UO2 Surface Research Articles

Uranium secondary phase formation during anoxic hydrothermal leaching processes of UO2 nuclear fuel

Abstract Mobilisation of uranium in geologic environments from UO2 solid phases usually takes place by oxidative dissolution involving a change of U oxidation state from +IV to +VI; however, anoxic or reducing geochemical conditions are expected in many of the planned European disposal sites. This work investigates potential alteration mechanisms of UO2 in contact with groundwater ions (Ca2+, CO 3 2 - , and silicate) under anoxic conditions, at ambient (25 °C) and hydrothermal (180 °C) temperature conditions. SEM–EDX analysis detected (in the case of treatment at 180 °C in high silicate content solutions) a compound with U:Si ratio of 1:1 on the UO2 surfaces after leaching. Minor quantities of phases containing U, Ti, Fe, Si, and Ca were formed, these could not be characterized completely. A further experiment, performed in the presence of dissolving CaO/TiO2/SiO2/Fe(0)–Fe2O3, formed a compound with U:Si:Ca of 1:2:8, a ratio not matching any known uranyl compound. The two phases, possibly identical with coffinite, USiO4, and U-bearing ekanite, UCa2Si8O20, were found to form at different [Ca]/[Si] conditions. The implications upon the final geologic storage of actual, heterogeneous spent fuel are discussed.

Theoretical Studies of Hydrogen and Water Adsorption on Actinide Oxide Surfaces

AbstractThe adsorption of H2O on UO2 and PuO2 surfaces is studied using density functional theoretical approaches. Periodic slab models are employed to represent the oxide surface in which five layers of the oxide are explicitly treated. For both oxides the dihydroxyl form corresponding to dissociated H2O is more stable than the molecularly adsorbed H2O on the (111) and (110) surfaces. Adsorption of H2 on PuO2 and Pu2O3 is also investigated where stable forms corresponding to dissociated H atoms are found.

Effect of alpha irradiation on UO2 surface reactivity in aqueous media

Abstract The option of direct disposal of spent nuclear fuel in a deep geological formation raises the need to investigate the long-term behavior of the UO2 matrix in aqueous media subjected to α-β-γ radiation. The β-γ emitters account for most of the activity of spent fuel at the moment it is removed from the reactor, but diminish within a millennial time frame by over three orders of magnitude to less than the long-term activity. The latter persists over much longer time periods and must therefore be taken into account over a geological disposal time scale. Leaching experiments with solution renewal were carried out on UO2 pellets doped with alpha emitters (238Pu and 239Pu) to quantify the impact of alpha irradiation on UO2 matrix alteration. Three batches of doped UO2 pellets with different alpha flux levels (3.30×104, 3.30×105, and 3.2×106 α cm-2 s-1) were studied. The results obtained in aerated and deaerated media immediately after sample annealing or interim storage in air provide a better understanding of the UO2 matrix alteration mechanisms under alpha irradiation. Interim storage in air of UO2 pellets doped with alpha emitters results in variations of the UO2 surface reactivity, which depends on the alpha particle flux at the interface and on the interim storage duration. The variation in the surface reactivity and the greater uranium release following interim storage cannot be attributed to the effect of alpha radiolysis in aerated media since the uranium release tends toward the same value after several leaching cycles for the doped UO2 pellet batches and spent fuel. Oxygen diffusion enhanced by alpha irradiation of the extreme surface layer and/or radiolysis of the air could account for the oxidation of the surface UO2 to UO2+x. However, leaching experiments performed in deaerated media after annealing the samples and preleaching the surface suggest that alpha radiolysis does indeed affect the dissolution, which varies with the flux at the UO2/water interface.

