatior which is sensitive to both photons and neutrons, and n-γ pulse shape discrimination allows to get additional information about the system