I. Casas

The oxidative dissolution mechanism of uranium dioxide. I. The effect of temperature in hydrogen carbonate medium

Abstract 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 HCO3− at the U(VI) sites of the oxidized layer; and 3) detachment of the U(VI)-carbonato surface complex. From this mechanism a general rate equation has been derived: r= k 1 k 2 { >UO 2 } tot [ O 2 ][HCO 3 − ] k −1 +k 2 [ HCO 3 − ]+k 1 [ O 2 ] Which allows to rationalization of some of the discrepancies found in the literature concerning the dependence of the dissolution rate of uranium dioxide on the hydrogen carbonate concentration. The application of this bicarbonate-promoted oxidative dissolution mechanism allows us to predict unirradiated UO2 and spent nuclear fuel dissolution rates which are in satisfactory agreement with rates determined experimentally. Some differences observed in spent fuel dissolution rates have been attributed to radiolysis effects.

The role of pe, pH, and carbonate on the solubility of UO2 and uraninite under nominally reducing conditions

Abstract Experimental data obtained from uranium dioxide solubility studies as a function of pH and under nominally reducing conditions in a 0.008 mol dm −3 perchlorate medium and in a 1 mol dm −3 chloride solution are presented. The solubility of extensively characterized uraninite samples from Cigar Lake (Canada), Jachymov (Czech Republic), and Oklo (Gabon) was determined in a solution matching the composition of a groundwater associated with granitic terrain. The redox potential of the test solution was monitored throughout the experimental period. The results obtained were modeled using aqueous formation constants compiled by the NEA, using stability constants corrected to appropriate ionic strengths. The solubility curves have been adjusted by calculating the value of K s4 (UO 2(s) + 2H 2 O ⇔ U(OH) 4(aq) ) that gave the best fit with the experimental data. For a low temperature synthetic UO 2 , a value of logK s4 of −7.3 was determined, while for uraninites the best fit was obtained with a value of logK s4 of −8.5. A wide range of published UO 2 solubilities can be reproduced by the available database, where experimental conditions were adequately defined in the original experiments. A lower value of the solubility product of the uranium dioxide phase defined as fuel in the SKB uranium database provides reasonable solubilities for a wide span of experimental results at near to neutral pH. Based on the modeling and using the β 1,4 for the U(IV)-OH complexation given by Grenthe et al. (1992a) , a logK s0 (UO 2 (s) + 4H + ⇔ U 4+ + 2H 2 O) value of −2.3 ± 0.2 is proposed. Differences in solubility between natural and synthetic samples are attributed to the presence of carbonate in the experiments performed with uraninites, while differences in solubility observed among the natural samples can be correlated to radiation effects at atomic scale.

Kinetics of corrosion and dissolution of uranium dioxide as a function of pH

A continuous flow-through reactor with a thin layer of solid particles (size ranging from 100 to 300 μm) was used to obtain a deeper knowledge on the mechanism of dissolution of UO2 under oxidizing conditions. Using this methodology the dissolution rate of uranium dioxide was determined at three different oxygen partial pressures (5, 21, and 100% in nitrogen) and as a function of pH (between 3 and 12) in a noncomplexing medium. From the results of these experiments the following rate equation was derived: In addition, XPS characterizations were performed to determine the U(IV)/U(VI) ratio on the solid surface at different experimental times and conditions. These results showed that at acidic conditions (pH below 6.7) the final solid surface presents a stoichiometry close to UO2, while at alkaline conditions the final solid surface average composition is close to UO2.25. This information was integrated with the results of the leaching experiments to present a model for the mechanism of dissolution of uranium dioxide under the experimental conditions.

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.

The thermodynamics and kinetics of uranophane dissolution in bicarbonate test solutions

Abstract The thermodynamic and kinetic properties of a synthetic uranophane (Ca(H3O)2(UO2)2(SiO4)2 · 3H2O) have been determined from dissolution experiments in test solutions of different bicarbonate concentrations at 25°C. The experiments were performed using batch and continuously stirred tank flow-through reactors. From the experimental data obtained with the batch reactor the solubility constant for the reaction: Ca(H3O)2(UO2)2(SiO4)2 · 3H2O + 6H3O+ ⇔ Ca+2 + 2UO2+2 + 2H4SiO4 + 11H2O was determined to be log Ks00 = 11.7 ± 0.6 and the rate equation for the dissolution process is log rdissol (mol s−1 m−2) = −8.3 (±0.6) + 0.7 (±0.3) log [HCO3−]. By using the continuously stirred tank flow-through reactor we obtained a rate equation in reasonably good agreement with that obtained in the batch reactor: log rdissol (mol s−1 m−2) = −9.2 (±0.4) + 0.7 (±0.2) log [HCO3−].

