Dissolution Of UO2 Research Articles

Effect of hydrogen on oxidative dissolution of epsilon particles-doped UO2 pellets under carbonate condition with hydrogen peroxide

The epsilon particles that result from nuclear fission of UO2 fuel possess both advantages and disadvantages from a spent nuclear fuel (SNF) management perspective. In this study, the effect of epsilon particles, namely Ru, Mo, and Pd, inherent in simulated UO2 pellets is examined. Various analytical methods have been used to explore the changes in the structural, surface, and electrochemical properties of which contain epsilon particles. A notable finding is that the epsilon particles are not evenly distributed and tend to clump together, transforming into a metallic state after sintering, as detailed in the X-ray diffraction analyses. Energy-dispersive X-ray spectroscopy analyses highlight interesting aspects of the distribution of elements, especially the disappearance of Pd during sintering, which is likely due to its high vapor pressure. Although the lattice structure of UO2 remains unchanged, the sizes of the grains and pores visibly change, which may influence the tendency of UO2 pellet-cracking. Despite the addition of the epsilon particles, the electrical conductivity analyses show no significant changes, suggesting that they act as minor impurities without affecting the structural lattice. However, their possible role as catalysts in electrochemical reactions opens new and interesting areas that require thorough investigation. Moreover, examining the anodic dissolution under various conditions provides detailed insights into UO2 dissolution and oxidation, revealing how epsilon particles subtly influence the oxidative dissolution process. This study clarifies the basic interactions and effects of epsilon particles in UO2 pellets and broadens the path for a deeper understanding and improvement of nuclear fuel matrices and steering advancements in the safe and effective use of nuclear energy.

The effect of cerium, neodymium, and ytterbium doping on UO2 dissolution

The dissolution rate of spent nuclear fuel under environmental conditions has been studied heavily to understand the impacts of a failed waste package on potential radionuclide release from a geologic repository. Multiple countries have evaluated these scenarios in oxidizing and reducing environments respective to their relevant repository conditions. The repository environment (e.g., water chemistry, temperature, oxygen content) and the fuel itself (i.e., chemical composition of the fuel) both heavily impact the dissolution rate of the spent fuel. This work examined the impact that rare earth element dopants (Ce, Nd, Yb) have on the fuel dissolution under repository relevant conditions with decreasing oxidizing conditions using a single pass flowthrough system. UO2 samples were doped with Ce, Nd, or Yb with concentrations between 1 and 5 at%. The addition of dopants to the samples reduced the dissolution rates on most samples relative to pure UO2 samples. Preliminary experiments that occurred in a less oxidizing environment showed a reduction in dissolution rate compared to fully oxidizing conditions. The results within highlight the importance of dopant behavior in used fuel dissolution modeling.

The influence of bicarbonate concentration and ionic strength on peroxide speciation and overall reactivity towards UO2.

H2O2 produced from water radiolysis is expected to play a significant role in radiation induced oxidative dissolution of spent nuclear fuel under the anoxic conditions of a deep geological repository if the safety-barriers fail and ground water reaches the fuel. It was recently found that the coordination chemistry between U(vi), HCO32- and H2O2 can significantly suppress H2O2 induced dissolution of UO2 in 10 mM bicarbonate. This was attributed to the much lower reactivity of the U(vi)O22+-coordinated O22- as compared to free H2O2. We have extended the study to lower bicarbonate concentrations and explored the impact of ionic strength to elucidate the rationale for the low reactivity of complexed H2O2. The experimental results clearly show that dissolution of U(vi) becomes suppressed at [HCO3-] < 10 mM. Furthermore, we found that the reactivity of the peroxide in solutions containing U(vi) becomes increasingly more suppressed at lower carbonate concentration. The suppression is not influenced by the ionic strength, which implies that the low reactivity of O22- in ternary uranyl-peroxo-carbonato complexes is not caused by electrostatic repulsion between the negatively charged complex and the negatively charged UO2-surface as we previously hypothesized. Instead, the suppressed reactivity is suggested to be attributed to inherently higher stability of the peroxide functionality as a ligand to UO22+ compared to as free H2O2.

