Dissolution Of UO2 Research Articles

γ-radiation induced corrosion of copper in bentonite-water systems under anaerobic conditions

In this work we have experimentally studied the impact of bentonite clay on the process of radiation-induced copper corrosion in anoxic water. The motivation for this is to further develop our understanding of radiation-driven processes occurring in deep geological repositories for spent nuclear fuel where copper canisters containing the spent nuclear fuel will be embedded in compacted bentonite. Experiments on radiation-induced corrosion in the presence and absence of bentonite were performed along with experiments elucidating the impact irradiation on the Cu2+ adsorption capacity of bentonite. The experiments presented in this work show that the presence of bentonite clay has no or very little effect on the magnitude of radiation-induced corrosion of copper in anoxic aqueous systems. The absence of a protective effect similar to that observed for radiation-induced dissolution of UO2 is attributed to differences in the corrosion mechanism. This provides further support for the previously proposed mechanism where the hydroxyl radical is the key radiolytic oxidant responsible for the corrosion of copper. The radiation effect on the bentonite sorption capacity of Cu2+ (reduced capacity) is in line with what has previously been reported for other cations. The reduced cation sorption capacity is partly attributed to a loss of Al-OH sites upon irradiation.

Kinetics of dissolution of UO2 in nitric acid solutions: A multiparametric study of the non-catalysed reaction

UO2 pellets were prepared by densification of oxides obtained from the conversion of the oxalate precursor. Then characterized in order to perform a multiparametric study of the dissolution in nitric acid medium. In this frame, for each sample, the densification rate, the grain size and the specific surface area of the prepared pellets were determined prior to the final dissolution experiments. By varying the concentration of the nitric acid solution and temperature, three different and successive steps were identified during the dissolution. Under the less aggressive conditions considered, a first transient step corresponding to the dissolution of the most reactive phases was observed at the solid/solution interface. Then, for all the tested conditions, a steady state step was established during which the normalised dissolution rate was found to be constant. It was followed by a third step characterized by a strong and continuous increase of the normalised dissolution rate. The duration of the steady state, also called “induction period”, was found to vary drastically as a function of the HNO3 concentration and temperature. However, independently of the conditions, this steady state step stopped at almost similar dissolved material weight loss and dissolved uranium concentration. During the induction period, no important evolution of the topology of the solid/liquid interface was evidenced authorizing the use of the starting reactive specific surface area to evaluate the normalised dissolution rates thus the chemical durability of the sintered pellets. From the multiparametric study of UO2 dissolution proposed, oxidation of U(IV) to U(VI) by nitrate ions at the solid/liquid interface constitutes the limiting step in the overall dissolution mechanism associated to this induction period.

Effect of metallic iron on the oxidative dissolution of UO2 doped with a radioactive alpha emitter in synthetic Callovian-Oxfordian groundwater

In the hypothesis of direct disposal of spent fuel in a geological nuclear waste repository, interactions between the fuel mainly composed of UO2 and its environment must be understood. The dissolution rate of the UO2 matrix, which depends on the redox conditions on the fuel surface, will have a major impact on the release of radionuclides into the environment. The reducing conditions expected for a geological disposal situation would appear to be favorable as regards the solubility and stability of the UO2 matrix, but may be disturbed on the surface of irradiated fuel. In particular, the local redox conditions will result from a competition between the radiolysis effects of water under alpha irradiation (simultaneously producing oxidizing species like H2O2, hydrogen peroxide, and reducing species like H2, hydrogen) and those of redox active species from the environment. In particular, Fe2+, a strongly reducing aqueous species coming from the corrosion of the iron canister or from the host rock, could influence the dissolution of the fuel matrix. The effect of iron on the oxidative dissolution of UO2 was thus investigated under the conditions of the French disposal site, a Callovian-Oxfordian clay formation chosen by the French National Radioactive Waste Management Agency (Andra), here tested under alpha irradiation. For this study, UO2 fuel pellets doped with a radioactive alpha emitter (238/239Pu) were leached in synthetic Callovian-Oxfordian groundwater (representative of the French waste disposal site groundwater) in the presence of a metallic iron foil to simulate the steel canister. The pellets had varying levels of alpha activity, in order to modulate the concentrations of species produced by water radiolysis on the surface and to simulate the activity of aged spent fuel after 50 and 10,000years of alpha radioactivity decay. The experimental data showed that whatever the sample alpha radioactivity, the presence of iron inhibits the oxidizing dissolution of UO2 and leads to low uranium concentrations (between 4×10−10 and 4×10−9M), through a reactional mechanism located in the very first microns of the UO2/water reactional interface. The mechanism involves consumption of oxidizing species, in particular of H2O2 by Fe2+ at the precise place where these species are produced, and is accompanied by the precipitation of an akaganeite-type Fe3+ hydroxide on the surface. The higher the radioactivity of the samples, the greater the precipitation induced. Modeling has been developed, coupling chemistry with transport and based on the main reactional mechanisms identified, which enables accurate reproduction of the mineralogy of the system under study, giving the nature of the phases under observation as well as the location of their precipitation. Obviously without excluding a potential contribution from the hydrogen produced by the anoxic corrosion of the iron foil, this study has shown that iron plays a major role in this oxidizing dissolution inhibition process for the system investigated (localized alpha radiolysis). This inhibitor effect associated with iron is therefore strongly dependent on the location of the redox front, which is found on the surface in the case of alpha irradiation UO2/water reactional interface.

