Dissolution Of Uranium Research Articles

Hydrogeological controls of arsenic and uranium dissolution into groundwater of the Pine Ridge Reservation, South Dakota

AbstractThis study integrates geochemical modeling, spatial analysis and several statistical methods including principal component analysis, multivariate regression and cluster analysis to investigate hydrogeologic controls of arsenic and uranium contamination within groundwater of the Arikaree aquifer on the Pine Ridge Reservation (PRR). Located in southwestern South Dakota, the PRR is largely rural and many people rely on domestic supply wells completed in the Arikaree aquifer as their primary drinking water source. Locally, the White River Group, which unconformably underlies the Arikaree Group, is enriched in arsenic and uranium related to volcanic ash deposits and acts as a geogenic metal source. Geochemical data from over 250 groundwater samples were obtained through collaboration with the Oglala Sioux Tribe. Cluster spatial statistics analyses delineated four regions of statistically significant variations in groundwater chemistry that represent upgradient, intermediate, and downgradient portions of the Arikaree aquifer. Groundwater evolves as it flows through the Arikaree aquifer with increasing alkalinity, sodium, and pH along flow paths. These chemical changes are likely due to dissolution of carbonate minerals and volcanic ash. Thermodynamic calculations suggest increasing supersaturation of the groundwater with respect to calcite; thus, volcanic ash dissolution may be an important secondary source of alkalinity. Elevated alkalinity and pH levels were found to be the driving factors of arsenic and uranium mobility, and downgradient sections of the aquifer in the northern portions of the PRR are most likely to be impacted by metal(loid) contamination, with 73% of the wells in this grouping failing a USEPA maximum contaminant level for arsenic, uranium, and/or gross alpha.

γ-Radiation and H2O2 induced oxidative dissolution of uranium(iv) oxide in aqueous solution containing phthalic acid.

This study aims to reveal possible involvements of organic acids in the oxidative dissolution of UO2. Using phthalic acid as a model compound, we have measured adsorption on UO2 and investigated effects on the reaction between H2O2 and UO2 and on oxidative dissolution induced by γ-irradiation. Significant adsorption of phthalic acid was observed even at neutral pH. However, the reaction between H2O2 and UO2 in phthalic acid solution induced oxidative dissolution of U(vi) similar to in aqueous bicarbonate solution. Moreover, degradation products of phthalic acid were not detected after the reaction of H2O2. These results indicate that even though phthalic acid adsorbs on the UO2 surface, it is not involved in the interfacial reaction of H2O2. In contrast, the dissolution of U by irradiation was inhibited in aqueous phthalic acid solution, whereas H2O2 generated by radiolysis was consumed by UO2. Based on these contrasting results, possible roles of radical species generated by water radiolysis were discussed.

Field experiments of surface water to groundwater recharge to characterize the mobility of uranium and vanadium at a former mill tailing site

Characterizing the mobility of uranium and vanadium in groundwater with a hydraulic connection to surface water is important to inform the best management practices of former mill tailing sites. In this study, the recharge of river water to the unsaturated and saturated zones of a uranium-contaminated alluvial aquifer was simulated in a series of forced-gradient single- and multi-well injection-extraction tests. The injection fluid (river water) was traced with natural and artificial tracers that included halides, fluorobenzoates, lithium, and naphthalene sulfonate to characterize the potential mass transport mechanisms of uranium and vanadium. The extraction fluid (river water/groundwater mixture) was analyzed for the tracers, uranium, and vanadium. The results from the tracers indicated that matrix diffusion was likely negligible over the spatiotemporal scales of the tests as evident by nearly identical breakthrough curves of the halides and fluorobenzoates. In contrast, the breakthrough curves of lithium and naphthalene sulfonate indicated that sorption by cation exchange and sorption to organic matter, respectively, were potential mass transport mechanisms of uranium and vanadium. Uranium was mobilized in the saturated zone containing gypsum (gypsum-rich zone), the vadose zone (vadose-rich zone), and the saturated zone containing organic carbon (organic-rich zone) whereas vanadium was mobilized only in the saturated gypsum-rich zone. The mechanisms responsible for the mobilization of uranium and vanadium were likely dissolution of uranium- and vanadium-bearing minerals and/or desorption from the gypsum-rich zone, flushing of uranium from the vadose-rich zone, and desorption of uranium from the organic-rich zone due to the natural contrast in the geochemistry between the river water and groundwater. The experimental design of this study was unique in that it employed the use of multiple natural and artificial tracers coupled with a direct injection of native river water to groundwater. These results demonstrated that natural recharge and flooding events at former mill tailing sites can mobilize uranium, and possibly vanadium, and contribute to persistent levels of groundwater contamination.

