Title: Analysis of Differential Die Away Instrument Simulated Performance Using Boiling Water Reactor Spent Fuel Assemblies

In this study, Monte Carlo simulations were performed for the differential die-away (DDA) technique to analyse the time-dependent behaviour of the neutron population in fresh and spent nuclear fuel assemblies as part of the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Simulations were performed to investigate both a possibly portable as well as a permanent DDA instrument. Taking advantage of a custom made modification to the MCNPX code, the variation in the neutron population, simultaneously in time and space, was examined. The motivation for this research was to improve the design of the DDA instrument, as it is bemore » ing considered for possible deployment at the Central Storage of Spent Nuclear Fuel and Encapsulation Plant in Sweden (Clab), as well as to assist in the interpretation of the both simulated and measured signals.« less

New Monte Carlo simulations of the differential die-away (DDA) instrument response to the assay of spent and fresh fuel helped to redefine the signal-to-Background ratio and the effects of source neutron tailoring on the system performance. Previously, burst neutrons from the neutron generator together with all neutrons from a fission chain started by a fast fission of ²³⁸U were considered to contribute to active background counts. However, through additional simulations, the magnitude of the ²³⁸U first fission contribution was found to not affect the DDA performance in reconstructing ²³⁹Pu _eff. As a result, the newly adopted DDA active background definitionmore » considers now any neutrons within a branch of the fission chain that does not include at least one fission event induced by a thermal neutron, before being detected, to be the active background. The active background, consisting thus of neutrons from a fission chain or its individual branches composed entirely of sequence of fast fissions on any fissile or fissionable nuclei, is not expected to change significantly with different fuel assemblies. Additionally, while source tailoring materials surrounding the neutron generator were found to influence and possibly improve the instrument performance, the effect was not substantial.« less

The differential die-away (DDA) technique has been simulated by using the MCNPX code to quantify its capability to measure the fissile content in spent fuel assemblies, For 64 different spent fuel cases of various initial enrichment, burnup and cooling time, the count rate and signal to background ratios of the DDA system were obtained, where neutron backgrounds are mainly coming from the {sup 244}Cm of the spent fuel. To quantify the total fissile mass of spent fuel, a concept of the effective {sup 239}Pu mass was introduced by weighting the relative contribution to the signal of {sup 235}U and {supmore » 241}Pu compared to {sup 239}Pu and the calibration curves of DDA count rate vs. {sup 239}Pu{sub eff} were obtained by using the MCNPX code. With a deuterium-tritium (DT) neutron generator of 10{sup 9} n/s strength, signal to background ratios of sufficient magnitude are acquired for a DDA system with the spent fuel assembly in water.« less

Four helium-3 ( ³He) detector/preamplifier packages (¾”/KM200, DDSI/PDT-A111, DDA/PDT-A111, and DDA/PDT10A) were experimentally tested to determine the deadtime effects at different DT neutron generator output settings. At very high count rates, the ¾”/KM200 package performed best. At high count rates, the ¾”/KM200 and the DDSI/PDT-A111 packages performed very well, with the DDSI/PDT-A111 operating with slightly higher efficiency. All of the packages performed similarly at mid to low count rates. Proposed improvements include using a fast recovery LANL-made dual channel preamplifier, testing smaller diameter ³He tubes, and further investigating quench gases.