Effects of Fe(II) and Hydrogen Peroxide Interaction upon Dissolving UO2under Geologic Repository Conditions

Iron redox cycling is supposed to be one of the major mechanisms that control the geochemical boundary conditions in the near field of a geologic repository for UO2 spent nuclear fuel. This work investigates the impact of reactions between hydrogen peroxide (H2O2) and iron (Fe2+/Fe3+) on UO2 dissolution. The reaction partners were contacted with UO2 in oxygen-free batch reactor tests. The interaction in absence of UO2 gives a stoichiometric redox reaction of Fe2+ and H2O2 when the reactants are present in equal concentration. Predomination of H202 results in its delayed catalytic decomposition. With UO2 present, its dissolution is controlled by either a slow mechanism (as typical for anoxic environments) or uranium peroxide precipitation, depending strongly on the reactant ratio. Uranium peroxide (UO4 x nH2O, m-studtite), detected on UO2 surfaces after exposure to H2O2, was not found on the surfaces exposed to solutions with stoichometric Fe(II)/ H2O2 ratios. This suggests that H2O2 was deactivated in redox reactions before a formation of UO4 took place. ESR measurements employing the spin trapping technique revealed only the DMPO-OH adduct within the first minutes after the reaction start (high initial concentrations of the OH radical); however, in the case of Fe(II) and H2O2 reacting at 10(-4) mol/L with UO2, dissolved oxygen and Fe2+ concentrations indicate the participation of further Fe intermediates and, therefore, Fenton redox activities.

The reduction of U(VI) by near field hydrogen in the presence of UO2(s)

Summary In several recent works a large impact of dissolved hydrogen on spent fuel dissolution has been observed: the solution concentrations of all redox sensitive components of the spent fuel decrease with time, while gas phase analyses indicate levels of radiolytic oxygen below the detection limit. This indicates that at the relatively low temperatures of these studies (≤70 °C) hydrogen is activated. The aim of this study is to test one of the proposed mechanisms for hydrogen activation, namely by UO2(s) surfaces, which may act as hydrogen catalysts. Dissolved U(VI) carbonate species were used as oxidized species in order to test the reducing ability of hydrogen in the presence and absence of UO2(s). In most cases parallel experiments with the same conditions, but under Ar(g) atmosphere were carried out as blanks. Quartz lined autoclaves avoiding any contact of the solution with metallic parts were used. The experiments were carried out in two stages: first the stability of the (10-7-10-5) M U(VI) in 10 mM NaCl, 2 mM NaHCO3 solutions in the presence of dissolved hydrogen was tested. Then a gold net basket containing UO2(s) fragments was introduced in the autoclave and the concentration of U(VI) was followed at different time intervals by ICP-MS. Various dissolved hydrogen concentrations, temperatures and U(VI) concentrations were used, always spanning the range of these parameters in a proposed deep rock repository. Especially at the lowest U(VI) concentrations tested (∼100 ppb) the activation of dissolved hydrogen by UO2(s) surfaces is clearly visible. As a conclusion of this work it may be stated that in the temperature range (20-70 °C) investigated a) dissolved hydrogen does not reduce U(VI) carbonate species in the absence of a catalyst and b) in the presence of UO2(s) surfaces dissolved hydrogen reduces U(VI) in carbonate solutions, very probably to UO2(s).

Segregation of Ba2+, Sr2+, Ce4+ and Zr4+ to UO2 surfaces

Segregation of the cation substitution defectsBa2+,Sr2+,Ce4+ andZr4+ to the stable lowindex surfaces of UO2 has been predicted using atomistic simulation techniques. WhileCe4+ andZr4+ substitutesimply for U4+, charge compensation in the form of oxygen vacancies is required in the case ofBa2+ and Sr2+. Three surfaces are considered: (110), (111) and (100). Although the (111) and (110) areperfect cleaved surfaces, the (100) necessarily incorporates a series of surface defects toneutralize the inherent dipole. The segregation energies of these cations depend strongly onthe surface to which the segregation is proceeding. Furthermore, it is also a function of theorientation of the segregating defect cluster with respect to the surface and, inthe case of the (100) dipolar surface, the configuration of the surface defects.