Gd2O3 Doped UO2(s) Corrosion in the Presence of Silicate and Calcium under Alkaline Conditions

The anodic reactivity of UO2 and UO2 doped with Gd2O3 was investigated by electrochemical methods in slightly alkaline conditions in the presence of silicate and calcium. At the end of the experiments, the electrodes were analysed by X-ray photoelectron spectroscopy to determine the oxidation state of the uranium on the surface. The experiments showed that the increase in gadolinia doping level led to a reduction in the reactivity of UO2, this effect being more marked at the highest doping level studied (10 wt.% Gd2O3). This behaviour could be attributed to the formation of dopant-vacancy clusters (GdIII-Ov), which could limit the accommodation of excess O2− into the UO2 lattice. In addition, the presence of Ca2+ and SiO32− decreased the anodic dissolution of UO2. In summary, the Gd2O3 doping in presence of silicate and calcium was found to strongly decrease the oxidative dissolution of UO2, which is a beneficial situation regarding the long-term management of spent nuclear fuel in a repository.

UO2 dissolution in aqueous halide solutions exposed to ionizing radiation

In this work, we have experimentally studied UO2 dissolution in pure water and in 1 M aqueous solutions of either Cl- or Br- exposed to γ-radiation. It has previously been found that high ionic strength can facilitate adsorption of dissolved UO22+ on UO2 surfaces. The adsorption is also affected by the solution pH relative to the point of zero charge of UO2. In our experiments, Br3- was observed in 1 M Br- solution exposed to γ-radiation. Experiments confirmed that Br3- can quantitively oxidize UO2. XPS and UPS were used to characterize potential surface modifications after exposure. The XPS results show that the UO2 surfaces after exposure to γ-radiation in pure water and in 1 M aqueous solutions of either Cl- or Br- were significantly oxidized with U(V) as the dominating state. U 4f7/2 and O 1 s spectra of the UO2 surface after exposure to γ-radiation in pure water demonstrates the formation of uranyl peroxide secondary phases. UPS results indicate that there is a large percentage of U(VI) on the ultra-thin outer layer of UO2 after exposure to γ-radiation in 1 M aqueous solutions of Br- and Cl-, and 100 % of U(VI) in the pure water case.

Understanding the solid/liquid interface evolution during the dissolution of Nd-doped UO2 by macro-/microscopic dual approach

The chemical durability of Nd-doped UO2 samples was studied using an innovative macro-/microscopic dual approach method. The starting precursor was first prepared by a hydroxide co-precipitation route leading to a nano-sized powder associated to high specific surface area. The powder was then converted into U0.8Nd0.2O2±y solid solution by calcination. After shaping of the oxide powder into the pellet form through uniaxial pressing at 500 MPa, the pellets were sintered at 1600 °C under inert (Ar) or reducing (Ar-4%H2) atmosphere. At the macroscopic scale, multiparametric dissolution tests were carried out under static conditions in 2 mol.L−1 HNO3 at 60 °C. Simultaneously, an operando monitoring of the solid/liquid interface evolution during dissolution was also performed by ESEM at the microscopic scale, in order to follow the consequences of the dissolution on the ceramics microstructure. This dual approach was validated by comparing the calculated normalized dissolution rates obtained for each series of experiments. Changing the sintering atmosphere did not appear to significantly affect the dissolution kinetics of Nd-doped UO2 samples at the macroscopic scale whereas notorious differences in the dissolution mechanisms have been highlighted at the microscopic scale by ESEM monitoring solid/liquid interface.

