urnup-Profile Measurement inside and outside the Spent Fuel Assembly

Axial burnup distribution of a spent fuel assembly plays very important role in the application of burnup credit and the safeguards. When the burnup credit is included for a spent fuel with high burnup, it was argued that the axial burnup distribution should be considered. The burnup is to obtain the total fissile material contents in the spent fuel. In most of the previous works, the axial burnup distribution was determined from the fuel-rod measurement or the measurement outside the spent fuel assembly. In the present work, the axial burnup distribution was measured inside and outside the spent fuel assembly. The gamma-ray dose distribution along the assembly axis was measured with an ionization chamber, and the neutron distribution was measured with a fission chamber, and the burnup distribution was deduced from the measurment. MCNPX simulation was performed to determine the relation between the measured distribution and the burnup distribution. A detector system to measure the burnup distribution outside the spent fuel assembly was also installed. A CZT detector, which was for the gamma-ray spectroscopy, an ionization chamber, which was to measure the gamma-ray dose distribution, and fission chambers, which were to determine the absolute value of burnup accurately, were installed in a detector system. The measured data from the detectors were collected and stored automatically as the detector box was moved along the spent fuel assembly. The burnup distribution, which was determined outside the spent fuel assembly, was presented, and compared with the burnup distribution determined inside the spent fuel assembly. Our work could be helpful to determine the burnup distribution more accurately.