lmprovement of linear heat rate calculation for fast reactor MOX fueI using Monte Carlo code "MCNP"

not registered ; Kitamura, Ryoichi; Aoyama, Takafumi 

A three dimensional, continuous energy Monte Carlo code "MCNP" was applied to accurately evaluate the linear heat rate of mixed oxide (MOX) fuel used in a fast reactor. The test fuel pins to be analyzed were irradiated at the core center position in the experimental fast reactor JOYO MK-II core as the power-to-melt test, which was performed in order to optimize the thermal analysis method employed in the MOX fuel pin design. ln the calculation, the heterogeneous structure of the irradiation test subassembly (B5D-2) which accommodated test fuel pins inside was modeled precisely. The neutron flux distribution within the subassembly was calculated using MCNP with the neutron source which was based on the diffusion theory by the JOYO MK-II core management code system "MAGI" The linear heat rates of fuel pins were then obtained by multiplying the fission rate and fission energy of individual fissile nuclide. The gamma heating was calculated by MAGl considering the delayed fission gamma and it was included in the neutron heating obtained by MCNP. The accuracy of MCNP was verified by the post irradiation examination. using the fission product yields (Nd) measured for the melted part of test fuel pins, the linear heat rates were evaluated and compared with the MCNP results. The average ratios of MCNP calculations to the experimental values were 0.955 0.020, indicating MCNP can precisely calculate the linear heat rates. By correcting the MCNP calculation results with the C/E values obtained above, the linear heat rates of test fuel pins were evaluated to be 620 - 685 W/cm at the core center height.