Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 29
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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 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).
Author: J. A. L. Robertson
Publisher: New York : Gordon and Breach
ISBN:
Category : Technology & Engineering
Languages : en
Pages : 328
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Author: D. I. Sinizer
Publisher:
ISBN:
Category : Nuclear fuels
Languages : en
Pages : 52
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Author: W. C. Thurber
Publisher:
ISBN:
Category : Nuclear reactions
Languages : en
Pages : 58
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Author: U.S. Nuclear Regulatory Commission
Publisher:
ISBN:
Category : Nuclear energy
Languages : en
Pages : 950
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Author: V. M. Golyanov
Publisher:
ISBN:
Category :
Languages : en
Pages : 22
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The effects of irradiation on UO2 were investigated by electron microscopy of thin films. Annealing of defects, the nature and formation of defects, fission gas formation, and the relation of radiation dose to defect concentration were considered.
Author: R. D. Leggett
Publisher:
ISBN:
Category : Uranium
Languages : en
Pages : 66
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Author: J. A. Horak
Publisher:
ISBN:
Category : Irradiation
Languages : en
Pages : 46
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Twelve 0.22-in.-diameter fuel specimens containing a longitudinal central vent and clad with 0.010 in. of Type 304 stainless steel were irradiated to evaluate the effect of restraint and a central vent on fuel element stability. The cladding of 10 of the specimens contained porous end plugs to vent any released fission gas and thus to minimize the buildup of gas pressure within the stainless steel cladding. The specimens consisted of a 20% enriched uranium--2 wt% zirconium alloy core surrounded by a natural uranium--2 wt% zirconium alloy sleeve. Eight of the specimens were irradiated to burnups of the enriched core of 6.9 to 12.8% of all atoms (1.2 to 2.2 at.% of the duplex assembly) at maximum fuel temperatures ranging from 280 to 760 deg C. Most of the clad specimens exhibited negligible volume increases as a result of irradiation. Two specimens containing central vents but unclad were irradiated together with the clad specimens in an attempt to differentiate between the effects due to a central vent and the effects due to cladding. The central vent in itself did not appear to reduce the swelling characteristics of the alloy. Mechanical restraint appeared to have extended the useful operating temperatures of the metallic fuel alloy by at least 200 deg C and also greatly extended the burnup levels to which the fuel could be irradiated.
Author: Roger Conant Waugh
Publisher:
ISBN:
Category : Uranium
Languages : en
Pages : 146
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Author: Gordon G. Bentle
Publisher:
ISBN:
Category : Nuclear fuel elements
Languages : en
Pages : 76
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