Analysis of 235U, 239Pu and 241Pu content in a spent fuel assembly using Lead Slowing Down Spectrometer and time intervals matrix

Nowadays 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 obtained through using of various techniques, one of which is the non-destructive assay (NDA) 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 [3], was based on a matrix of time intervals where large differences in the number of fissions of 235U and 239Pu 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 proposed method for characterization of spent fuel assemblies has only been studied using Monte Carlo simulations, it was possible to demonstrate the determination of 239Pu using a DT pulsed neutron source, a Lead Slowing Down Spectrometer, and fast timing scintillator that is sensitive to both photons and neutrons. Additional information about the system can be obtained from n-γ pulse shape discrimination