Constituent redistribution in U-Pu-Zr fuel during irradiation

    Authors

    Y. S. Kim; G. L. Hofman; S. L. Hayes;Y. H. Sohn

    Comments

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    Abbreviated Journal Title

    J. Nucl. Mater.

    Keywords

    FAST-REACTOR FUEL; TEMPERATURE-GRADIENT; THERMAL-GRADIENT; ALLOYS; BEHAVIOR; Materials Science, Multidisciplinary; Nuclear Science & Technology; Mining & Mineral Processing

    Abstract

    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 Title

    Journal of Nuclear Materials

    Volume

    327

    Issue/Number

    1

    Publication Date

    1-1-2004

    Document Type

    Article

    Language

    English

    First Page

    27

    Last Page

    36

    WOS Identifier

    WOS:000220576600004

    ISSN

    0022-3115

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