Physicochemical principles of preparation of U(VI) carbonate solutions for extraction reprocessing in the Carbex process

Physicochemical principles of preparation of U(VI) carbonate solutions in the step of oxidative dissolution of U3O8 and UO2 in the Carbex process are considered. Carbonate solutions with the U(VI) concentration higher than 100 g L–1, suitable for subsequent final purification of uranium by extraction, can be prepared under the conditions of formation of U(VI) carbonate–peroxide complexes in the course of dissolution with prevention of hydrolysis of U(VI) compounds. The behavior of impurities simulating some fission products in the course of oxidative dissolution was studied, and the decontamination factors of U(VI) from the chosen simulated fission products were determined.

Numerical analysis of the dissolution of uranium dioxide by molten zircaloy using MPS method

Dissolution of uranium dioxide by molten zircaloy at high temperatures is an important chemical process affecting the fuel melting during nuclear reactor severe accidents. In this study, the dissolution model, based on the mass transfer model and the dissolution criteria, was developed and incorporated into the Moving Particle Semi-implicit (MPS) method. To verify the dissolution model, the one-dimensional mass transfer between two semi-infinite materials was simulated by the improved MPS code. The sensitivity analyses of the particle size have been conducted since particle size determines the simulation accuracy. Then analyses on dissolution of UO2 by molten zircaloy were performed according to results by the improved MPS code. In this study, two series of experiments on dissolution of UO2 by molten zircaloy conducted by Kim and Olander were simulated respectively. The first series of experiments focuses on mass transfer. The corresponding simulation results of MPS show that the dissolution kinetics of UO2 by molten zircaloy follows the parabolic law, and the penetration depths of chemical elements (U, O, Zr) were limited by time. The second series of experiments concerns on natural convection. The simulation results show that the natural convection of melting process has a significant influence on the dissolution behavior. The results indicate that the dissolution of UO2 by molten zircaloy fits the parabolic law before the concentration of UO2 in the liquid reaches the saturation point. After the concentration of UO2 reaches the saturation point, the dissolution of UO2 by molten zircaloy also obeys the parabolic law. The difference between the two parabolic laws is the dissolution rate. The dissolution rate decreases after the UO2 is saturated in the liquid. For natural convection, the penetration depths of chemical elements (U and O) become deeper, which could accelerate the progress of dissolution.

Oxidative dissolution of ADOPT compared to standard UO2 fuel

In this work we have studied oxidative dissolution of pure UO2 and ADOPT (UO2 doped with Al and Cr) pellets using H2O2 and gammaradiolysis to induce the process. There is a small but significant difference in the oxidative dissolution rate of UO2 and ADOPT pellets, respectively. However, the difference in oxidative dissolution yield is insignificant. Leaching experiments were also performed on in-reactor irradiated ADOPT and UO2 pellets under oxidizing conditions. The results indicate that the U(VI) release is slightly slower from the ADOPT pellet compared to the UO2. This could be attributed to differences in exposed surface area. However, fission products with low UO2 solubility display a higher relative release from ADOPT fuel compared to standard UO2-fuel. This is attributed to a lower matrix solubility imposed by the dopants in ADOPT fuel. The release of Cs is higher from UO2 which is attributed to the larger grain size of ADOPT.