Thermal Evolution of Hydrogen Following Water Adsorption on Defective UO2(100)

The evolution of molecular hydrogen from UO2(100) surfaces exposed to water vapor has been studied using temperature programmed desorption. No hydrogen evolution is observed from well-annealed, pristine surfaces. Significant amounts of hydrogen, however, desorb from highly defected, ion-sputtered surfaces at ∼400 K. Annealing studies show that the most of the defects are healed after heating to 750 K. Successive water exposures also heal these defects. These results indicate that the defect is likely an oxygen vacancy (reduced uranium site). On annealing, oxygen diffuses from the near-surface region to fill the surface vacancy. The dissociative adsorption of water also fills the oxygen vacancy. As the damage resulting from ion sputtering is similar to that caused by the recoil atom formed during radioactive decay, the thermal chemistry at defects may be an indirect, but significant, factor in the radiation chemistry of adsorbed species on alpha-emitting materials.

Effects of colloidal and dissolved silica on the dissolution of UO2nuclear fuel in groundwater leaching tests

The influence of aqueous silica of two different physical forms (dissolved ions and SiO2 colloid) on the dissolution of UO2 nuclear fuel material was investigated at 95 °C temperature in autoclaves. It was tested that SiO2 colloids can contribute to the surface degradation or act as carrier for uranium ions during a near field geochemical dissolution process. In the presence of colloids, well-crystallized secondary phases containing U and Si were formed on the surfaces, the latter attacked by the treatment. This was not the case when dissolved Si was used. SiO2 colloids were partly found in their original form on the surfaces after 1000 hours at 95 °C. A surface charge model suggests that this different effects are due to the development of electrostatic interactions between the UO2 and SiO2 surfaces.

Reduction of uranium(VI) by mixed iron(II)/iron(III) hydroxide (green rust): formation of UO2 nanoparticles.

Green rusts, which are mixed ferrous/ferric hydroxides, are found in many suboxic environments and are believed to play a central role in the biogeochemistry of Fe. Analysis by U LIII-edge X-ray absorption near edge spectroscopy of aqueous green rust suspensions spiked with uranyl (U(VI)) showed that U(VI) was readily reduced to U(IV) by green rust The extended X-ray absorption fine structure (EXAFS) date for uranium reduced by green rust indicate the formation of a UO2 phase. A theoretical model based on the crystal structure of UO2 was generated by using FEFF7 and fitted to the data for the UO2 standard and the uranium in the green rust samples. The model fits indicate that the number of nearest-neighbor uranium atoms decreases from 12 for the UO2 structure to 5.4 forthe uranium-green rust sample. With an assumed four near-neighbor uranium atoms per uranium atom on the surface of UO2, the best-fit value for the average number of uranium atoms indicates UO2 particles with an average diameter of 1.7 +/- 0.6 nm. The formation of nanometer-scale particles of UO2, suggested by the modeling of the EXAFS data, was confirmed by high-resolution transmission electron microscopy, which showed discrete particles (approximately 2-9 nm in diameter) of crystalline UO2. Our results clearly indicate that U(VI) (as soluble uranyl ion) is readily reduced by green rust to U(IV) in the form of relatively insoluble UO2 nanoparticles, suggesting that the presence of green rusts in the subsurface may have significant effects on the mobility of uranium, particularly under iron-reducing conditions.

