A General Evaluation of the Irradiation Behavior of Dispersion Fuels
A General Evaluation of the Irradiation Behavior of Dispersion Fuels

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Published: 1991

Total Pages: 15

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This 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.

Predicted Irradiation Behavior of U Sub 3 O Sub 8 -Al Dispersion Fuels for Production Reactor Applications
Predicted Irradiation Behavior of U Sub 3 O Sub 8 -Al Dispersion Fuels for Production Reactor Applications

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Published: 1990

Total Pages: 7

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Candidate 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.

Initial Assessment of Radiation Behavior of Very-high-density Low-enriched-uranium Fuels
Initial Assessment of Radiation Behavior of Very-high-density Low-enriched-uranium Fuels

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Published: 1999

Total Pages: 17

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Results from the postirradiation examinations of microplates irradiated in the RERTR-1 and -2 experiments in the ATR have shown several binary and ternary U-Mo alloys to be promising candidates for use in aluminum-based dispersion fuels with uranium densities up to 8 to 9 g/cm3. Ternary alloys of uranium, niobium, and zirconium performed poorly, however, both in terms of fuel/matrix reaction and fission-gas-bubble behavior, and have been dropped from further study. Since irradiation temperatures achieved in the present experiments (approximately 70 C)are considerably lower than might be experienced in a high-performance reactor, a new experiment is being planned with beginning-of-cycle temperatures greater than 200 C in 8-g U/cm3 fuel.

Correlation Between Annealing and Irradiation Behavior of Dispersion Fuels
Correlation Between Annealing and Irradiation Behavior of Dispersion Fuels

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Published: 1987

Total Pages:

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Studying 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.

Detailed Analysis of Uranium Silicide Dispersion Fuel Swelling
Detailed Analysis of Uranium Silicide Dispersion Fuel Swelling

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Published: 1989

Total Pages: 21

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Swelling of U3Si and U3Si2 is analyzed. The growth of fission gas bubbles appears to be affected by fission rate, fuel loading, and microstructural change taking place in the fuel compounds during irradiation. Several mechanisms are explored to explain the observations. The present work is aimed at a better understanding of the basic swelling phenomenon in order to accurately model irradiation behavior of uranium silicide disperson fuel. 5 refs., 10 figs.

Comparison of Irradiation Behavior of Different Uranium Silicide Dispersion Fuel Element Designs
Comparison of Irradiation Behavior of Different Uranium Silicide Dispersion Fuel Element Designs

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Published: 1995

Total Pages: 12

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Calculations of fuel swelling of U3SiAl-Al and U3Si2 were performed for various dispersion fuel element designs. Breakaway swelling criteria in the form of critical fuel volume fractions were derived with data obtained from U3SiAl-Al plate irradiations. The results of the analysis show that rod-type elements remain well below the pillowing threshold. However, tubular fuel elements, which behave essentially like plates, will likely develop pillows or blisters at around 90% 235U burnup. The U3Si2-Al compounds demonstrate stable swelling behavior throughout the entire burnup range for all fuel element designs.