Neutron intensity of spent fast reactor fuel

Takamatsu, Misao ; Aoyama, Takafumi ; Masui, Tomohiko*

Neutron intensity of spent fuel is important not only for the shielding design and dose evaluation of the reprocessing plant and the transportation of the mixed oxide (MOX) fuel, but also for the core management, because it contains more minor actinides than that of LWR fuel. In order to obtain the experimental data and to improve the accuracy of burnup calculation, the neutron counting rate from a spent fuel subassembly of the JOYO MK-II core with a burnup of 62,5ooMWd/t and cooling time of 5.2 years was measured in the spent fuel storage pond at JOYO. The measured neutron counting rate was then converted to the neutron intensity using the detector response which was obtained by the Monte Carlo calculation code "MCNP-4A". And the neutron intensity was compared with the calculated value based on the 3-D diffusion theory with 7 energy groups using the JOYO core management code system "MAGI". The major results obtained in this study are summarized as follows; (1) The measured neutron intensity per fuel subassembly was about 2.710 n/s and is about 3 times as much as that of fresh (unirradiated) fuel. (2) The average C/E value of neutron intensity was about 1.07. (3)It was found that the axial neutron intensity didn't simply follow the burnup distribution, and the neutron intensity was locally increased at the bottom end of the fuel region due to an accumulation of Cm.