Modelling radiation-induced oxidative dissolution of UO2-based spent nuclear fuel on the basis of the hydroxyl radical mediated surface mechanism: Exploring the impact of surface reaction mechanism and spatial and temporal resolution

A combined kinetic and diffusion model with an accurate α-dose rate profile was used to model radiation induced dissolution of UO2. Previous experimental data were used to fit the surface site reaction system involving the surface bound hydroxyl radical as an intermediate species for both UO2 oxidation and surface catalysed decomposition of H2O2. The performance of the model was explored in terms of sensitivity to spatial and temporal resolution as well as simplifications in the models describing the surface reactions and the reactions in solution. As a result, optimal conditions for running the numerical simulations were identified.

UO2 dissolution in bicarbonate solution with H2O2: the effect of temperature.

Upon nuclear waste canister failure and contact of spent nuclear fuel with groundwater, the UO2 matrix of spent fuel will interact with oxidants in the groundwater generated by water radiolysis. Bicarbonate (HCO3-) is often found in groundwater, and the H2O2 induced oxidative dissolution of UO2 in bicarbonate solution has previously been studied under various conditions. Temperatures in the repository at the time of canister failure will differ depending on the location, yet the effect of temperature on oxidative dissolution is unknown. To investigate, the decomposition rate of H2O2 at the UO2 surface and dissolution of UVI in bicarbonate solution (0.1, 1, 10 and 50 mM) was analysed at various temperatures (10, 25, 45 and 60 °C). At [HCO3-] ≥ 1 mM, the concentration of dissolved UVI decreased with increasing temperature. This was attributed to the formation of UVI-bicarbonate species at the surface and a change in the mechanism of H2O2 decomposition from oxidative to catalytic. At 0.1 mM, no obvious correlation between temperature and U dissolution was observed, and thermodynamic calculations indicated this was due to a change in the surface species. A pathway to explain the observed dissolution behaviour of UO2 in bicarbonate solution as a function of temperature was proposed.

Simple, Fast, and Selective Dissolution of Eu2O3 in an Ionic Liquid as a Sustainable Paradigm for Lanthanide-Actinide Separations in Radioactive Waste Remediation.

The liquid-liquid extraction (LLE) process for lanthanide-actinide separation from the nuclear fuel cycle has several drawbacks such as, the requirement of cooling for decay heat control, the handling of large volumes of toxic volatile organic compounds (VOCs), and secondary waste generation. Alternatively reprocessing without spent fuel cooling is done by pyroprocessing, which uses high-temperature corrosive molten salts and requires elevated temperature, and is an energy-intensive process. In recent years, some of the shortcomings of both LLE and pyroprocessing are overcome by the use of room temperature ionic liquids (RTILs) as the solvents. In the present work, an attempt was made to exploit the potential of the neoteric, less-corrosive, low-VOC RTILs toward direct dissolution-based separations at ambient conditions. The present paper involves the selective dissolution of Eu2O3 in an RTIL, i.e., C4mim·NTf2 containing 2-thenoyltrifluoroacetone (HTTA) within ca. 30 min at ambient conditions; while the dissolution of AmO2 and UO2 were found to be very poor, making this an attractive method for lanthanide-actinide separation, a key step in radioactive waste management, i.e., an actinide partitioning and transmutation strategy. The quantitative dissolution of Eu2O3 from simulated spent nuclear fuel with different Eu2O3 loading was also shown. Water plays a crucial role in deciding the kinetics of dissolution and amount of the dissolved oxide. The combination of X-ray absorption, fluorescence, and UV-vis spectroscopic studies suggested the formation of the dehydrated anionic complex Ln(TTA)4- to play pivotal role in the oxide dissolution process. The structure of the complex was analyzed by density functional theory and extended X-ray absorption fine structure. The mechanism of oxide dissolution was proposed and electrochemical studies were performed to understand the possible recovery option using electrodeposition of the dissolved Eu3+.

