M.E. Torrero

Solid surface evolution model to predict uranium release from unirradiated UO2 and nuclear spent fuel dissolution under oxidizing conditions

The dissolution of UO2 under oxidizing conditions has been studied in the last years in different waste disposal conditions. These studies have indicated the importance of the solid surface evolution during leaching experiments. In this work, a mathematical model based on X-ray photoelectron spectroscopy determinations of the solid surface was developed. This model allows the uranium release under oxidizing conditions at acidic pH or in carbonate medium to be predicted. At alkaline pH without carbonate, the formation of a UO2.33 surface layer and its equilibrium with the uranium concentration in solution could be responsible for the disagreement observed between the model and the experimental data. This model has been also applied to uranium release from spent nuclear fuel dissolution experiments carried out in granitic groundwater.

Effect of H2O2, NaClO and Fe on the dissolution of unirradiated UO2 in NaCl 5 mol kg−1. Comparison with spent fuel dissolution experiments

Copyright (c) 1996 Elsevier Science B.V. All rights reserved. The effect of H 2 O 2 , NaClO and Fe on the dissolution of unirradiated UO 2 (sr in NaCl 5 mol kg −1 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=(−8.0p0.2r+log[Ox 0.93& plusmn; 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 −1 which corresponds to the solubility value of UO 2 (fr under reducing conditions.

Uranium (iv) Dioxide and Simfuel as Chemical Analogues of Nuclear Spent Fuel Matrix Dissolution. A Comparison of Dissolution Results in a Standard Naci/NaHCO 3 Solution

We have carried out an experimental comparison study of the dissolution rates of unirradiated UO 2 and SIMFUEL pellets and particles (100–300 μm) in a standard NaCI/NaHC0 3 solution, under oxidizing conditions. We have performed the experiments using batch and flow methodologies. Both methodologies gave similar results, indicating that the overall oxidation/dissolution process is the same in both cases. The results from the experiments indicate that under these conditions the dissolution process is both oxygen and bicarbonate promoted. The dissolution rates we obtained are: R=2.4 ± 0.8 mg U/m 2 d for U0 2 and R= 0.17 ± 0.05 mg U/m 2 d for SIMFUEL. The results of the experiments indicate that the dissolution rate under oxic conditions is clearly dependent on the number of U(VI) surface sites which for spent nuclear fuel is a function of the extent of radiolytic oxidation.

Dissolution studies of soddyite as a long-term analogue of the oxidative alteration of the spent nuclear fuel matrix

The thermodynamic and kinetic dissolution properties of a synthetically obtained soddyite have been determined at different bicarbonate concentrations. This uranium-silicate is expected to be a secondary solid phase of the oxidative alteration pathway of uranium dioxide in waters with low phosphate content and, consequently, it is likely to constitute one of the long-term uranium solubility limiting solid phases. The experimental data obtained at the end of the experiments correspond fairly well to the theoretical model calculated with a log K{sub S0}{sup 0} of 3.9 {+-} 7. On the other hand, the general trend of the total uranium in solution measured in the experiments as a function of time has been fitted by using a kinetic equation obtained from the principle of detailed balancing of the dissolution reaction. In addition, the EQ3/6 code has also been used to model the uranium concentrations as a function of time. In both modeling exercises comparable results were obtained. The dissolution rate, normalized to the total surface area used in the experiments as measured with the BET method, gave an average value of 6.8 ({+-}4.4) 10{sup {minus}14} mol/cm{sup 2} s.

Effect of temperature and bicarbonate concentration on the kinetics of UO{sub 2}(s) dissolution under oxidizing conditions

The dissolution rate of unirradiated UO{sub 2}(s) has been studied as a function of hydrogen carbonate concentration at three different temperatures (298.15 K, 313.15 K and 333.15 K) under oxidizing conditions in a continuous flow-through reactor with a thin layer of solid particles (particle size from 100 to 300 {micro}m). From the results of these experiments, two different rate laws have been determined. At high temperature (313.15 K and 333.15 K), the authors obtained a dissolution rate proportional to hydrogen carbonate concentration while at 298.15 K, the rate almost depends on the square root of the hydrogen carbonate concentration. This indicates a different reaction mechanism depending on temperature which can be related to the oxidation step of the overall process. The apparent activation energy obtained was 41 kJ/mol.