Uranium dioxide in Fe(III)-containing ionic liquids with DMSO: Dissolution, separation, and structural characterization

UO2 can be successfully dissolved in imidazolium-based Fe(III)-containing ionic liquids (ILs) with the help of DMSO. Spectroscopic studies and X-ray diffraction show that UO2Cl42− is the principal product. The dissolved uranyl species can be easily separated from the ILs via a combination of crystallization and solvent extraction. Moreover, even if 15.2 wt% of the rare-earth elements of Sm, Eu, and Gd, compared with the total amount of uranium and the rare-earth elements, exist in the IL, only uranium-containing crystals would be selectively formed and separated from the system. The solvents of acetone and acetonitrile could be used to separate the rare-earth elements from uranium in the IL with the help of imidazolium chloride. Considering the complete process from the dissolution of UO2 and some rare-earth oxides to the separation of uranium and rare-earth elements in the IL, the facile approach is promising for the spent nuclear fuel reprocessing.

Can H2 enhance the oxidative dissolution of UO2?

Can H2 enhance the oxidative dissolution of UO2?

The effect of fuel chemistry on UO2 dissolution

The dissolution rate of both unirradiated UO2 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 UO2 under oxidizing repository conditions and compare them to the rates predicted by current dissolution models.Both unirradiated UO2 and UO2 doped with varying concentrations of Gd2O3, to simulate used fuel composition after long time periods when radiolysis has minor contributions to dissolution, were examined. In general, a rise in temperature increased the dissolution rate of UO2 and had a larger effect on pure UO2 than on those doped with Gd2O3. Oxygen dependence was observed in the UO2 samples with no dopant and increased as the temperature rose; in the doped fuels less dependence was observed. The addition of gadolinia into the UO2 matrix resulted in a significant decrease in the dissolution rate. The matrix stabilization effect resulting from the dopant proved even more beneficial in lowering the dissolution rate at higher temperatures and dissolved O2 concentrations in the leachate where the rates would typically be elevated.

Role of Microstructure and Surface Defects on the Dissolution Kinetics of CeO2, a UO2 Fuel Analogue.

The release of radionuclides from spent fuel in a geological disposal facility is controlled by the surface mediated dissolution of UO2 in groundwater. In this study we investigate the influence of reactive surface sites on the dissolution of a synthesized CeO2 analogue for UO2 fuel. Dissolution was performed on the following: CeO2 annealed at high temperature, which eliminated intrinsic surface defects (point defects and dislocations); CeO2-x annealed in inert and reducing atmospheres to induce oxygen vacancy defects and on crushed CeO2 particles of different size fractions. BET surface area measurements were used as an indicator of reactive surface site concentration. Cerium stoichiometry, determined using X-ray Photoelectron Spectroscopy (XPS) and supported by X-ray Diffraction (XRD) analysis, was used to determine oxygen vacancy concentration. Upon dissolution in nitric acid medium at 90 °C, a quantifiable relationship was established between the concentration of high energy surface sites and CeO2 dissolution rate; the greater the proportion of intrinsic defects and oxygen vacancies, the higher the dissolution rate. Dissolution of oxygen vacancy-containing CeO2-x gave rise to rates that were an order of magnitude greater than for CeO2 with fewer oxygen vacancies. While enhanced solubility of Ce(3+) influenced the dissolution, it was shown that replacement of vacancy sites by oxygen significantly affected the dissolution mechanism due to changes in the lattice volume and strain upon dissolution and concurrent grain boundary decohesion. These results highlight the significant influence of defect sites and grain boundaries on the dissolution kinetics of UO2 fuel analogues and reduce uncertainty in the long term performance of spent fuel in geological disposal.

Irradiation Defects in UO2 Leached in Oxidizing Water: An in-situ Raman Study

In this study, the behavior of irradiation induced defects in UO2 in a specific oxidizing water environment was investigated under alpha irradiation. Raman spectroscopy was used to measure the formation kinetics of irradiation defects in UO2 when exposed to an aerated medium. The acquired data were then compared with a reference UO2/(Ar+H2) system. The results revealed that the presence of either aerated deionized water or reducing environment modifies the formation kinetics of irradiation defects. In addition, the precipitation of the studtite phase is also observed during the leaching experiment and water chemical analysis showed that the dissolution of UO2 proceeds without the formation of an oxidized UO2+x.layer.