Measurement of Dissolution Rates of UO2 Matrix with the Isotope Dilution Method

ABSTRACTThe objective of this work was to measure the actual rates of matrix dissolution of UO2 in synthetic groundwater by using an isotope dilution method to provide a quantitative estimate of precipitation effects. A preliminary series of tests was performed under oxidising conditions, in contact with air. The second series of tests was under reducing conditions produced by actively corroding iron. The solid phases were fragments of unirradiated fuel pellets and intact pellets. The aqueous phase was a dilute, synthetic groundwater - so-called modified Allard water that is buffered by sodium bicarbonate/carbonate. This paper gives results for the experiments under oxidising conditions and preliminary tests under reducing conditions. In contact with air, the U concentrations reached higher levels than measured in previous experiments with spent fuel or with UO2 pellets. The comparison of the U concentrations calculated from isotopicratios with the experimental results suggests precipitation has begun at later stages of restarted tests. The measurements in the presence of actively corroding iron gave very low concentrations in the aqueous phase. At contact times longer than one week, all U seemed to disappear from solution and sorb or precipitate on UO2 or Fe surfaces in the test vessel.

Modelling Experimental Results on Radiolytic Processes at the Spent Fuel Water Interface. I. Radiolysis Products and U Release

ABSTRACTExperimental and modelling efforts in the last decade in the frame of nuclear waste management field have been focused on studying the role of the UO2 surfaces in poising the redox state of solid/water systems. For this purpose, an experimental programme was developed consisting on dissolution experiments with PWR spent fuel fragments in an anoxic environment and by using different solution compositions.The collected data so far, indicate that production and fate of radiolysis products follow the same trends independently on the solution composition used in the tests. Hydrogen and oxygen concentrations show an initial increase with time until reaching a constant concentration. The trend observed for hydrogen peroxide is a decrease at short contact times to reach again a constant concentration with time. These steady-states indicate an overall balance of the generated radiolytic species.Modelling work indicates that uranium dissolution is controlled by the oxidation of the spent fuel matrix in 10mM bicarbonate solutions while in the tests carried out at lower or without carbonate concentrations uranium in the aqueous phase is governed by the precipitation of schoepite.These results are determinant to highlight that reducing conditions are restored in the aqueous phase in relatively short periods of time and at short distances away from the dynamic redox spent fuel/water interface.

Raman microspectrometric identification of corrosion products formed on UO2 nuclear fuel during leaching experiments

The corrosion of directly disposed spent nuclear fuel by contact with intruding groundwater will alter the physical and chemical properties of this material. Secondary phases which formed during alteration of UO2 surfaces were measured with Raman microspectrometry and the characteristic vibrational spectra of the materials were recorded. U phases were synthesized in hydrothermal autoclave syntheses. A Raman spectral library of UO2 corrosion phases was set up for the identification of unknown products found on altered nuclear fuel samples. In a case study, U peroxide (UO4) was identified by comparison with a natural sample as the main alteration phase by its characteristic O–O Raman vibration at 870 cm−1. The results demonstrate the differentiation between UO2 and its alteration products U(VI) oxyhydroxide and U(VI) peroxide (UO4) on one sample with a relatively quick, non-destructive, spatially resolving measurement method which delivers oxidation state and molecular bonding information. Implications for the analysis of complex heterogenous matrices are discussed.

On the Interaction between uranyl carbonate and UO2(s) in anaerobic solution

It was repeatedly observed that in nearly neutral anoxic solutions (10 mM NaCl, 2 mM NaHCO3-, pH ~ 8.5) U(VI) concentration decreases from 10-5 M to 10-7 M after 24 days interaction with UO2(s) surfaces. The formation of a layer of hyper-stoichiometric uranium oxide UO2+x(s) on fresh UO2(s) surfaces is proposed as a possible reaction mechanism. Gibbs free energy calculations and other batch experiments were carried out to test the validity of this hypothesis. The calculated negative AG values indicate that the formation of a layer of UO2+x (x=0.25 or 033) is thermodynamically possible. The U(VI)/U(IV) ratio of the reacted uranium oxide surface is 0375, corresponding to UO2.27(s). The average U(VI)/U(IV) ratio obtained for the suspended particles worn off from the reacted UG2(s) surface on an ultrasonic bath resulted 0.162. Our preliminary results from this work support the above hypothesis, which might be also an explanation for some of the results obtained during spent fuel leaching experiments under anaerobic conditions.

