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,...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 exhibits 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 10 MW material test reactor designed by the International Atomic Energy Agency.Consequently,the feasibility of the method and rationality of the detector design were verified.展开更多
基金supported by the National Natural Science Foundation of China(No.12075105)the Major Science and Technology Projects of Gansu Province(No.22ZD6GB020)+1 种基金the NSFC-Nuclear Technology Innovation Joint Fund(No.U2167203)the Fundamental Research Funds for the Central Universities(lzujbky-2023-stlt01,lzujbky-2024-jdzx10)。
文摘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 exhibits 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 10 MW material test reactor designed by the International Atomic Energy Agency.Consequently,the feasibility of the method and rationality of the detector design were verified.
