Analyses of Burnup at plutonium Spots in Uranium-Plutonium Mixed Oxide Fuels in Light Water Reactors by Neutron Transport and Burnup Calculations
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Plutonium concentrations and burnup at Pu spots were calculated in U-Pu mixed oxide (MOX) fuel pellets for light water reactors with the neutron transport and burnup calculation code VIMBURN. The calculation models were suggested for Pu spots and U matrices in a heterogeneous MOX fuel pellet. The calculated Pu concentrations and burnup at Pu spots were compared with the PIEs data in a MOX pellet (38.8MWd/kgHM). The calculated Pu concentrations agreed by 5-18% with the measured ones, and the calculated burnup did by less than 10% with the estimated one with the measured Nd concentrations. Commercial PWR types of MOX fuels were also analyzed with the calculation code and the models. Burnup at Pu spot increased as the distance was greater from the radial center of a MOX fuel pellet. Burnup at Pu spots in the peripheral region became 3-5 times higher than pellet average burnup of 40MWd/kgHM. The diameters (20-100μm) of Pu spots were not found a significant factor for burnup at Pu spots. In the outer half volume region (outer than r/r<SUB>0</SUB>=0.7) of a MOX fuel pellet, burnup at Pu spots exceeded 70MWd/kgHM (the threshold burnup of microstructure change in UO<SUB>2</SUB> fuel pellet) at pellet average burnup of 14-30MWd/kgHM.
- Journal of Nuclear Science and Technology
Journal of Nuclear Science and Technology 34(6), 551-558, 1997-06-25
Atomic Energy Society of Japan