Determination of Fissile Nuclide ²³⁵U Content in Re-Irradiated Spent Fuel Assemble with Nondestructive Assay

Adequate knowledge of burnup levels of fuel elements within a research reactor is of great importance for its safe operation. The traditional nondestructive assay of burnup is to measure the radiation emitted either as neutrons or gamma rays. But the results are not satisfactory in accuracy because of variability of core loading and operation history. This paper presents a method for the experimental determination of fissile nuclide ²³⁵U content in Spent Fuel Assembles (SFAs). The method is based on re-irradiation of SFAs and measurement of the delayed gamma-rays emitted by the generated fission products. The most important advantage of this method is its independence of SFA irradiation history. This paper emphasizes how to discriminate the resource of characteristic gamma ray and introduces the experimental device. A SFA with about 15% burnup unloading from CMRR is measured by the above method and the uncertainty is less than 5%.