J. Giménez

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.

Kinetic Studies of Unirradiated UO2 Dissolution under Oxidizing Conditions in Batch and Flow Experiments

The U02-matrix dissolution rate is a critical parameter to predict the stability of nuclear spent fuel under final disposal conditions. In this study, the kinetic of dissolution of unirradiated U02 has been studied in 0.01 mol L NaC104 at 298 Κ and under oxidizing conditions in sequential batch and continuous flow experiments. Initial dissolution rates have been attributed to the dissolution of an oxidized layer on the solid surface, while final dissolution rates correspond to the dissolution of U307. X-Ray Photoelectron Spectroscopy studies confirmed the changes of solid surface during leaching experiments. A general mechanism of oxidation-dissolution is proposed based on experimental determinations.

Determination of the equilibrium formation constants of two U(VI)–peroxide complexes at alkaline pH

The formation of uranyl-peroxide complexes was studied at alkaline media by using UV-Visible spectrophotometry and the STAR code. Two different complexes were found at a H(2)O(2)/U(VI) ratio lower than 2. A graphical method was used in order to obtain the formation constants of such complexes and the STAR program was used to refine the formation constants values because of its capacity to treat multiwavelength absorbance data and refining equilibrium constants. The values obtained for the two complexes identified were: UO(2)(2+) + H(2)O(2) + 4OH(-) <−> UO(2)(O(2))(OH)(2)(2-) + 2H(2)O: log β°(1,1,4) = 28.1 ± 0.1 (1). UO(2)(2+) + 2H(2)O(2) + 6OH(-) <−> UO(2)(O(2))(2)(OH)(2)(4-) + 4H(2)O: log β°(1,2,6) = 36.8 ± 0.2 (2). At hydrogen peroxide concentrations higher than 10(-5) mol dm(-3), and in the absence of carbonate, the UO(2)(O(2))(2)(OH)(2)(4-) complex is predominant in solution, indicating the significant peroxide affinity of peroxide ions for uranium and the strong complexes of uranium(VI) with peroxide.

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.

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.

Surface Site Densities of Uranium Oxides: UO 2 , U 3 O 8

The estimation of the surface site density for UO 2 (main component of the SF matrix) is an important aspect to take into account in the development of radiolytical models for the dissolution of the spent nuclear fuel (SF). Also, other oxides can be formed on the SF surface due to the effect of radiolytically-formed oxidizing species. Due to the lack or reliable data in literature we have studied the surface site densities of two uranium oxides: UO 2 and U 3 O 8 . For this determination, the knowledge of both the reactive surface area and the surface charge of the solid are necessary. In this work the reactive surface area of the two solids was determined by the BET method while the surface charge was determined by potentiometric acid-base titrations of suspensions of the solids at different ionic strengths (0.001, 0.01, and 0.1 mol·dm −3 ) under N 2 atmosphere. The amount of adsorbed protons was calculated by subtracting blank titration from thesolid suspension titration. The single-site nonelectrostatic model (NEM) was used to describe titration data. Uranium speciation in solution was included in the model as well. The surface area values obtained were 0.15 ± 0.01 m 2 ·g −1 for UO 2 and 0.77 ± 0.02 m 2 ·g −1 for U 3 O 8 , while the surface acidic site densities were determined to be 165 ± 10 sites·nm −2 and 48 ± 3 sites·nm −2 for UO 2 and U 3 O 8 , respectively.

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.

RN Fractional Release of High Burn-Up Fuel: Effect of HBS and Estimation of Accessible Grain Boundary

The so-called Instant Release Fraction (IRF) is considered to govern the dose released from Spent Fuel repositories. Often, IRF calculations are based on estimations of fractions of inventory release based in fission gas release [1]. The IRF definition includes the inventory located within the Gap although a conservative approach also includes both the Grain Boundary (GB) and the pores of restructured HBS inventories. A correction factor to estimate the fraction of Grain Boundary accessible for leaching has been determined and applied to spent fuel static leaching experiments carried out in the ITU Hot Cell facilities [2]. Experimental work focuses especially on the different properties of both the external rim area (containing the High Burn-up Structure (HBS)) and the internal area, to which we will refer as Out and Core sample, respectively. Maximal release will correspond to an extrapolation to simulate that all grain boundaries or pores are open and in contact with solution. The correction factor has been determined from SEM studies taking into account the number of particles with HBS in Out sample, the porosity of HBS particles, and the amount of transgranular fractures during sample preparation.