Measurement of Dissolution Rates of UO 2 Matrix with the Isotope Dilution Method
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0HDVXUHPHQWRI'LVVROXWLRQ5DWHVRI820DWUL[ZLWKWKH,VRWRSH'LOXWLRQ0HWKRG K. Ollila1 and V. Oversby2 VTT Processes, Technical Research Centre of Finland, PO Box 1608, FIN-02044 VTT, Finland 2 VMO Konsult, Karlavägen 70, SE-11459 Stockholm, Sweden $%675$&7 The objective of this work was to measure the actual rates of matrix dissolution of UO2 in synthetic groundwater by using an isotope dilution method to provide a quantitative estimate of precipitation effects. A preliminary series of tests was performed under oxidising conditions, in contact with air. The second series of tests was under reducing conditions produced by actively corroding iron. The solid phases were fragments of unirradiated fuel pellets and intact pellets. The aqueous phase was a dilute, synthetic groundwater - so-called modified Allard water that is buffered by sodium bicarbonate/carbonate. This paper gives results for the experiments under oxidising conditions and preliminary tests under reducing conditions. In contact with air, the U concentrations reached higher levels than measured in previous experiments with spent fuel or with UO2 pellets. The comparison of the U concentrations calculated from isotopic ratios with the experimental results suggests precipitation has begun at later stages of restarted tests. The measurements in the presence of actively corroding iron gave very low concentrations in the aqueous phase. At contact times longer than one week, all U seemed to disappear from solution and sorb or precipitate on UO2 or Fe surfaces in the test vessel. ,1752'8&7,21 Spent fuel arising from the operation of power reactors, e.g. in Finland and Sweden, is planned to be disposed of in a repository to be constructed at a depth of about 500 metres in crystalline granitic bedrock. The fuel assemblies will be placed in canisters consisting of an outer corrosion-resistant copper shell and an inner cast iron form, which gives mechanical strength. The canister will be placed in a disposal borehole and the space between canister and rock will be filled with compacted bentonite. Any oxygen trapped in the repository will be consumed by reaction with the host rock and/or pyrite in the bentonite, giving long-term conditions with low redox potential. Under such conditions, the uranium dioxide matrix of spent fuel is a stable phase with extremely low solubility. Radiolysis of water due to the alpha activity of the fuel has been proposed as a means of inducing locally oxidising conditions at the surface. This may cause UO2 to dissolve in the more soluble U(VI) oxidation state. The solubility of U(VI) is enhanced in the presence of bicarbonate/carbonate [1,2]. In safety analyses, it is commonly assumed that U(VI) once formed will be released from the canister. The potential role of the canister material, iron, and its corrosion products in reducing oxidised, more-soluble actinide species is poorly understood. The experiments described in this paper are the first part of a project that will continue using UO2 doped with 233U to measure the
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