Effect of the Microstructural Morphology on UO2 Powders

Several UO2 powders with different morphologies were synthetized and characterized. Three different morphologies were synthesized thanks to sol gel process (big heap of about 200μm wide consisting of sintered crystallites) on the one hand, and to oxalic precipitations (one square platelet morphology and one hexagonal stick morphology) on the other hand. Significant differences in dissolution kinetics were observed. Therefore, the morphology of the powders was found to be a key parameter that has to be considered in studies of UO2 dissolution kinetics. The second part of the study consists in dissolving in nitric acid in in the same operating conditions three UO2 powders having different crystallites sizes. It was shown that dissolution kinetics is dependent on the morphology but also on the powder stoichiometry.

Kinetics of Alkaline Leaching of UO2 and FeS2 in Co-existing System

The dissolution of uranium dioxide under oxidative alkaline conditions is critically influenced by iron pyrite (FeS2), which is a gangue mineral commonly found in uranium ores. This paper makes an effort in understanding the kinetics of dissolution of UO2 in a co-existing system of UO2 and FeS2 under the lixiviant combination of Na2CO3 and NaHCO3–O2 at elevated temperature and pressure. Dissolution experiments were carried out in a laboratory batch autoclave reactor using synthetic mixtures of minerals consisting of UO2, FeS2 (reactive gangue—varied from 1 to 6 %), silica (inert gangue) and calcite (inert gangue). The kinetic profiles indicated that the rate of dissolution of UO2 increased with initial increase in FeS2 content in the feed and decreased when the FeS2 weight increased beyond 3 %. The dissolution phenomenon was analysed using scanning electron microscopy and X-ray diffraction studies.

Dissolution kinetics of Indian PHWR natural UO2 fuel pellets in nitric acid – Effect of initial acidity and temperature

Understanding the dissolution behaviour of plutonium rich MOX fuel is one of the major challenges in fast reactor spent fuel reprocessing. As an initial step, kinetics for the dissolution of Indian PHWR natural UO2 sintered pellets in nitric acid was studied experimentally. In this paper we have reported the effect of initial nitric acid concentration and temperature on dissolution rate of sintered UO2 pellets. The shrinking core model was used to correlate the experimental results. The apparent activation energy estimated from the temperature effect of the chemical reaction was found to be about 90.5 kJ/mol. The estimated apparent activation energy indicates that the dissolution rate of UO2 pellets was controlled by chemical reaction under the experimental conditions.

Dissolution of oxide nuclear fuel in subacidic iron(III) nitrate solutions and extraction of uranium from it with tributyl phosphate-containing supercritical carbon dioxide

The solubility of the uranyl nitrate complex with tributyl phosphate (TBP) in liquid and supercritical carbon dioxide (SC-CO2) has been measured as a function of temperature and pressure. This complex is formed upon the dissolution of uranium dioxide in the two-phase system consisting of aqueous iron(III) nitrate and TBP-saturated liquid or supercritical CO2. The kinetics of the dissolution of ceramic UO2 in such a two-phase solvent is studied. It has been shown that uranium can be extracted in the presence of TBP-containing supercritical CO2 via the dissolution of spent oxide nuclear fuel in subacidic iron(III) nitrate solutions with the simultaneous extraction of the uranium complex into the TBP-SC-CO2 phase.

Surface passivation limited UO2 oxidative dissolution in the presence of FeS.

Iron sulfide minerals produced during in situ bioremediation of U can serve as an oxygen scavenger to retard uraninite (UO2) oxidation upon oxygen intrusion. Under persistent oxygen supply, however, iron sulfides become oxidized and depleted, giving rise to elevated dissolved oxygen (DO) levels and remobilization of U(IV). The present study investigated the mechanism that regulates UO2 oxidative dissolution rate in a flow-through system when oxygen breakthrough occurred as a function of mackinawite (FeS) and carbonate concentrations. The formation and evolution of surface layers on UO2 were characterized using XAS and XPS. During FeS inhibition period, the continuous supply of carbonate and calcium in the influent effectively complexed and removed oxidized U(VI) to preserve an intermediate U4O9 surface. When the FeS became depleted by oxidization, a transient, rapid dissolution of UO2 was observed along with DO breakthrough in the reactor. This rate was greater than during the preceding FeS inhibition period and control experiments in the absence of FeS. With increasing DO, the rate slowed and the rate-limiting step shifted from surface oxidation to U(VI) detachment as U(VI) passivation layers developed. In contrast, increasing the carbonate concentrations facilitated detachment of surface-associated U(VI) complexes and impeded the formation of U(VI) passivation layer. This study demonstrates the critical role of U(VI) surface layer formation versus U(VI) detachment in controlling UO2 oxidative dissolution rate during periods of variable oxygen presence under simulated groundwater conditions.

