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Reduced Enrichment for Research and Test Reactors
Nuclear Engineering Division at Argonne
U.S. Department of Energy

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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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Last modified on July 29, 2008 11:33 +0200