Uranium recovery from sandstone-type uranium deposit by acid in-situ leaching - an example from the Kujieertai

The factors influencing uranium recovery in water-rock systems during acid in-situ leaching (ISL) were studied at the Kujieertai uranium deposit in Xinjiang. Using an ISL unit, a field leach trial (FLT) had been carried out to test the sequential effects of a leaching solution without oxidant (H2SO4 solution 4–8 g/L) and a leaching solution with oxidant (H2SO4 3–7 g/L, and Fe (III) 2–6 g/L). The observation of the leaching process revealed clearly defined stages of uranium release from the solid mineral to solution. Uranium mobilization from solid mineral into solution can be described in four stages. At the beginning of the acid ISL process, there was no oxidant to be added to the leaching solution and the desorption of hexavalent uranyl ions in the open pores, as well as dissolution of hexavalent uranium minerals, led to a short-term peak in the pregnant solution, which happened while pH decreased from about 5.3 to 2.62. Following the depletion of the adsorbed hexavalent uranium and a decline in uranium dissolution intensity, the addition of Fe(III) facilitated the oxidation of tetravalent uranium, which enabled intensive uranium mobilization again. During this process, the dissolution of uranium had a strong positive correlation with the reduction of Fe(III) and Eh in the leach solution. Beside hydrochemical factors, the deportment of uranium was also an important factor affecting uranium recovery. Uranium located in the open pores can be completely exposed to the solution and the mobilization intensity was significantly affected by hydrogeochemical conditions; but the uranium present in microfissures and in the ore matrix could not be fully exposed to the solution, so, their dissolution intensity was primarily controlled by corrosion and permeability of the ore. In general, the hydrogeochemical conditions and the deportment of uranium were the external and internal factors that significantly affected the dissolution and recovery of uranium in the early and middle stages of the FLT. However, in the latest stages, due to uranium depletion, enhancing the chemical potential of the leaching solution, specifically acidity and/or the amount of oxidant, had little improvement on uranium recovery.

Leaching of low permeable sandstone uranium ore using auxiliary materials: anionic surfactants

A low-grade and permeable sandstone uranium samples in Inner Mongolia prospect, China was subjected to Na2CO3/NaHCO3 leaching. The uranium dissolution mechanism was investigated by effects of leaching parameters and the shrinking core model was selected to model uranium leaching reactions. Agitation leach tests showed that the maximum uranium leaching rate reached 96.5% at 25 °C and had an 19% yield increase when added of 1 mg/L the anionic surfactant to the leaching solution. The optimum process operating parameters were: molar ratio of Na2CO3:NaHCO3 1:1, solid–liquid ratio 1:6, contact time 4 h, agitation speed 400 rpm and the surfactant concentration in leach liquor with 1 mg/L. The kinetics equations indicated that the reactions appear to mix control turn to layer diffusion control with the surfactants adding.

Microstructural evolution of UO2 pellets containing metallic particles of Ru, Rh and Pd during dissolution in nitric acid solution: 3D-ESEM monitoring

Uranium dioxide containing 3 mol% platinum group metals (PGMs) (Ru, Rh, Pd) was synthesized by hydroxide precipitation. The powders were converted to oxides, pelletized and sintered to prepare dense pellets of UO2 incorporating PGMs particles. The characterization techniques performed revealed a microstructure similar to that of spent nuclear fuel (SNF). Dissolution tests in nitric acid demonstrated that in the presence of PGMs, the uranium dissolution rate was increased and the induction period was shortened. We used a new method based on the acquisition of Environmental Scanning Electron Microscopy (ESEM) micrographs recorded at three tilt angles and at different dissolution times. This method allowed the reconstruction of the topography of the solid/liquid interface. By monitoring the evolution of the solid/liquid interface during dissolution by means of 3D reconstructions, we were able to observe preferential dissolution zones in the vicinity of the PGMs particles and to determine microscopic dissolution rates for several regions of interest. PGMs particles were found mainly at the grain boundaries. In 0.1 M HNO3 solution at 60 °C, the normalized dissolution rate for uranium at the grain boundaries reached RL(U) = (7 ± 1) × 10−2 g·m−2·d−1, a value similar to the normalized dissolution rate determined for the whole image over the first 30 days of the experiment. This result showed that the dissolution occurred mainly at the UO2 grain boundaries in the vicinity of PGMs particles. Furthermore, the 3D reconstructions of the solid/liquid interface were used to determine the evolution of the surface area of the pellet. By combining the weight losses determined at the macroscopic scale using the uranium concentrations in solution with the reactive surface area values, it was possible to estimate an effective normalized dissolution rate for the whole pellet.

