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The Melt-Dilute Treatment of Al-Base Highly Enriched DOE Spent Nuclear Fuels: Principles and Practices

Published online by Cambridge University Press:  10 February 2011

Thad M. Adams
Affiliation:
Westinghouse Savannah River Company, Savannah River Technology Center Aiken, SC 29808
Harold B. Peacock Jr.
Affiliation:
Westinghouse Savannah River Company, Savannah River Technology Center Aiken, SC 29808
Frederick C. Rhode
Affiliation:
Westinghouse Savannah River Company, Savannah River Technology Center Aiken, SC 29808
Natraj C. lyer
Affiliation:
Westinghouse Savannah River Company, Savannah River Technology Center Aiken, SC 29808
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Abstract

The melt-dilute treatment technology program is focused on the development and implementation of a treatment technology for diluting highly enriched (>20% 235U) aluminum spent nuclear fuel to low enriched levels (<20% 235U and qualifying the LEU Al-SNF form for geologic repository storage. In order to reduce the enrichment of these assemblies prior to ultimate geologic repository disposal, the melt-dilute technology proposes to melt these SNF assemblies and then dilute with additions of depleted uranium. The benefits accrued from this treatment process include the potential for significant volume reduction, reduced criticality potential, and the potential for enhanced SNF form characteristics. The emphasis within the development program to date has been on determining the process metallurgy and off-gas system design for the treatment of all types of Al SNF (UAIx, Al-U3O8, and AI-U3Si2). In determining the process metallurgy a wide range of alloys, representative of those expected in the Al-SNF form, have been fabricated and their product characteristics, namely microstructure, homogeneity, phase composition, and “ternary” constituent effects have been analyzed. As a result of the presence of species within the melt which will possess significant vapor pressures in the desired operating temperature range an off-gas system is necessary. Of the volatile species the one of greatest concern is 137Cs.

Type
Research Article
Copyright
Copyright © Materials Research Society 1999

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References

REFERENCES

1. Binary Alloy Phase Diagrams, Massalski, Thaddeus, ed., American Society for Metals, Metals Park, OH, 1986.Google Scholar
2. Brewer, A., Draft Functional Performance Requirements for SRS Spent Nuclear Fuel Treatment and Storage Services, Westinghouse Savannah River Company, 1998.Google Scholar
3. Petrov, Y., Alekseeva, Z., and Petrov, D., J. Nucl. Mat., 182 (1991), p. 60.Google Scholar
4. G, Copeland. and Snelgrove, J., Proceedings of the International Meeting on research and Test Reactor Core Conversions form HEU to LEU Fuels, November 1982, Argonne, IL, p.79.Google Scholar
5. Hofman, , Copeland, G. G., and Sanecki, J., Nuclear Technology, 72 (1986), p. 338.Google Scholar
6. Pasto, A., Copeland, G., and Martin, M., Bull. Cer. Soc., 61 (1982), p.491.Google Scholar
7. Lorenz, R. A., Fission Product Release Experiments, Presentation at the ANS Workshop on Safety of Uranium Aluminum Fast Reactors, March 1989.Google Scholar
8. Taleyarkhan, R.P., Analysis and Modeling of Fission Product Release from Various Uranium-Aluminum Plate-Type Reactor Fuels, Nuclear Safety, Vol.33, No.1, January-March 1992.Google Scholar
9. Shibata, T., Tamai, T., and Hayashi, M., Release of Fission Products from Irradiated Aluminide Fuel at High Temperatures, Nuclear Science and Engineering: 87, 405417 (1984).Google Scholar
10. Hodges, M.E.and Hyder, M.L., Offgas-Studies for the Melt-Dilute Program, ANS Third Topical Meeting, DOE Spent Nuclear Fuel and Fissile Materials Management, Charleston, SC, September 8-11, 1998.Google Scholar