Constituent redistribution in U-Pu-Zr fuel during irradiation
Abbreviated Journal Title
J. Nucl. Mater.
FAST-REACTOR FUEL; TEMPERATURE-GRADIENT; THERMAL-GRADIENT; ALLOYS; BEHAVIOR; Materials Science, Multidisciplinary; Nuclear Science & Technology; Mining & Mineral Processing
The thermo-migration fluxes of U, Pu and Zr in U-Pu-Zr metallic alloy fuel during irradiation in the Experimental Breeder Reactor II (EBR-II) were calculated using the constituent redistribution profiles measured in postirradiation examinations. Based on these fluxes, the diffusion coefficients, and the sums of heat of transport and enthalpy of solution for the gamma, gamma + zeta and delta + zeta phases in U-Pu-Zr were obtained. Using these data, the predicted concentration redistribution profiles are consistent with the measurements. The effect of minor actinide (Am and Np) addition was also examined. Am migration generally followed that of Zr with local precipitation, while Np behaved similarly to Pu. (C) 2004 Elsevier B.V. All rights reserved.
Journal of Nuclear Materials
"Constituent redistribution in U-Pu-Zr fuel during irradiation" (2004). Faculty Bibliography 2000s. 4493.