Sequential Bioleaching of Phosphorus and Uranium

Phosphorus and uranium are both vital elements for society. In recent decades, fears have arisen about the future availability of low-cost phosphorus and uranium. This has resulted in pressure to de-centralize production of both elements by utilizing lower-grade or complex deposits. The research presented here focused on phosphorus-containing apatite ores with uranium impurities; in order to separate uranium by selective and sequential bioleaching before phosphorus leaching. This would create an alternative process route for solvent-extraction, used to remove/recover uranium from the phosphorus acid product of apatite H2SO4 wet process. In this work, it was seen that the used fluorapatite ore required 24 h leaching at pH 1 by H2SO4 to result in 100% leaching yield for phosphorus. As this ore did not contain much uranium, an artificial fluorapatite-uranium ore was prepared by mixing standard uranium ore and fluorapatite. The research with this ore showed that 89% of uranium dissolved in 3 days at pH > 2 and leaching was improved by applying Fe3+ oxidant. In these conditions only 4% of phosphorus was leached. By prolonged (28 days) leaching 95% uranium yield was reached. According to the experiments, the iron in the uranium leach solution would be mainly Fe3+, which allows the use of H2O2 for uranium recovery and then direct use of spent leachate for another uranium leaching cycle. After the dissolution of uranium, 90% of phosphorus was dissolved by decreasing the pH to 1.3. This was done by bioleaching, by utilizing biogenic sulfur oxidation to sulfuric acid.

Surface and electrochemical controls on UO2 dissolution under anoxic conditions

The escape of radionuclides from underground spent nuclear fuel disposal facilities will likely result from anoxic dissolution of spent nuclear fuel by intruding groundwater. Anoxic dissolution of various forms of uranium dioxide (UO2), namely bulk pellet, powder and thin film, has been investigated. Long-duration static batch dissolution experiments were designed to investigate the release of uranium ions in deionized water and any surface chemistry that may occur on the UO2 surface. The dissolved uranium concentration for anoxic dissolution of nearly stoichiometric UO2 was found to be of the order of 10−9 mol/l for the three different sample types. Further, clusters (∼500 nm) of homogenous uranium-containing precipitates of ∼20–100 nm grains were observed in thin film dissolution experiments. Such a low solubility of UO2 across sample types and the observation of secondary phases in deionized water suggest that anoxic UO2 dissolution does not only occur through a U(IV)(solid) to U(VI)(aqueous) process. Thus, we propose that dissolution of uranium under anoxic repository conditions may also proceed via U(IV)(solid) to U(IV)(aqueous), with subsequent U(IV) (precipitates) in a less defective form. Quantitative analysis of surface-sensitive EBSD diffractograms was conducted to elucidate lattice-mismatch induced cracks observed in UO2 thin film studies. Variable temperature anoxic dissolution was conducted, and no increased uranium concentration was observed in elevated temperatures.

Radiation induced dissolution of U3Si2 - A potential accident tolerant fuel

To assess the integrity of the accident tolerant fuel candidate U3Si2 under geological repository conditions, the kinetics of γ-radiation- and H2O2- induced oxidative dissolution was studied. The experiments were performed in aqueous solutions containing 10 mM HCO3− and in solutions without added HCO3−. The same experiments were also performed on UO2 for comparison. All experiments were performed using powder suspensions.The experiments show that U3Si2 is less than one order of magnitude more reactive towards H2O2 than is UO2. The dissolution yield of U3Si2 slightly exceeds the theoretical yield (23%). In experiments with consecutive additions of H2O2 in HCO3− solutions, the reactivity remains constant implying that no significant amount of a secondary phase is formed on the U3Si2 surface. The dissolution of Si closely follows that of U in HCO3− solution. In solutions without added HCO3− the reactivity towards H2O2 is reduced by a factor less than 2. The dissolution is slightly slower than in HCO3− containing solutions but precipitation of U is observed after some time. The results of consecutive additions of H2O2 to the HCO3− free system shows that the reactivity is decreasing for every addition. This indicates that a secondary phase is formed. XRD shows that the secondary phase is studtite.The irradiation experiments show that the surface area normalized radiation chemical yields for uranium dissolution for U3Si2 and UO2 in HCO3− solution differ by a factor 5–10 in favour of UO2. This difference can largely be attributed to the difference in dissolution yield.

