Abstracts and Available Papers Presented at the
1996 International RERTR Meeting

IRRADIATION BEHAVIOR OF
URANIUM OXIDE-ALUMINUM DISPERSION FUEL

Gerard L. Hofman, Jeffrey Rest and James L. Snelgrove

Argonne National Laboratory
9700 South Cass Avenue
Argonne, Illinois 60439-4841 USA

ABSTRACT

An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO₂-Al dispersion fuel. The aluminum-fuel interaction models were developed based on U₃O₈-Al irradiation data. Deformation of the fuel element occurs due to fuel particle swelling driven by both solid and gaseous fission products, as well as a consequence of the interaction between the fuel particles and the aluminum matrix. The calculations show, that with the assumption that the correlations derived from U₃O₈ are valid for UO₂, the LEU UO₂-Al with 42% fuel volume loading (4 gm/cc) irradiated at fuel temperatures greater than 413 K should undergo breakaway swelling at core burnups greater than about 1.12 x10²⁷ fissions m^-3 (~63%²³⁵U burnup).

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