Plate spent fuel burnup measurement equipment based on a compact D-D neutron generator

Abstract: Burnup measurement is crucial for the management and disposal of spent fuel. The conventional approach
indirectly estimates burnup by examining the fission product or actinide content. Compared to the first two
methods, the active neutron method exhibit a lower dependence on the irradiation history and initial enrichment
degree of the spent fuel. In addition, it can be used to directly determine the content of fissile nuclides in spent
fuel. This study proposed the design of a burnup measurement equipment specifically crafted for plate segments
by utilizing a compact D-D neutron generator. The equipment initiates the fission of fissile nuclides within the
spent fuel plate segment through thermal neutrons provided by the moderators. Subsequently, the burnup is
determined by analyzing the transmitted thermal neutrons and counting the fission fast neutrons. The Monte
Carlo program Geant4 was used to simulate the relationship between spent fuel plate segment assembly burnup
and the detector count of 10MW material test reactor designed by the International Atomic Energy Agency.
Consequently, the feasibility of the method and rationality of the detector design were verified.