Corrosion and dissolution studies of UO2 containing α-emitters

Summary Corrosion and dissolution of UO2 doped with short-lived α-emitters, the so called α-doped UO2, was investigated in leaching experiments combined with X-ray Photoelectron Spectroscopy (XPS) analyses. These experiments are part of an ongoing set of tests aiming at evaluating the effects of α-radiolysis on the dissolution of the UO2 spent fuel matrix when it will become exposed to groundwater in a deep final repository. Pellets of UO2 containing ∼10 and ∼0.1 wt.% 238Pu were used. These compositions simulate the α-radiation field of different types of commercial LWR spent fuel after different storage times. Undoped UO2 pellets were also used as reference. Static sequential leaching experiments in demineralised water at room temperature under anoxic conditions confirmed that higher fractions of uranium are released in the case of α-doped materials compared with undoped UO2. Relatively large amounts of plutonium were found in the leachate of the materials doped with 238Pu. Significant contributions to the total amount released of uranium were found in the rinse solutions of the leaching vessels. The evolution of the U(VI)/U(IV) ratio on the leached surfaces as a function of contact time was studied by deconvolution of the 4f7/2 uranium peak obtained from the XPS spectra. In the case of undoped UO2, after ∼1000 h of leaching the increase of the U(VI)/U(IV) ratio stabilized to a level corresponding to ∼UO2.33. The XPS results for both α-doped materials showed a continuing increase of the U(VI) fraction on the surface with increasing leaching times after the first ∼10 h. The rate of increase depended on the α-activity of the fuel sample, hence on the amount of radiolysis in the water. After ∼1000 h of leaching, essentially only U(VI) was detected on the surface of the high activity α-doped UO2. The presence of a yellowish layer of reprecipitated U(VI) phases covering the surface could be observed optically already after ∼100 h of leaching.

Carbon Efficiency and the Surface Chemistry of the Actinides: Direct Formation of Furan from Acetylene over β-UO3

This work presents the direct oxidative coupling of two molecules of acetylene to furan (C4H4O) over the surface of pure polycrystalline β-UO3. For comparison, only traces of furan are formed over α-U3O8 and none on UO2 surfaces. A comparison to the reactions of other C2 compounds (ethanol (H. Madhavaram and H. Idriss, J. Catal.184, 553 (1999)), acetaldehyde (H. Madhavaram and H. Idriss, Catal. Today63, 309 (2000)), and ethylene (H. Madhavaram and H. Idriss, J. Catal.184, 553 (1999); H. Madhavaram and H. Idriss, Stud. Surf. Sci. Catal.110, 265 (1997)) over β-UO3 indicates the presence of two routes for making furan from C2 compounds: one via aldolization and the other via oxidative coupling.

Radiolytic Modelling: Application To Spent Fuel Dissolution Experiments

ABSTRACTIn this work we describe our first efforts to simulate the chemical evolution of the spent fuel/water system in dissolution experiments using a radiolytic model. These experiments were carried out with spent fuel fragments in deionized water [1]. In order to build the radiolytic model, a set of reactions were selected, together with their kinetic constants, including the catalytic decomposition of H2O2 in the UO2 surface and a reaction mechanism for UO2 oxidation and dissolution. Averaged alpha and beta G-values were used, neglecting gamma radiation. Diffusion to the gas phase was treated thermodynamically assuming Henry's law. The computer code used in the simulations was CHEMSIMUL [9]. First attempts to simulate the experimental results were not successful and a sensitivity analysis was carried out. As a result of this, the kinetic constant of the reaction OH• + H2 → H• + H2O, the only reaction that consumes hydrogen in the system, was identified as a key parameter, together with the kinetic constants of the UO2 oxidation and dissolution reactions. Varying these parameters we were able to reproduce the experimental data. Although more simulations are needed, these results are encouraging because they give us more confidence in the use of radiolytic models as useful tools to predict the long-term alteration rate of the spent fuel under repository conditions, which is our ultimate goal.