Vadose-zone alteration of metaschoepite and ceramic UO2 in Savannah River Site field lysimeters

Uranium dioxide (UO2) and metaschoepite (UO3•nH2O) particles have been identified as contaminants at nuclear sites. Understanding their behavior and impact is crucial for safe management of radioactively contaminated land and to fully understand U biogeochemistry. The Savannah River Site (SRS) (South Carolina, USA), is one such contaminated site, following historical releases of U-containing wastes to the vadose zone. Here, we present an insight into the behavior of these two particle types under dynamic conditions representative of the SRS, using field lysimeters (15 cm D x 72 cm L). Discrete horizons containing the different particle types were placed at two depths in each lysimeter (25 cm and 50 cm) and exposed to ambient rainfall for 1 year, with an aim of understanding the impact of dynamic, shallow subsurface conditions on U particle behavior and U migration. The dissolution and migration of U from the particle sources and the speciation of U throughout the lysimeters was assessed after 1 year using a combination of sediment digests, sequential extractions, and bulk and μ-focus X-ray spectroscopy. In the UO2 lysimeter, oxidative dissolution of UO2 and subsequent migration of U was observed over 1–2 cm in the direction of waterflow and against it. Sequential extractions of the UO2 sources suggest they were significantly altered over 1 year. The metaschoepite particles also showed significant dissolution with marginally enhanced U migration (several cm) from the sources. However, in both particle systems the released U was quantitively retained in sediment as a range of different U(IV) and U(VI) phases, and no detectable U was measured in the lysimeter effluent. The study provides a useful insight into U particle behavior in representative, real-world conditions relevant to the SRS, and highlights limited U migration from particle sources due to secondary reactions with vadose zone sediments over 1 year.

Reactive Transport Modeling during Uranium In Situ Leaching (ISL): The Effects of Ore Composition on Mining Recovery

Unconsolidated sandstone uranium deposits exploited by the in situ leaching (ISL) method, contain complex tetravalent and hexavalent uranium compounds, mostly as UO2 and UO3 oxides that have different dissolution rates in sulfuric acid solutions. This work investigates a reactive transport model that takes into account the dissolution of both UO2 and UO3 in sulfuric acid solution together with possible interactions with rock minerals during the ISL uranium extraction. Several empirical reaction rate constants were determined during lab experiments on uranium extraction assays, including dissolution rates of tetravalent and hexavalent uranium oxides, and the dissolution rate of rock components by sulfuric acid solution. Effects on the recovery of solution flow rates and ratios between tetravalent and hexavalent uranium compounds are also investigated. The experimental dissolution constants were then used in the proposed reactive transport model to be applied to a real case study in Kazakhstan for comparing the 16 months history matching of an exploitation block consisting of 18 well injectors and 4 producers. The obtained numerical results show good agreement with empirical data gathered during exploitation.

Double diffusive dissolution model of UO2 pellet in molten Zr cladding

ABSTRACT The rapid dissolution of UO2 in molten Zr that could occur during fuel-cladding liquefaction at high temperatures and its kinetics were reformulated considering the convective mass transfer and the scouring effect at the UO2/Zr interface. The mass transfer coefficient of U was obtained as a correlation including the aspect ratio term by computational fluid dynamics (CFD) analysis. To explain the gap between the rapid dissolution rate observed in the experiments and the density-driven convective mass transfer, we introduced an idea in which the eutectic melting at the UO2/Zr interface promotes the solid oxide detachment owing to infiltration of the U-Zr-O liquid into the UO2 grain boundaries (scouring effect). The developed model was validated with UO2-Zr crucible experiments at 2273 and 2373 K. The calculated mass percentage ratios of U/Zr agreed with the measurements, and the transition timings from rapid dissolution stage were consistent with the metallographic observations.