Modelling of the Ni(II) removal from aqueous solutions onto grape stalk wastes in fixed-bed column.

Grape stalk wastes generated in the wine production process were used for the removal of nickel (II) from aqueous solution. The experimental breakthrough curves were obtained in fixed-bed columns. Experiments we carry out in order to evaluate the influence of inlet metal concentration (30 and 70 mg L(-1)) and the regeneration process in a double sorption cycle. The CXTFIT code was used to fit the experimental data and to determine the transport and sorption parameters of the convective-dispersive equation (CDE) and the two-site deterministic non-equilibrium (TSM/CDE) model by adjusting the models to the experimental breakthrough curves (BTC). The results showed that bed capacity as well as transport and sorption parameters were affected by the initial metal concentration, at the highest Ni(II) concentration the grape stalks column saturated quickly leading to earlier breakthrough. The sorption capacity of the sorbent was slightly reduced in a double sorption cycle, while the recovery of the metal in the desorption step was ranging between 80% and 85% in both cycles.

Estimation of the concentrations of trace metals in natural systems: The application of codissolution and coprecipitation approaches to El Berrocal (Spain) and Poços de Caldas (Brazil)

Abstract Trace element concentrations in natural systems indicate that, in most cases, solubility controls are not exerted by pure solid phases. The cycling of trace metals in the environment is coupled to the chemistry of the major components. To model this behaviour we have developed a methodology based on codissolution and coprecipitation approaches. We have applied this methodology to the El Berrocal (Spain) and Pocos de Caldas (Brazil) sites. Trace metals under investigation were: Ba, Cu, Mn, Ni, Sr, and U. The models took into account: (1) the observed association of copper and uranium with iron oxyhydroxides (both in El Berrocal and in Pocos de Caldas), (2) the interaction of barium and manganese with calcite in El Berrocal, and (3) the association between strontium and fluorite in Pocos de Caldas. In most cases, the results were in good agreement with field data, showing that application of these approaches can reproduce more accurately the measured concentration of trace metals in groundwater than the assumption of a pure solid phase controlling the concentration of trace metals in solution. However, some problems still remain, such as the description of nickel concentrations in groundwater.

Extraction of cadmium(II) by organophosphorus compounds

Abstract The extraction of cadmium(II) by di-(2-ethylhexyl) phosphoric acid dissolved in tetradecane from aqueous chloride and perchlorate solutions has been studied at 25°C. The distribution of the metal has been determined as a function of metal and DEHPA concentrations. Distribution data have been treated both graphically and numerically using the program LETAGROP-DISTR ( Acta Chem. Scand. 1971, 25 , 1521) and the composition of the extracted species into the organic phase has been determined. The extraction constants for these species are given in Table 1.

Sorption of strontium on uranyl peroxide: implications for a high-level nuclear waste repository

Strontium-90 is considered the most important radioactive isotope in the environment and one of the most frequently occurring radionuclides in groundwaters at nuclear facilities. The uranyl peroxide studtite (UO2O2 . 4H2O) has been observed to be formed in spent nuclear fuel leaching experiments and seems to have a relatively high sorption capacity for some radionuclides. In this work, the sorption of strontium onto studtite is studied as a function of time, strontium concentration in solution and pH. The main results obtained are (a) sorption is relatively fast although slower than for cesium; (b) strontium seems to be sorbed via a monolayer coverage of the studtite surface, (c) sorption has a strong dependence on ionic strength, is negligible at acidic pH, and increases at neutral to alkaline pH (almost 100% of the strontium in solution is sorbed above pH 10). These results point to uranium secondary solid phase formation on the spent nuclear fuel as an important mechanism for strontium retention in a high-level nuclear waste repository (HLNW).