Irradiation behavior of uranium oxide-aluminum dispersion fuel

Conference · Sun Dec 01 04:00:00 EST 1996

OSTI ID:414400

Hofman, G L; Rest, J; Snelgrove, J L

An oxide version of the DART code has been generated in order to assess the irradiation behavior of UO{sub 2}-Al dispersion fuel. The aluminum-fuel interaction models were developed based on U{sub 3}O{sub 8}-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 U{sub 3}O{sub 8} are valid for UO{sub 2}, the LEU UO{sub 2}-Al with a 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 x 10{sup 27} fissions m{sup {minus}3} ({approximately} 63% {sup 235}U burnup).

• Research Organization: Argonne National Lab., IL (United States)
• Sponsoring Organization: USDOE, Washington, DC (United States)
• DOE Contract Number: W-31109-ENG-38
• OSTI ID: 414400
• Report Number(s): ANL/TD/CP--91547; CONF-9610205--3; ON: DE97001402
• Country of Publication: United States
• Language: English

Related Subjects

05 NUCLEAR FUELS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
ALUMINIUM ALLOYS
DISPERSION NUCLEAR FUELS
FUEL ELEMENTS
HIGHLY ENRICHED URANIUM
MATERIAL SUBSTITUTION
MODERATELY ENRICHED URANIUM
PHYSICAL RADIATION EFFECTS
POST-IRRADIATION EXAMINATION
RESEARCH AND TEST REACTORS
SLIGHTLY ENRICHED URANIUM
SWELLING
URANIUM OXIDES