Dopant effect on the spent fuel matrix dissolution of new advanced fuels: Cr-doped UO2 and Cr/Al-doped UO2

To improve the fundamental understanding of the processes controlling spent fuel alteration under deep geological conditions, the influence of dopants (Cr and Cr/Al) on UO2 matrix stability is evaluated for “modern” types of light water reactor (LWR) fuels with additives. The uranium dissolution behavior of as-prepared 0.06 wt%Cr and 0.05 wt%Cr/0.02 wt%Al doped UO2 pellets is studied in simplified groundwater containing 19 mM HCO3− in autoclaves under hydrogen atmosphere. Sintered disks were exposed to simulated highly carbonated conditions, representative of a repository scenario of water intrusion after a hypothetical canister failure. The uranium concentration released was ∼10−7 M for the 0.05 wt%Cr/0.02 wt%Al doped UO2 pellet and ∼10−6 M for 0.06 wt%Cr doped UO2 pellets, after 170 days. The results indicate that the amount of dissolved uranium is slightly lower compared to previous studies in absence of a reductant gas phase, but clearly above the solubility of UO2(am, hyd). The initial measured pH was 8.9 ± 0.1, which gradually approached a constant value of ∼9.2 ± 0.1. Solid characterization at the end of the dissolution experiment, by SEM, Raman spectroscopy and XRD shows that the surface of all pellets remains almost unaltered. The experimental results indicate a potential oxidative dissolution of UO2, which could be attributed to the presence/intrusion of dissolved oxygen in the prepared synthetic groundwater. In order to identify the mechanism of uranium release, the datasets from the batch experiments are simulated with a PHREEQC model previously calibrated with results of existing spent fuel UOx leaching experiments. The model includes the geochemical processes that are relevant for the studied experimental conditions: (i) non-oxidative dissolution of UO2, (ii) UO2 oxidation with O2, (iii) dissolution of U(VI) by carbonate water, (iv) reduction of oxidized U(VI) on the surface pellet by activated H2. The ability to activate the dissolved H2 is studied by implementing a kinetically controlled process of H2 activation on the Cr surface in the model.

Impact of ruthenium metallic particles on the dissolution of UO2 in nitric acid

UO2 pellets incorporating 3 mol.% of Ru was prepared by using a wet chemistry route and then characterised. The speciation, morphology, as well as spatial distribution of Ru in the sintered samples, were determined. The synthesised samples were submitted to dissolution tests in 0.1 M nitric acid at 60 °C and the dissolution of pure UO2 pellets was also studied with and without the presence of Ru metallic particles in the solution. The evolution of the U, Ru, and nitrous acid concentrations in solution was measured and the residues of dissolution were further characterised. The obtained results unambiguously demonstrated the catalytic activity of Ru-metal particles during UO2 dissolution in nitric acid provided that a solid/solid interface existed between UO2 and Ru-metal particles. This positive impact on the dissolution kinetics of UO2 was supported by redox reactions taking place at both nitric acid solution/Ru-metal particles and at Ru-metal particles/UO2 interfaces.

New deep eutectic solvents based on imidazolium cation: Probing redox speciation of uranium oxides by electrochemical and theoretical simulations

Dissolution of Uranium oxides and their redox speciation in two imidazolium-based DES. • New DES based on imidazolium cation are studied for the uranium oxide dissolution. • Redox speciation of UO 3 and UO 2 in imidazolium-based DES was thoroughly probed. • Interestingly, Uranyl species formed in the case of UO 3 but not in the case of UO 2 . • Theoretical simulations adequately validated the experimental results. Two new deep eutectic solvents, different from hitherto known and based on common choline chloride, were prepared. 1-methyl-3-decylimidazolium bromide mixed with hydrogen bond donors, malonic acid and diglycolic acid to form DES. Dissolution of uranium oxides (UO 3 and UO 2 ) was explored in both of them and characterization of the resultant solutions was carried out by IR, TGA and UV–Vis spectroscopy. Redox speciation of dissolved uranium oxides was probed in these redox-active DES through cyclic voltammetry, differential pulse voltammetry and in situ spectroelectrochemical measurements. Interestingly, the formation of uranium oxo species was observed through the dissolution of UO 3 in both the DES. The electrochemical characteristics viz. redox thermodynamics (peak potential and formal redox potential), transport property (diffusion coefficient D 0 ), heterogeneous electron transfer kinetic parameters (αn and k 0 ) and mechanistic electron transfer of the dissolved uranium species in two DES were investigated. The speciation of uranium was decoded through the electronic absorption spectra of uranium in its lower oxidation states [U(V) and U(IV)] acquired through in situ spectroelectrochemical electrolysis with varying cathodic potentials. Studies were also carried out using Molecular Dynamics (MD) and density functional theory (DFT) simulations to authenticate the electrochemical outcomes and to acquire the binding energy, optimized structure and molecular orbital diagram of dissolved uranium species. Further, the MD simulation sheds light on the probable equatorial coordinating atoms of dissolved uranium species. This is the first report, to the best of our knowledge, on imidazolium-based DES and actinide ions.

