Radiolytic modelling of spent fuel oxidative dissolution mechanism. Calibration against UO2 dynamic leaching experiments (original) (raw)
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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 .
Modelling the Oxidative Dissolution of UO2
MRS Proceedings, 1999
An electrochemically based model for predicting the effects of α-radiolysis, the precipitation of U(VI) corrosion products and redox processes with Fe and Fe(II) on the dissolution of UO2 is described. Various aspects of the model are presented, including: the underlying mechanism, the reaction-diffusion equations used to describe the mass transport and homogeneous reactions of the various species considered in the model, the geometrical grid used to simulate both experimental and used fuel/container geometries and the electrochemical boundary conditions used for the numerical solution of the reaction-diffusion equations. The results of preliminary simulations are also discussed.
The oxidative dissolution of unirradiated UO2 by hydrogen peroxide as a function of pH
Journal of Nuclear Materials, 2005
The dissolution of non-irradiated UO 2 was studied as a function of both pH and hydrogen peroxide concentration (simulating radiolytic generated product). At acidic pH and a relatively low hydrogen peroxide concentration (10 À5 mol dm À3 ), the UO 2 dissolution rate decreases linearly with pH while at alkaline pH the dissolution rate increases linearly with pH. At higher H 2 O 2 concentrations (10 À3 mol dm À3 ) the dissolution rates are lower than the ones at 10 À5 mol dm À3 H 2 O 2 , which has been attributed to the precipitation at these conditions of studtite (UO 4 AE 4H 2 O, which was identified by X-ray diffraction), together with the possibility of hydrogen peroxide decomposition. In the literature, spent fuel dissolution rates determined in the absence of carbonate fall in the H 2 O 2 concentration range 5 · 10 À7 -5 · 10 À5 mol dm À3 according to our results, which is in agreement with H 2 O 2 concentrations determined in spent fuel leaching experiments.
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 kinetic model for the stability of spent fuel matrix under oxic conditions
Journal of Nuclear Materials, 1996
A kinetic model for the UO2-spent fuel dissolution has been developed by integrating all the fundamental and experimental evidence about the redox buffer capacity of the UO 2 matrix itself within the methodological framework of heterogeneous redox reactions and dissolution kinetics. The purpose of the model is to define the geochemical stability of the spent fuel matrix and its resistance to internal and external disturbances. The model has been built in basis the reductive capacity (RDC) of the spent fuel/water system. A sensitivity analysis has been performed in order to identify the main parameters that affect the RDC of the system, the oxidant consumption and the radionuclide release. The number of surface co-ordination sites, the surface area to volume ratio, the kinetics of oxidants generation by radiolysis and the kinetics of oxidative dissolution of UO 2, have been found to be the main parameters that can affect the reductive capacity of the spent fuel matrix. The model has been checked against some selected UO 2 and spent fuel dissolution data, performed under oxidizing conditions. The results are quite encouraging.
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.
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
Journal of Nuclear Materials, 2012
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Physical and Chemical Aspects of Radiation Induced Oxidative Dissolution of UO2
2006
The general subject of this thesis is oxidative dissolution of UO 2. The dissolution of UO 2 is mainly investigated because of the importance of the UO 2 matrix of spent nuclear fuel as a barrier against radionuclide release in a future deep repository. U(IV) is extremely insoluble under the reducing conditions prevalent in a deep repository, whereas U(VI) is more soluble. Hence, oxidation of the UO 2-matrix will affect its solubility and thereby its function as a barrier. In this thesis the relative efficiency of one-and two electron oxidants in dissolving UO 2 is studied. The oxidative dissolution yield of UO 2 was found to differ between one-and two-electron oxidants. At low oxidant concentrations the dissolution yields for one-electron oxidants are significantly lower than for two-electron oxidants. However, the dissolution yield for one-electron oxidants increases with increasing oxidant concentration, which could be rationalized by the increased probability for two consecutive one-electron oxidations at the same site and the increased possibility for disproportionation. This licentiate thesis is based on the following publications:
Geochimica Et Cosmochimica Acta, 1999
The oxidative dissolution of uranium (IV) dioxide has been experimentally investigated as a function of hydrogen carbonate concentration at 4 different temperatures (10, 25, 45, and 60°C) by using a continuous thin-layer flow-through reactor. The experimental results have been interpreted as evidence for a bicarbonate-promoted oxidative dissolution mechanism which can be differentiated in to 3 steps: 1) initial oxidation of the uranium dioxide solid surface; 2) binding of HCO 3