CANDU Fuel Attribution Through the Analysis of Delayed Neutron Temporal Behaviour

Abstract

Delayed Neutron Counting (DNC) is an established technique in the Canadian nuclear industry as it is used for the detection of defective fuel in several CANDU reactors and the assay of uranium in geological samples. This paper describes the possible expansion of DNC to the discipline of nuclear forensics analysis. The temporal behaviour of experimentally measured delayed neutron spectra were used to determine the relative contributions of ²³³U and ²³⁵U to the overall fissile content present in mixtures with average absolute errors of ±4 %. The characterization of fissile content in current and proposed CANDU fuels (natural UO₂, thoria and mixed oxide (MOX) based) by DNC analysis is evaluated through Monte Carlo simulations.