Irradiation Behavior of Dispersion Fuels
Author: D. W. White (Jr.)
Publisher:
Published: 1957
Total Pages: 62
ISBN-13:
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Author: D. W. White (Jr.)
Publisher:
Published: 1957
Total Pages: 62
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Published: 1990
Total Pages: 7
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DOWNLOAD EBOOKCandidate fuels for the new heavy-water production reactor include uranium/aluminum alloy and U3O--Al dispersion fuels. The U3O--Al dispersion fuel would make possible higher uranium loadings and would facilitate uranium recycle. Research efforts on U3O8--Al fuel include in-pile irradiation studies and development of analytical tools to characterize the behavior of dispersion fuels at high-burnup. In this paper the irradiation performance of U3O8--Al is assessed using the mechanistic Dispersion Analysis Research Tool (DART) code. Predictions of fuel swelling and alteration of thermal conductivity are presented and compared with experimental data. Calculational results indicate good agreement with available data where the effects of as-fabricated porosity and U3O8--Al oxygen exchange reactions are shown to exert a controlling influence on irradiation behavior. The DART code is judged to be a useful tool for assessing U3O8--Al performance over a wide range of irradiation conditions. 8 refs., 8 figs., 1 tab.
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Published: 1991
Total Pages: 15
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DOWNLOAD EBOOKThis document discusses the irradiation behavior of aluminum-based dispersion fuels and evaluates metallurgical processes that control the dispersion behavior. Phase transformations and microstructural changes resulting from fuel-matrix interactions and the effect of fissioning in fuel are discussed.
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Published: 1991
Total Pages: 15
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Published: 1987
Total Pages:
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DOWNLOAD EBOOKStudying the effects of annealing of scaled-down dispersion fuel plates is an important part of the data base for fuel performance. One of the most critical aspects of fuel performance is the stability of a fuel/matrix dispersion which is usually measured by volumetric changes of the fuel zone. A correlation has been proposed that fission-induced amorphization is responsible for the instability of the fuel and that such transformations can be predicted by the thermodynamic properties of the fuel. It is proposed that annealing studies may be used as a screening test for new fuels for which no thermodynamic properties have been measured and/or no irradiation data are available. Estimations of irradiation performance could be obtained faster and without the expense of irradiating the fuels under investigation. Miniature fuel plates were fabricated by standard procedures and annealed at 400°C for up to 1981 hrs in a resistance wound furnace. At periodic intervals the plates were removed and the fuel zone volumes were calculated based on immersion density measurement data. 7 refs., 1 tab.
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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).
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Published: 2001
Total Pages: 5
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Published: 2000
Total Pages: 5
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DOWNLOAD EBOOKAuthor: J. A. L. Robertson
Publisher: New York : Gordon and Breach
Published: 1969
Total Pages: 328
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Published: 1983
Total Pages:
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DOWNLOAD EBOOKUranium silicides, because of their relatively high uranium density, were selected as candidate dispersion fuels for the higher fuel densities required in the Reduced Enrichment Research and Test Reactor (RERTR) Program. Irradiation experience with this type of fuel, however, was limited to relatively modest fission densities in the bulk from, on the order of 7 x 102° cm−3, far short of the approximately 20 x 102° cm−3 goal established for the RERTR program. The purpose of the irradiation experiments on silicide fuels on the ORR, therefore, was to investigate the intrinsic irradiation behavior of uranium silicide as a dispersion fuel. Of particular interest was the interaction between the silicide particles and the aluminum matrix, the swelling behavior of the silicide particles, and the maximum volume fraction of silicide particles that could be contained in the aluminum matrix.