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<p>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&#xa0;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.</p>

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

  • 期刊论文

  • 发表日期：

    2025年1月31日

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  • Yi-Nong Li,
  • Zheng Wei,
  • Gen-Tao Gao,
  • Lu Wu,
  • Kang Wu,
  • Jun Ma,
  • Xing-Yu Liu,
  • Ze-En Yao,
  • Yu Zhang,
  • Jun-Run Wang,
  • Xiao-Dong Su,
  • Zhi-Yong Deng,
  • Guo-Rong Wan

摘要

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.