The oxidative dissolution mechanism of uranium dioxide. I. The effect of temperature in hydrogen carbonate medium (original) (raw)
Journal of Nuclear Materials, 2005
The effects of alpha dose-rate on UO 2 dissolution were investigated by performing dissolution experiments with 238 Pu-doped UO 2 materials containing nominal alpha-activity levels of $1-100 Ci/kg UO 2 (actual levels 0.4-80 Ci/kg UO 2 ), in 0.1 M NaClO 4 and in 0.1 M NaClO 4 + 0.1 M carbonate. Dissolution rates increased less than 10-fold for an almost 100-fold increase in doping level and fall within the range of predictions of the Mixed Potential Model (a detailed mechanistic model for used fuel dissolution). Dissolution rates were lower in carbonate-free solutions and enrichment of 238 Pu on the UO 2 surface was suggested in carbonate solutions. Effective G values, defined as the ratio of the total amount of U dissolved divided by the maximum possible amount of U dissolved by radiolytically produced H 2 O 2 , increased with decreasing doping levels. This suggests that the dissolution reaction at high dose rates is limited by the reaction rate between UO 2 and H 2 O 2 , but becomes increasingly limited by the rate of production of H 2 O 2 at lower dose rates.
The effect of ion irradiation on the dissolution of UO2 and UO2-based simulant fuel
Journal of Alloys and Compounds, 2018
The aim of this work was to study the separate effect of fission fragment damage on the dissolution of simulant UK advanced gas-cooled reactor nuclear fuel in water. Plain UO 2 and UO 2 samples, doped with inactive fission products to simulate 43 GWd/tU of burn-up, were fabricated. A set of these samples were then irradiated with 92 MeV 129 Xe 23þ ions to a fluence of 4.8 Â 10 15 ions/cm 2 to simulate the fission damage that occurs within nuclear fuels. The primary effect of the irradiation on the UO 2 samples, observed by scanning electron microscopy, was to induce a smoothening of the surface features and formation of hollow blisters, which was attributed to multiple overlap of ion tracks. Dissolution experiments were conducted in single-pass flow-through (SPFT) mode under anoxic conditions (<0.1 O 2 ppm in Ar) to study the effect of the induced irradiation damage on the dissolution of the UO 2 matrix with data collection capturing six minute intervals for several hours. These time-resolved data showed that the irradiated samples showed a higher initial release of uranium than unirradiated samples, but that the uranium concentrations converged towards~10 À9 mol/l after a few hours. Apart from the initial spike in uranium concentration, attributed to irradiation induced surficial micro-structural changes, no noticeable difference in uranium chemistry as measured by X-ray electron spectroscopy or 'effective solubility' was observed between the irradiated, doped and undoped samples in this work. Some secondary phase formation was observed on the surface of UO 2 samples after the dissolution experiment.
Journal of Nuclear Materials, 2012
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On the Stability of Uranium Carbide in Aqueous SolutionEffects of HCO3 and H2O2
Uranium carbide (UC) is a candidate fuel material for future Generation IV nuclear reactors. As part of a general safety assessment, it is important to understand how fuel materials behave in aqueous systems in the event of accidents or upon complete barrier failure in a geological repository for spent nuclear fuel. As irradiated nuclear fuel is radioactive, it is important to consider radiolysis of water as a process where strongly oxidizing species can be produced. These species may display high reactivity toward the fuel itself and thereby influence its integrity. The most important radiolytic oxidant under repository conditions has been shown to be H 2 O 2. In this work, we have studied the dissolution of uranium upon exposure of UC powder to aqueous solutions containing HCO 3 − and H 2 O 2 , separately and in combination. The experiments show that UC dissolves quite readily in aqueous solution containing 10 mM HCO 3 − and that the presence of H 2 O 2 increases the dissolution further. UC also dissolves in pure water after the addition of H 2 O 2 , but more slowly than in solutions containing both HCO 3 − and H 2 O 2. The experimental results are discussed in view of possible mechanisms.
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.
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.
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.
Behavior of Uranium Dioxide: Chemistry and Catalysis in the UO2-water System
MRS Proceedings, 2003
ABSTRACTInteractions during extended exposure of UO2 to 2:1 H2+O2 mixtures at room temperature and 0.13 bar pressure are investigated in an effort to describe chemical and kinetic behavior of spent fuel following contact with groundwater. Oxidation of UO2 to UO2+x by O2 occurs initially when oxide is directly exposed to the gas mixture or submerged in water, but immersion is accompanied by a 25-fold reduction in the rate. The initial rate is proportional to [O2]2 for gasphase oxidation and to [O2]1.5 for the submerged oxidation. Continued measurement during direct oxide-gas contact indicates sequential reactions in which the UO2+x product is further oxidized by H2O and ultimately reacts with H2 to form an oxide hydride.
Journal of Nuclear Materials, 2005
Calibration and testing are inherent aspects of any modelling exercise and consequently they are key issues in developing a model for the oxidative dissolution of spent fuel. In the present work we present the outcome of the calibration process for the kinetic constants of a UO 2 oxidative dissolution mechanism developed for using in a radiolytic model. Experimental data obtained in dynamic leaching experiments of unirradiated UO 2 has been used for this purpose. The iterative calibration process has provided some insight into the detailed mechanism taking place in the alteration of UO 2 , particularly the role of Å OH radicals and their interaction with the carbonate system. The results show that, although more simulations are needed for testing in different experimental systems, the calibrated oxidative dissolution mechanism could be included in radiolytic models to gain confidence in the prediction of the long-term alteration rate of the spent fuel under repository conditions.
The dissolution of uranium oxides: Thermodynamic and kinetic investigations
Hydrometallurgy, 2016
This work investigates the dissolution of various uranium oxides in nitric acid medium and the most predominant occurring reaction was determined on the basis of the thermodynamic and kinetic studies. Six uranium oxides were dissolved and studied. The Gibbs free energies of all the reactions Δ r G°(T) were analyzed by Ulich model (Ulich, 1930) and the predominant dissolution reaction was found to be: 3U 3 O 8(s) + 20HNO 3(aq) → 9 UO 2 (NO 3) 2(aq) + 2NO (g) + 10H 2 O Three reaction order rate models namely the first, the second and the third order were applied on the predominant reaction reported above. According to the kinetics results, our reaction best fits the second order equation rate.