Irradiation Behavior of Uranium Oxide-aluminum Dispersion Fuel
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Published: 1996
Total Pages: 29
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DOWNLOAD EBOOKAn 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 U3O-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 U3O are valid for UO2, the LEU UO2-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 1027 fissions m−3 ((approximately) 63% 235U burnup).