Radiolytic modelling of spent fuel oxidative dissolution mechanism. Calibration against UO2 dynamic leaching experiments (original) (raw)
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Dissolution of irradiated fuel: a radiolytic mass balance study
Journal of Nuclear Materials, 1995
We have studied the production of H2,O 2 and H202 by radiolysis of the leach solution in a closed system containing fragments of irradiated PWR fuel and distilled water purged with argon. The experimental data is not reflected in the release of U(VI) to the solution, clearly indicating that most of the oxidant production has been taken up by the UO2 spent fuel surface. This proves that the UO 2 surface constitutes a major redox buffer capacity to prevent radiolytic oxidation under repository conditions. 0022-3115/95/$09.50
Modelling Oxidative Dissolution of Spent Fuel
MRS Proceedings, 1996
Spent nuclear fuel will, by the radiation, split nearby water into oxidizing and reducing compounds. The reducing compounds are mostly hydrogen that will diffuse away. The remaining oxidizing compounds can oxidize the uranium oxide of the fuel and make it more soluble. The oxidised uranium will dissolve and diffuse away. The nuclides previously incorporated in the spent fuel matrix can then be released and also migrate away from the fuel.
The effect of fuel chemistry on UO2 dissolution
Journal of Nuclear Materials, 2016
The dissolution rate of both unirradiated UO 2 and used nuclear fuel has been studied by numerous countries as part of the performance assessment of proposed geologic repositories. In the scenario of waste package failure and groundwater contact with the fuel, the effects of variables such as temperature, dissolved oxygen, and water and fuel chemistry on the dissolution rates of the fuel are necessary to provide a quantitative estimate of the potential release over geologic time frames. The primary objective of this research was to determine the influence these parameters, with primary focus on the fuel chemistry, have on the dissolution rate of unirradiated UO 2 under oxidizing repository conditions and compare them to the rates predicted by current dissolution models.
A multiphase interfacial model for the dissolution of spent nuclear fuel
Journal of Nuclear Materials, 2015
The Fuel Matrix Dissolution Model (FMDM) is an electrochemical reaction/diffusion model for the dissolution of spent uranium oxide fuel. The model was developed to provide radionuclide source terms for use in performance assessment calculations for various types of geologic repositories. It is based on mixed potential theory and consists of a two-phase fuel surface made up of UO 2 and a noble metal bearing fission product phase in contact with groundwater. The corrosion potential at the surface of the dissolving fuel is calculated by balancing cathodic and anodic reactions occurring at the solution interfaces with UO 2 and NMP surfaces. Dissolved oxygen and hydrogen peroxide generated by radiolysis of the groundwater are the major oxidizing agents that promote fuel dissolution. Several reactions occurring on noble metal alloy surfaces are electrically coupled to the UO 2 and can catalyze or inhibit oxidative dissolution of the fuel. The most important of these is the oxidation of hydrogen, which counteracts the effects of oxidants (primarily H 2 O 2 and O 2). Inclusion of this reaction greatly decreases the oxidation of U(IV) and slows fuel dissolution significantly. In addition to radiolytic hydrogen, large quantities of hydrogen can be produced by the anoxic corrosion of steel structures within and near the fuel waste package. The model accurately predicts key experimental trends seen in literature data, the most important being the dramatic depression of the fuel dissolution rate by the presence of dissolved hydrogen at even relatively low concentrations (e.g., less than 1 mM). This hydrogen effect counteracts oxidation reactions and can limit fuel degradation to chemical dissolution, which results in radionuclide source term values that are four or five orders of magnitude lower than when oxidative dissolution processes are operative. This paper presents the scientific basis 3 of the model, the approach for modeling used fuel in a disposal system, and preliminary calculations to demonstrate the application and value of the model.
Journal of Nuclear Materials, 2005
The objective of this work is to study the UO 2 oxidation by O 2 and dissolution in bicarbonate media and to extrapolate the results obtained to improve the knowledge of the oxidative dissolution of spent nuclear fuel. The results obtained show that in the studied range the oxygen consumption rate is independent on the bicarbonate concentration while the UO 2 dissolution rate does depend on. Besides, at 10 À4 mol dm À3 bicarbonate concentration, the oxygen consumption rate is almost two orders of magnitude higher than the UO 2 dissolution rate. These results suggest that at low bicarbonate concentration (<10 À2 mol dm À3 ) the alteration of the spent nuclear fuel cannot be directly derived from the measured uranium concentrations in solution. On the other hand, the study at low bicarbonate concentrations of the evolution of the UO 2 surface at nanometric scale by means of the SFM technique shows that the difference between oxidation and dissolution rates is not due to the precipitation of a secondary solid phase on UO 2 .
Journal of Nuclear Materials, 2012
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Journal of Nuclear Materials, 1996
The effect of H20 2, NaC10 and Fe on the dissolution of unirradiated UO2(s) in NaC1 5 mol kg-J has been studied at neutral to alkaline pH. Dissolution rates have been determined as a function of oxidant concentration. A general equation to correlate both parameters has been obtained: log r = (-8.0 + 0.2)+ log[Ox] 0"93+ 0.07. The values obtained have been compared to those given for spent fuel under the same experimental conditions. The effect of iron is similar in both unirradiated UO 2 and spent fuel with a final uranium concentration around 5 × 10 -8 mol kg-~ which corresponds to the solubility value of UO2(f) under reducing conditions.