The Surface Precipitation During UO2 Leaching Process

To understand the influence of cations and silica in groundwater on spent fuel corrosion, the leaching behavior of newly reduced UO2.00 in different solutions was investigated. Uranium concentrations in the solutions were measured and the leached UO2 surface and precipitates on it were analyzed by XPS, laser Raman spectroscopy, SEM-EDS and TEM-EDS. It was found that, in air saturated 2 mM HCO3- solutions, the leaching rates of UO2(s) depend on the composition of the leaching solutions. Very similar leaching rates, 0.5 μg • cm-2 • day-1, were obtained in 0.1 M NaCl and 0.1 M KCl solutions, while that in 0.45mM Ca2+ -0.18 mM Mg2+ -0.2 mM SiO2 containing synthetic groundwater was 10 times smaller. Calcite and quartz like precipitates were detected on the corroded UO2(s) surface.

Redox Reactions of Iron and Uranium Dioxide in Simulated Cement Pore Water Under Anoxic Conditions

Results on the chemical behavior of Fe(0) and UO2(s), as well as the interaction between fresh and corroded iron with U(VI) in simulated cement contacting alkaline solution are reported. Batch experiments were conducted under anoxic conditions at different alkalinities and salt concentrations to investigate: (a) the corrosion of iron foils (b) the U(VI) removal by fresh, aged and pre-treated (with FeS or Fe3O4 layers) iron surfaces in a simulated cement pore fluid, (c) the dissolution rates of newly reduced UO2.00 slices in simulated cement pore fluid and KOH solutions (pH 12.7) and (d) the isotope exchange reactions between dissolved 235U(VI) and 238UO2(s). The reacted iron and UO2(s) surfaces were analyzed by X-ray diffraction (XRD), scanning electron microscopy-energy dispersive spectra (SEM-EDS), laser Raman spectroscopy and X-ray photoelectron, spectroscopy (XPS).

Reactions of acetic acid on UO2(111) single crystal surfaces

The reactions of acetic acid have been investigated on the (111) surface of uranium dioxide single crystals, which was characterized by low energy electron diffraction and Auger electron spectroscopy. Temperature programmed desorption (TPD) of this molecule displays a rich chemistry on both the stoichiometric and electron beam sputtered surfaces. Acetic acid-TPD on a stoichiometric surface yields ketene (dehydration) as the main product, plus acetaldehyde as the minor product. On an electron beam sputtered surface, two additional association products were observed—butene and crotonaldehyde. The carbon yield of ketene on the stoichiometric surface was comparable to that of CeO2(111) and TiO2(001), higher than that of ZnO(0001) and Cu2O(100), but lower than that of MgO(100). This investigation has shown the ability of UO2(111) surfaces to dehydrate, reduce, and reductively couple organic molecules.

Analysis and Modeling of Decontamination Experiments of Depleted Uranium Dioxide in RF Plasma

ABSTRACT: Analysis was performed of the results of UO2 etching experiments, which used NF3 RF glow discharge. In the experiments, the gas pressure was varied from ∼10 ‐ 40 Pa and the absorbed power was varied from 24 ‐ 210 W. The UO2 etch reactions to form volatile UF6 were strongly dependent on the absorbed power and, to a lesser extent, on pressure. At and below 50 W absorbed power, UO2 etching increased with pressure up to ∼ 23 Pa, then decreased with further increase in pressure. At higher power, however, UO2 etching increased monotonically with pressure. The UO2 etching process was self‐limiting, due to the formation of non‐volatile products UF2–5, UO2F, and UO2F2. The accumulation of UF6 and O2 in the plasma sheath also reduced the diffusion coefficient of atomic fluorine radicals to the UO2 surface thereby contributing to the decrease in etch rate with immersion time.