Highly Efficient and Selective Dissolution Separation of Fission Products by an Ionic Liquid [Hbet][Tf2N]: A New Approach to Spent Nuclear Fuel Recycling.

Here, we propose the use of carboxyl-functionalized ionic liquid, [Hbet][Tf2N], to separate the fission products from spent nuclear fuels. This innovative method allows the selective dissolution of neutron poisons, lanthanides oxide, as well as some fission products with high yield, leaving most of the UO2 matrix and minor actinides behind in the spent nuclear fuel and accomplishing the actinides recovery as a group. Water-saturated [Hbet][Tf2N] can dissolve lanthanides oxide from simulated spent nuclear fuel with a dissolution ratio of 100% at 40 °C. However, the dissolution of uranium is almost negligible (<1%) under the same conditions. This big difference in dissolution provides a novel separation approach to spent nuclear fuel recycling and may open new perspectives for spent nuclear fuel reprocessing. The recovery of Nd and U from metal-loaded ionic liquids and the recyclability of the ionic liquid [Hbet][Tf2N] have also been investigated. Furthermore, a U/ x value related to the lattice energy U of metal compound M xO y is used to elaborate the solubility. This work represents the first case for efficient fission products removal by selective dissolution, avoiding the complete dissolution of spent nuclear fuel, the producing of the large high-level radioactive waste, and reducing environmental hazards.

Mineralogy Controlled Dissolution of Uranium from Airborne Dust in Simulated Lung Fluids (SLFs) and Possible Health Implications.

The recent increase in cardiovascular and metabolic disease in the Navajo population residing close to the Grants Mining District (GMD) in New Mexico is suggested to be due to exposure to environmental contaminants, in particular uranium in respirable dusts. However, the chemistry of uranium-containing-dust dissolution in lung fluids and the role of mineralogy are poorly understood, as is their impact on toxic effects. The current study is focused on the dissolution of xcontaining-dust, collected from several sites near Jackpile and St. Anthony mines in the GMD, in two simulated lung fluids (SLFs): Gamble's solution (GS) and Artificial Lysosomal Fluid (ALF). We observe that the respirable dust includes uranium minerals that yield the uranyl cation, UO2 2+, as the primary dissolved species in these fluids. Dust rich in uraninite and carnotite is more soluble in GS, which mimics interstitial conditions of the lungs. In contrast, dust with low uraninite and high kaolinite is more soluble in ALF, which simulates the alveolar macrophage environment during phagocytosis. Moreover, geochemical modeling, performed using PHREEQC, is in good agreement with our experimental results. Thus, the current study highlights the importance of site-specific toxicological assessments across mining districts with the focus on their mineralogical differences.

Radiation induced dissolution of (U, Gd)O2 pellets in aqueous solution – A comparison to standard UO2 pellets

The behavior of spent nuclear fuel exposed to groundwater is crucial in the safety assessment of a deep geological repository for spent nuclear fuel. For this reason, leaching experiments on spent nuclear fuel as well as non-radioactive analogues have been conducted for several decades. Although the processes involved can be considered to be fairly well understood, there is a need for further experimental studies whenever new fuel types are introduced. Fuels with burnable absorbers are now in use but very little is known about their behavior under repository conditions. In this work, the impact of burnable absorbers doping (Gd, 3–8%wt.) on the oxidative dissolution of UO2 in an aqueous system was studied in H2O2 and γ-irradiation induced dissolution experiments. The results showed a significant decrease in uranium dissolution and lower reactivity towards H2O2 for (U,Gd)O2 pellets compared to standard UO2. The resulting decrease in the final oxidative dissolution yield was mainly attributed to decreased redox reactivity of the UO2-matrix upon doping. The results of the gamma radiation exposures display an even larger effect of Gd-doping. These findings indicate that other processes are involved in the radiation-induced dissolution of Gd-doped UO2 compared to pure UO2.

