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Evaluation of Spent Fuel as a Waste Form In a Salt Repository

Published online by Cambridge University Press:  25 February 2011

Walter J. GRAY ,
Gary L. McVay ,
John. O. Barner ,
John W. Shade  and
Roger W. Cote
Walter J. GRAY
Affiliation:
Pacific Northwest Laboratory, P. O. Box 999, Richland, WA 99352;
Gary L. McVay
Affiliation:
Pacific Northwest Laboratory, P. O. Box 999, Richland, WA 99352;
John. O. Barner
Affiliation:
Pacific Northwest Laboratory, P. O. Box 999, Richland, WA 99352;
John W. Shade
Affiliation:
Pacific Northwest Laboratory, P. O. Box 999, Richland, WA 99352;
Roger W. Cote
Affiliation:
Battelle Memorial Institute, Office of Nuclear Waste Isolation, 505 King Avenue, Colunmus, OH 43201
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Abstract

Leach tests have been performed on spent fuel in synthetic Permian Basin salt brine at 25 and 75°C. Complementary tests on unirradiated UO2 pellets have been conducted in both salt brine and deionized water in the range 25 to 150°C. Iron and/or oxidized zircaloy coupons were included in some of the tests. Uranium release from spent fuel was more than 100 times greater than from U02. In brine, iron had no significant effect on the total uranium release but substantially reduced the amount in solution by causing the uranium to plate out on the iron coupon and container walls and to precipitate as filterable particles.

Type
Research Article
Information
MRS Online Proceedings Library (OPL) , Volume 26: Symposium D – Scientific Basis for Nuclear Waste Management VII , 1983 , 437
Copyright
Copyright © Materials Research Society 1984

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References

REFERENCES

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