Influence of the concentration of nitric acid on the composition of NOX gas evolved during the dissolution of nuclear fuel and its implications on the PUREX process

Understanding the mechanism of dissolution of UO2 in nitric acid is important for designing a continuous dissolution system for the reaction. Knowledge of the composition of the NOX gas evolved during the course of the reaction is imperative for this understanding as it strongly influences the reaction mechanism. In this work, the profiles of the composition of NOX gas evolved and the ratio of moles of acid consumed per mole of uranium dissolved during the dissolution of UO2 pellets in nitric acid have been determined as a function of acidity. The predominant stoichiometric reaction by which UO2 dissolves at every point of the reaction was clearly identified thereby improving the understanding of its mechanism and facilitates better control over the overall reprocessing process flowsheet.

Potential for uranium release under geologic CO2 storage conditions: The impact of Fe(III)

Saline aquifers are considered ideal subsurface sinks for large amounts of CO2 storage. It is common to have trace levels of uranium-bearing minerals (naturally occurring radioactive material, NORM) in sandstone saline aquifers and CO2 injection may cause uranium mobilization due to the coupled chemical and physical interactions at mineral-water interfaces. In this study, we developed a TOUGHREACT model to assess the uranium mobilization potential from uraninite (UO2) dissolution induced by CO2 injection into a hypothetical CO2 storage reservoir with Fe(III)-bearing hematite in a time scale of 30 years CO2 injection and 100 years after CO2 injection. Numerical simulation results show that injection of CO2 reduced pH and induced small amounts of hematite dissolution, which released Fe3+ into aqueous phase. Fe3+ together with dissolved bicarbonate species caused oxidative dissolution of UO2 (Fe3+ serving as an oxidant), resulting in an increase of dissolved uranium concentrations in the CO2 storage reservoir. However, the released uranium was limited to a small region close to Fe3+ supply source and the maximum concentration of released uranium was only 9.00 × 10−10 mol/kg in the CO2 storage reservoir at t = 30 years. The availability of Fe3+ is the main factor that limits mineralized uranium release because the pH drop induced by CO2 injection is not low enough to cause significant Fe3+ release. For sorbed uranium, the main factor that influences sorbed uranium release is pH because uranium sorption is amphoteric. Based on our simulation results, the pH range in the CO2 storage reservoir does not cause a substantial release of sorbed uranium. Also, the potential for released uranium to migrate through a permeable zone in the caprock to the layer above the caprock and cause groundwater contamination is negligible. In summary, results from this study imply a very low risk of environmental contamination by bicarbonate and Fe(III)-induced uranium mobilization.