Influence of the alpha radiation on the UO2 dissolution in high pH cementitious waters

To assess the long-term behaviour of spent fuel in conditions of the Belgian Supercontainer design, dissolution tests were performed to determine the solubility and dissolution rate of depleted and Pu-doped UO2 in synthetic cement waters. The results confirm the behaviour generally observed with UO2 at neutral pH in absence of H2. Due to the radiolysis of water, the produced oxidizing species oxidise the U(IV) to U(VI), inducing an increase of the uranium concentration. Consequently, the dissolution rate increases from the depleted UO2 to the most active doped UO2. Imposing a H2 overpressure had no significant effect on the UO2 behaviour.

Effects of Mn(II) on UO2 dissolution under anoxic and oxic conditions.

Groundwater composition and coupled redox cycles can affect the long-term stability of U(IV) products from bioremediation. The effects of Mn(II), a redox active cation present at uranium-contaminated sites, on UO2 dissolution in both oxic and anoxic systems were investigated using batch and continuous-flow reactors. Under anoxic conditions Mn(II) inhibited UO2 dissolution, which was probably due to adsorption of Mn(II) and precipitation of MnCO3 that decreased exposure of U(IV) surface sites to oxidants. In contrast, Mn(II) promoted UO2 dissolution under oxic conditions through Mn redox cycling. Oxidation of Mn(II) by O2 produced reactive Mn species, possibly short-lived Mn(III) in solution or at the surface, that oxidatively dissolved the UO2 more rapidly than could the O2 alone. At pH 8 the Mn cycling was such that there was no measurable accumulation of particulate Mn oxides. At pH 9 Mn oxides could be produced and accumulate, while they were continuously reduced by UO2, with Mn(II) returning to the aqueous phase. With the rapid turnover of Mn in the redox cycle, concentrations of Mn as low as 10 μM could maintain an enhanced UO2 dissolution rate. The presence of the siderophore desferrioxamine B (a strong Mn(III)-complexing ligand) effectively decoupled the redox interactions of uranium and manganese to suppress the promotional effect of Mn(II).

Factors governing the efficiency of dissolution of UO2 ceramic pellets in aqueous solutions of iron nitrate

Dissolution of ceramic UO2 in aqueous Fe(NO3)3 solutions at different temperatures under the conditions of limited contact with air and in the autoclave mode was studied. In the course of UO2 dissolution at 60–90°C, the U/Fe molar ratio appears to be ∼1, whereas at room temperature (25°C) this value is ∼0.5. By varying the acidity of Fe nitrate solutions at these temperatures, it is possible to increase the U/Fe molar ratio to ∼4 and to obtain uranyl nitrate solutions with simultaneous removal of Fe from the solution in the form of a precipitate of the basic salt, or to perform quantitative dissolution of UO2 under the conditions excluding the formation of such precipitate. In the course of dissolution of ceramic UO2 in Fe(NO3)3 solutions, the appearance or absence of Fe(II) ions, the formation or absence of the precipitate of the Fe basic salt, and variation of solution pH are interrelated and are determined by the process temperature.

Supercritical fluid extraction of uranium and fission products in reprocessing of simulated spent nuclear fuel in weakly acidic solutions of Fe(III) nitrate in the presence of tributyl phosphate

The kinetics of dissolution of ceramic UO2 in aqueous solutions of Fe(III) nitrate in the presence of TBP in supercritical CO2 was studied. Quantitative recovery of U into solutions of Fe nitrate in combination with its extraction from the solution into the fluid occurs under these conditions within ∼2 h, which is by more than an order of magnitude faster than in dissolution in a Fe(III) nitrate solution without TBP under common conditions. The behavior of uranium and simulated fission products (FPs) in reprocessing of simulated spent nuclear fuel (SSNF) in weakly acidic Fe(III) nitrate solutions using supercritical CO2 containing TBP was studied. Under these conditions, SSNF dissolves quantitatively with the simultaneous recovery of the U nitrate complex into the TBP-containing fluid phase. Uranium is recovered from the fluid phase by back extraction with an H2O2 solution with simultaneous precipitation of U in the form of the peroxide UO4·2H2O. In the process, high degree of uranium decontamination from all the FPs is reached, which allows repeated use of U in the nuclear fuel cycle.