Leaching of El-Missikat low-grade fluoritized uranium ore by sulfuric acid: mechanism and kinetic

A well-characterized low-grade fluoritized uranium samples from new occurrence in Gabal El-Missikat prospect, Eastern Desert, Egypt was subjected to sulfuric acid leaching. The effects of leaching parameters on uranium dissolution mechanism were investigated. The shrinking core model was used to model leaching reactions. The kinetics equations indicates that the reactions appear to be controlled by layer diffusion process. The activation energy for uranium dissolution was evaluated. Low activation energy value (2.54 kJ mol−1) confirm the diffusion layer mechanism. The presence of fluoride ions in the solution increases the dissolution of uranium. The optimum process operating parameters were: sulfuric acid concentration: 1.5 M, solid–liquid ratio: 1:3, contact time 8 h; agitation speed rate 200 rpm; and ore particle size − 75 µm at temperature 60 °C, in the absence of an external oxidant. Under these experimental conditions, the extraction efficiency of uranium was about 91%.

Separation of Radioactive Elements from Ethiopian Kenticha Pegmatite Ore by Hydrometallurgical Methods

The leaching and extraction behavior of uranium and thorium from a high-grade Ethiopian pegmatite ore in a mixture of hydrofluoric acid and sulfuric acid has been investigated. The effects of variables such as the temperature, particle size, acid concentration, and leaching time were studied. The leaching efficiency of uranium increased with increasing temperature to 150°C, at which 96% removal of uranium was achieved. Particles in the size range of − 100 + 75 μm resulted in the highest leaching of uranium, while formation of a colloidal suspension was observed when using a fine particle size fraction (− 75 μm). The dissolution of uranium increased with increasing leaching time. No significant systematic dependence of the leachability of thorium on the above variables was observed. Optimum extraction of uranium and thorium using D2EHPA was obtained when using aqueous/organic phase volume ratio of 1:1, solvent concentration of 0.3 M, and contact time of 20 min.

Kinetics of uranium extraction from coffinite—A comparison with other common uranium minerals

Abstract A synthetic coffinite was hydrothermally prepared and characterized before conducting a series of acid sulfate leach tests under conditions of relevance to uranium extraction. The results were then compared with similar studies on synthetic versions of the related U4+ minerals uraninite (UO2) and brannerite (UTi2O6) to identify and differentiate the rate and U extraction among these important uranium minerals. Tests examining the influence of residence time on uranium dissolution from synthetic coffinite, uraninite and brannerite showed that under similar experimental conditions, complete dissolution of uranium from coffinite was obtained between 36 and 48 h. The activation energy for this reaction was calculated to be 38.4 kJ/mol. This represented a significantly slower rate of dissolution than that indicated for uraninite which dissolved in 3 h (Ea=15.2 kJ/mol). The synthetic brannerite was leached at a much slower rate than the coffinite and reached a maximum dissolution of ∼18% U in 144 h (Ea=42–84 kJ/mol). The clear differentiation in rates and U extraction among the three minerals is consistent with previous literatures which suggest that in terms of leachability, uraninite>coffinite>brannerite. It is expected that the presence of impurities in natural coffinites would further inhibit leachability.

Influence of Dissolved Ions on Geochemical Dissolution of Uranium in KURT Granite

Influence of Dissolved Ions on Geochemical Dissolution of Uranium in KURT Granite

Fractal kinetic characteristics of hard-rock uranium leaching with sulfuric acid.

In order to study the fractal dynamic properties of uranium leach mining and discuss the influence of ore crushing on the dynamics of leach mining, uranium mine ore rocks in southern China were selected as the research object and an acid leaching experiment was performed on the ore samples with different fractal dimensions of 1.1, 1.4, 1.7, 2.0, 2.3 and 2.6. In the column leaching experiment, a PVC pipe with an inner diameter of 112 mm and a height of 1500 mm was used. The uranium content was determined by using titanium trioxide that was placed into a 0.1 mg ml−1 standard uranium solution, and a sampling rate of once daily with a 5 ml volume of leaching solution was adopted after 8 h drenching time. The results show that the flow rate of the leaching solution depends on the distribution of the ore's particle size, that is, a larger fractal dimension results in a smaller flow rate. The concentration of the uranium leaching solution reaches a maximum value which subsequently decreases with time on the third day of the experiment, and it seems that the changes in the uranium concentration tend to be stable at around 15 days. Moreover, the concentration seems to increase with the increasing fractal dimension, and the fractal dimension of the ore particle size has a significant impact on the leaching kinetics. When the fractal dimension is between 1.1 and 2.6, the uranium dissolution rate, K, increases with the increasing fractal dimension. The kinetic reaction of the uranium leaching is a liquid–solid one, which is controlled by chemical reactions in the earlier phase. While the middle reaction phase is mainly chemical-diffusion reaction coupling, and the latter part of the reaction is controlled by diffusion. As the fractal dimension increases, the liquid–solid reaction controlled by diffusion appears at earlier phases. When the fractal dimension is greater than 2.0, the time of entering the diffusion control phase only changed little with the increasing of the fractal dimension. At last, a fractal dimension of 2.0 is suggested for the acid leaching of uranium ore crushing.

