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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Contact:
Dr.
Gerard Hofman
Senior Metallurgist
Argonne National Laboratory
9700 South Cass Avenue
Argonne, IL 60439 USA
Phone: (630) 252-6683
Fax: (630) 252-5161
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