Fluorimetric determination of traces of uranium(VI) in brines and iron(III) oxides using separation on an activated silica gel column

Abstract A method for the determination of trace uranium levels in brines and waters with a high content of iron(III) based on the separation of uranium from a mixture of masking agents by sorption on an activated silica gel column was developed in connection with the Scintrex UA-3 uranium analyser. The detection limit of uranyl ions was 1.4 ng. This value can be achieved even with solutions containing 56 mg of iron, 480 mg of magnesium and/or 1000 mg of chloride. The relative standard deviation of the method is 5%. An analysis requires less than 20 min.

Dissolution of UO 2 (s) in MgCl 2 -Brines Under Different Redox Conditions.

The dissolution of unirradiated UO 2 (s), with a particle size of 1 mm, has been studied in MgCl 2 brines at 298 K under both reducing and oxidizing conditions. Results obtained under reducing conditions (H atmosphere in the presence of a palladium catalyst) show an initial increase of the total uranium concentration in solution and a subsequent decrease until equilibrium (or steady state) values are reached. Results obtained under oxidizing conditions (nominal oxygen partial pressures of 0.05, 0.21 and 1 atm) show two different trends. A relatively fast initial dissolution rate and, after approximately two or three weeks, a slower dissolution rate. X-Ray Photoelectron Spectroscopy (XPS) has shown that the UO 2 surface composition changes during the experiment.

Conceptual and Mathematical Model for the UO2(s) Dissolution in Brines Under Different Redox Conditions

The dissolution of unirradiated U02(s) in two different salt brines, a NaCl-brine and a MgCl2-brine, under different redox conditions has been modeled. The experimental data have been modeled taking into account two different processes: (1) an initial dissolution due to the existence of an oxidized layer initially present on the U02(s) surface and (2) the evolution of the uranium concentration in solution to reach the solubility of the thermodynamically stable solid phase. The good agreement between the model and the experimental data allows the establishment of an U02(s) dissolution mechanism under the experimental conditions. Results have shown that the long-term dissolution rate depends on the square root of the oxygen partial pressure. This fractional dependence has been attributed to the fact that the rate limiting step is the adsorption of the oxygen on the surface of the solid and the diffusion of this oxygen through the U02(s) lattice.

Mechanism of unirradiated UO{sub 2} (s) dissolution in NaCl and MgCl{sub 2} brines at 25 C

The dissolution of unirradiated UO{sub 2} (s) has been studied in NaCl and MgCl{sub 2} brines under both reducing and oxidizing conditions. The initial uranium release under reducing conditions has been attributed to the dissolution of an initial oxidized layer. The final uranium concentrations have been modeled by using the PHRQPITZ computer program giving the solubility of the solid phase UO{sub 2} (s). Under oxidizing conditions, the initial release is the sum of the oxidized layer dissolution and the oxidation/dissolution of the UO{sub 2}. The release rates calculated are 1.4 {times} 10{sup {minus}5} mol/m{sup 2} in NaCl-brine and 3.6 {times} 10{sup {minus}5} mol/m{sup 2} in MgCl{sub 2}-brine. After the initial release, uranium concentration in the NaCl-brine reaches a constant value, which has been attributed to the formation of a secondary solid phase. In MgCl{sub 2}-brine, the uranium concentration increases slowly indicating, in this case, no control by secondary phase formation.

The Solubility of Unirradiated UO 2 In Both Perchlorate And Chloride Test Solutions. Influence of the Ionic Medium

The solubility of a crystalline unirradiated UO 2 (s) has been studied under reducing conditions at 25°C in three different ionic media: 0.008 mol dm –3 NaCIO 4 , 1 and 5 mol dm -3 NaCl + . The species responsible for the solubility in perchlorate medium are U(OH) 3 and U(OH) 4 , with the stability constants log β 13 = -0.4(±0.2) and log β 14 = -5.7(±0.1), respectively. The solubility in both1! and 5 M chloride media is explained at pH values higher than 6 by the species U(OH) 4 with log β 14 = -5.5(±0.2) while at lower pH values an abnormal behaviour is observed, with higher uranium concentrations in solution and a proton independent solubility for pH values lower than 4. The XPS observations of the reacted solids do not show the presence of solid surface phases other than uranium dioxide, with an upper oxidation state of UO 2.1 .