The role of surface-bound hydroxyl radicals in the reaction between H2O2 and UO2

In this work, we have studied the reaction between H2O2 and UO2 with particular focus on the nature of the hydroxyl radical formed as an intermediate. Experiments were performed to study the kinetics of H2O2 consumption and uranium dissolution at different initial H2O2 concentrations. The results show that the consumption rates at a given H2O2 concentration are different depending on the initial H2O2 concentration. This is attributed to an alteration of the reactive interface, likely caused by blocking of surface sites by oxidized U/surface-bound hydroxyl radicals. The dissolution yield given by the amount of dissolved uranium divided by the amount of consumed hydrogen peroxide was used to compare the different cases. For all initial H2O2 concentrations, the dissolution yield increases with reaction time. The final dissolution yield decreases with increasing initial H2O2 concentration. This is expected from the mechanism of catalytic decomposition of H2O2 on oxide surfaces. As the experiments were performed in solutions containing 10 mM and a strong concentration dependence was observed in the 0.2–2.0 mM H2O2 concentration range, we conclude that the intermediate hydroxyl radical is surface bound rather than free.

Quantification of kinetic rate law parameters for the dissolution of natural autunite in the presence of aqueous bicarbonate ions at high concentrations

Uranium is a key contaminant of concern in the groundwater at U.S. Department of Energy (DOE) facilities within the United States and is a potential source of groundwater contamination and a risk to human health and the environment through discharges to surface water. Dissolved inorganic carbon (bicarbonate/carbonate) has a high affinity for complexing with uranium that is present as sorbed or unique uranium-bearing mineral phases within the sedimentary matrix. This process can result in the formation of soluble uranyl carbonate aqueous species, which are mobile under circumneutral pH conditions. This study was conducted to quantify the rate of release of uranium from the autunite mineral, (Ca[(UO2)(PO4)]2•3H2O), that was formed during polyphosphate injection to remediate uranium; the dissolution of uranium was studied as a function of the aqueous bicarbonate concentration, ranging from 25 to 100 mM. Experiments were carried out in the pH range from 7 to 11 in the temperature range of 23–90 °C via single-pass flow-through testing. Consistent with the results of previous studies (Gudavalli et al., 2013a, 2013b), the rate of uranium release from autunite exhibited minimal dependency on temperature, but was strongly dependent on pH and increasing concentrations of bicarbonate in the solution. Data obtained during these experiments were compared with results of previous experiments conducted using a low-concentration range of bicarbonate solutions (0.5–3.0 mM). An 8- to 30-fold increase in the rate of uranium release was observed in the presence of high bicarbonate concentrations at pH 7–8 compared to low bicarbonate values, while at pH 9–11, there was only a 5-fold increase in uranium rate of release with an increase in bicarbonate concentrations. The rate of uranium release was calculated to be between 5.18 × 10−8 and 1.69 × 10−7 mol m−2 s−1. The activation energy values at high and low bicarbonate concentrations were similar, with ratio values in the range of 0.6–1.0.

Dissolution of studtite [UO2(O2)(H2O)4] in various geochemical conditions

This study determined the dissolution rate of studtite, (UO2)O2(H2O)4, which can be formed by reaction between H2O2 and UO22+ that leaks from spent nuclear fuel (SNF) in deep geological repositories. The batch dissolution experiments were conducted using synthesized studtite under different solution conditions with varying pHs and concentrations of HCO3− and [H2O2] in synthetic groundwater. The experimental results suggested that carbonate ligand and H2O2 in groundwater accelerated the dissolution of studtite and uranium (U) release. Above 10−5 M of H2O2 initial concentration, the released uranium concentration in solution decreased, possibly as a result of reprecipitation of studtite due to reaction between uranium and H2O2. The results will be useful to assess the comprehensive transport of uranium from both nuclear waste and SNF stored in deep geological repositories.