Comparative study on aqueous acid free UO2 dissolution-extraction using DHOA adduct into room temperature ionic liquid/supercritical carbon dioxide/n-hexane

Abstract Feasibility was established for direct dissolution-extraction of uranium employing adduct of N,N-dihexyl octanamide (DHOA), thus eliminating discrete aqueous phase and free acid usage. Various aspects of dissolution of solid uranium dioxide and extraction of uranium into molecular diluent viz. n-hexane and neoteric solvents viz. room temperature ionic liquid (RTIL) and supercritical carbon dioxide (SC CO2) were studied. The organic adduct was found to have composition DHOA.(HNO3)0.78(H2O)0.4. Adduct miscibility and UO2 dissolution behavior was markedly different for RTIL and n-hexane. The dissolution process, studied by monitoring UV–Vis spectra, was found to be pseudo first order with a rate constant of of 0.074 min−1 and 0.036 min−1 for n-hexane and RTIL respectively. Irrespective of medium, dissolution-extraction efficiency of ≥90% was achievable. Using RTIL for dissolution-extraction medium and SC CO2 for stripping is promising in terms of overall efficiency as well as RTIL recovery by avoiding aqueous cross contamination.

Anaerobic Dissolution Rates of U(IV)-Oxide by Abiotic and Nitrate-Dependent Bacterial Pathways.

The long-term stability of U(IV) solid phases in anaerobic aquifers depends upon their reactivity in the presence of oxidizing chemical species and microbial catalysts. We performed flow-through column experiments under anaerobic conditions to investigate the mechanisms and dissolution rates of biogenic, noncrystalline UO2(s) by chemical oxidants (nitrate and/or nitrite) or by Thiobacillus denitrificans, a widespread, denitrifying, chemolithoautotrophic model bacterium. Dissolution rates of UO2(s) with dissolved nitrite were approximately 5 to 10 times greater than with nitrate alone. In the presence of wild-type T. denitrificans and nitrate, UO2(s) dissolution rates were similar to those of abiotic experiments with nitrite (from 1.15 × 10-14 to 4.94 × 10-13 mol m-2 s-1). Experiments with a T. dentrificans mutant strain defective in U(IV) oxidation supported microbially mediated U(IV) oxidation. X-ray absorption spectroscopy (XAS) analysis of post-reaction solids showed the presence of mononuclear U(VI) species rather than a solid U(VI) phase. At steady-state U release, kinetic and spectroscopic results suggest detachment of oxidized U(VI) from the UO2(s) surface as the rate-determining step rather than electron transfer or ion diffusion. Under anaerobic conditions, production of nitrite by nitrate-reducing microorganisms and enzymatically catalyzed, nitrate-dependent U(IV) oxidation are likely dual processes by which reduced U solids may be oxidized and mobilized in the aqueous phase.

Uranium release surrounding a single fracture in a uranium-rich reservoir under geologic carbon storage conditions

Though geologic carbon storage (GCS) is widely recognized as a promising strategy to reduce emissions of greenhouse gas (GHG), the potential for mobilization of radioactive uranium (U) from U-bearing minerals in deep subsurface due to CO2 injection remains a concern. In this study, supercritical CO2 and brine flowing through a fracture surrounded by reservoir rock containing uranium is simulated so as to study the potential of uranium release as a result of CO2 injection and the impact of various factors on total uranium release rate. Mineral compositions of the reservoir rock are from previously published literature, which mimics typical sandstone mineral compositions. The reservoir rock is assumed to have 2 × 10−4 vol% UO2 in solid phase. Simulation results show that CO2 injection induces UO2 dissolution, and both CO2 and mobilized uranium are able to migrate in both the fracture and the rock matrix surrounding the fracture. However, the released uranium concentration is quite low in a 60-day simulation period. Mineral dissolution causes a very small porosity increase surrounding the fracture in the simulation period. Sensitivity analysis shows that an increase of UO2 specific surface area and UO2 content in reservoir rock causes a significant increase in released uranium concentration. In other words, total uranium release rate is positively correlated with the specific surface area of UO2 and UO2 content in reservoir rock because the increase of specific surface area and UO2 content increases the total area of UO2 in contact with HCO3− and O2, which raises total uranium release rate. In summary, uranium release in and surrounding the fracture is mainly controlled by uranium supply of the reservoir rock, and the risk of environmental contamination by CO2-induced uranium release is quite low in the scenario reported.