Non-destructive measurement for plutonium 239 and uranium 235 in spent fuel assembly

Purpose: To measure in a non-destructive manner the quantities of plutonium 239 and uranium 235 contained in spent fuel disposed in a neutron decelerating medium in view of neutron emission rate, reactivity and decay. Constitution: For each of the spent and non-spent fuel assemblies, their reactivity, neutron counting rate and the ratio between the neutron counting rate after extraction of an external neutron source and the integrated value for the changes in the counting rate are measured respectively, and the degree of nuclear fission for the plutonium 239 and uranium 235 are determined based thereon. While on the other hand, spontaneous neutron generation rate for the spent fuel assembly is measured and the induced neutron generation rate is determined based on the above spontaneous rate and the above reactivity. The induced nuclear fission rate for the plutonium 239 and uranium 235 are determined based on the induced generation rate and the nuclear fission rate as determined above. While on the other hand, neutron flux in the spent fuel assembly is measured, based on which are determined visible macro-nuclear fission cross section for the plutonium 239 and uranium 235. Concentrations of the plutonium 239 and uranium 235 are determined using micro-nuclear fission cross section determined separately from theoretical calculation or the like. (Horiuchi, T.)