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Dissolution Kinetics of UO2. I. Flow-Through tests on UO2.00 Pellets and Polycrystalline Schoepite Samples in Oxygenated, Carbonate/Bicarbonate Buffer Solutions at 25°C

Published online by Cambridge University Press:  25 February 2011

Son N. Nguyen ,
Homer C. Weed ,
Herman R. Leider  and
Ray B. Stout
Son N. Nguyen
Affiliation:
Lawrence Livermore National Laboratory, P. O. BOX 808, Livermore, CA 94550
Homer C. Weed
Affiliation:
Lawrence Livermore National Laboratory, P. O. BOX 808, Livermore, CA 94550
Herman R. Leider
Affiliation:
Lawrence Livermore National Laboratory, P. O. BOX 808, Livermore, CA 94550
Ray B. Stout
Affiliation:
Lawrence Livermore National Laboratory, P. O. BOX 808, Livermore, CA 94550
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Abstract

The modelling of radionuclide release from waste forms is an important part of the performance assessment of a potential, high-level radioactive waste repository. Since spent fuel consists of UO2 containing actinide elements and other fission products, it is necessary to determine the principal parameters affecting UO2 dissolution and quantify their effects on the dissolution rate before any prediction of long term release rates of radionuclides from the spent fuel can be made.

Type
Research Article
Information
MRS Online Proceedings Library (OPL) , Volume 257: Symposium – Scientific Basis for Nuclear Waste Management XV , 1991 , 339
Copyright
Copyright © Materials Research Society 1992

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References

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