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Corrosion Behavior of Pre-oxidized High Burnup Spent Fuel in Salt Brine

Published online by Cambridge University Press:  01 February 2011

Andreas Loida ,
Bernhard Kienzler  and
Horst Geckeis
Andreas Loida
Affiliation:
Forschungszentrum Karlsruhe, Institut für Nukleare Entsorgung, P.O. Box 3640, D-76021 Karlsruhe, Germany
Bernhard Kienzler
Affiliation:
Forschungszentrum Karlsruhe, Institut für Nukleare Entsorgung, P.O. Box 3640, D-76021 Karlsruhe, Germany
Horst Geckeis
Affiliation:
Forschungszentrum Karlsruhe, Institut für Nukleare Entsorgung, P.O. Box 3640, D-76021 Karlsruhe, Germany
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Abstract

During long-term interim storage of spent fuel, pre-oxidation of the UO2-matrix may not be ruled out completely. This can happen if air could find access to the fuel in the case of cladding failure. The aim of this work is to study the impact of pre-oxidation of the fuel surface on the UO2 matrix dissolution rate and the associated mobilization or retention of radionuclides in highly concentrated salt solutions. The tests were performed with samples that suffered pre-oxidation during up to seven years. The dissolution rate of a fuel sample contacted by small quantities of air-oxygen was found to be roughly a factor of 10 higher in comparison to non oxidized samples, but concentrations of radionuclides, especially Pu and U were hardly affected. The majority of dissolved radionuclides, especially Pu, U appear to have been reimmobilized on the fuel sample itself.

Type
Research Article
Information
MRS Online Proceedings Library (OPL) , Volume 807: Symposium Scientific Basis for Nuclear Waste Management XXVII , 2003 , 53
Copyright
Copyright © Materials Research Society 2004

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References

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