Corrosion Of UO2 Research Articles

Oxidation of accident tolerant fuels models based on Cr-doped UO2 for the safety of nuclear storage facilities

Oxidation of Accident-Tolerant Fuels (ATF) is important from the nuclear safety point of view because it could provoke the release of radioactive material. Despite of the use of ATFs in current nuclear reactors, no studies regarding their oxidation are available. Thus, we propose here an experimental study on Cr2O3-doped UO2 samples, which were subjected to a thermal treatment ranging from room temperature up to 973 K in oxidant conditions. The aim is to assess the potential oxidation resistance of Cr-doped UO2 based model systems. Next, the effect of oxygen partial pressure and the influence of Cr2O3 dopant concentration concentrations on fresh UO2 fuel oxidation were evaluated. Two oxygen compositions (1% and 21%O2) were used to further evaluate the impact of Cr2O3 concentrations on UO2 oxidation. The results show that, despite no effect of the dopant is found in air (21% O2), there is a clear decrease in the oxidation degree at 1%O2. These results are intended to provide a fundamental reference for future investigations of Cr2O3-doped UO2 corrosion behavior and performance to be accounted in new designs of storage facilities.

Oxygen diffusion coefficient and characterization of leachant in UO2 corrosion studied by new methods

UO2 corrosion is a key issue in the study of the safe storage of spent nuclear fuel (SNF). In this work, corrosion experiments were performed on UO2 samples exposed to 18O labelled water during 290 days at room conditions, thereafter a complete analysis of the solid and leachant (labelled water) allow us obtaining new data and exploring new methods.Thus, a new application of the Nuclear Reaction Analysis (NRA), as a simple and low-cost approach, has been developed to estimate the oxidized layer thickness and the diffusion coefficient of oxygen through the oxidized layer. These results were verified by comparison with the ones obtained by traditional Secondary Ion Mass Spectrometry (SIMS).The diffusion coefficient obtained is two orders of magnitude larger than the scarce published ones at room temperature, highlighting the great effect of the composition of the oxidized layer due to the UO2 corrosion on this parameter. In addition, the water used in the corrosion experiments (leachant) has been analyzed by Raman spectroscopy, being the main purpose for the in situ analysis.The data obtained in this study bring out the great ability (higher than expected) of the oxygen to break through the oxidized layer which could endanger the SNF pellet integrity.Besides, this work shows new methods for analyzing corrosion experiments which can be used to obtain new data for completing the database.

Hydrogen peroxide decomposition on simulated nuclear fuel bicarbonate/carbonate solutions

The behavior of SIMFUEL (rare earth doped uranium dioxide containing noble metal particles) has been studied in bicarbonate/carbonate solutions (pH = 9.7) containing hydrogen peroxide, using a combination of electrochemical techniques, scanning electron microscopy (SEM), X-ray photoelectron spectroscopy (XPS), and inductively-coupled plasma mass spectrometry (ICP-MS). The hydrogen peroxide was found to decompose to oxygen and water both homogeneously and heterogeneously, accompanied by a minimal amount of corrosion of UO2. Homogeneous decomposition proceeded via a peroxycarbonate (CO42-) intermediate while heterogeneous decomposition was catalyzed by the reversible UIV⇌UV redox transformation in a thin U1−2xIVU2xVO2+x surface layer. The rate of the heterogeneous decomposition reaction depended on whether UVI surface species ((UVIO2CO3)ads, UVIO3•yH2O) were allowed to accumulate on the surface, blocking access of hydrogen peroxide to the catalytic surface layer.

Modelling uranium dioxide corrosion under repository conditions: A pore-scale study of the chemical and thermal processes

This paper investigates the corrosion behaviors of spent uranium dioxide (UO2) fuel when placed in geological repositories. We performed pore-scale reactive transport simulations of the UO2 corrosion process in a defective fuel rod with different orientations of breach or fracture on its clad. It is found that the corrosion rate has strong dependency on the breach orientation. The highest corrosion rate of UO2 is calculated when the angle between the groundwater flow direction and the clad fracture reaches 180°. UO2 corrosion simulations with different flow rates are compared. The results show the higher flow rate accelerates the corrosion of UO2 fuel. The effect of pH on the corrosion process is also determined. It is indicated that higher pH can limit the corrosion of UO2 by reducing the reaction rate. The dependence of reaction rate and time on reactive surface area is explored. Spent fuel with lower surface area demonstrates longer lifetime under corrosion conditions. A numerical model coupling thermal conduction and chemical reactions is developed to assess the impact of temperature on the process of UO2 corrosion. The results show that higher temperature leads to larger corrosion rates for UO2. The predicted reaction rates are higher in comparison with the isothermal results. Hence, our results can help improve the fundamental understanding of UO2 corrosion in geological repositories for long-term storage of spent nuclear fuels and provide guidance for the safe operations and selection of appropriate repositories.

Comparing the corrosion of uranium nitride and uranium dioxide surfaces with H2O2

Uranium mononitride, UN, is considered a potential accident tolerant fuel due to its high uranium density, high thermal conductivity, and high melting point. Compared with the relatively inert UO2, UN has a high reactivity in water, however, studies have not considered the significant effect of radiation, which is known to cause corrosion of UO2. This study uses 0.1 M H2O2 to simulate the effects of water radiolysis in order to compare the radiolytic corrosion rates of UO2, UN, and U2N3 thin films at room temperature. X-ray reflectivity was used to investigate the changes in film morphology as a function of H2O2 exposure time, allowing changes in film thickness and roughness to be observed on the Ångstrom length-scale. Results showed significant differences between UO2, UN, and U2N3, with corrosion rates of 0.083(3), 0.020(4), and 0.47(8) Å/s, respectively, showing that UN corrodes more slowly than UO2 in 0.1 M H2O2.

Structure-thermodynamics relationship of schoepite from first-principles.

The relationship between the structure and thermodynamic properties of schoepite, an important uranyl phase with formula [(UO2)8O2(OH)12]·12H2O formed upon corrosion of UO2, has been investigated within the framework of density functional perturbation theory (DFPT). Experimental crystallographic lattice parameters are well reproduced in this study using standard DFT. Phonon calculations within the quasi-harmonic approximation predict standard molar entropy and isobaric heat capacity of S0 = 179.60 J mol-1 K-1 and C0P = 157.4 J mol-1 K-1 at 298.15 K, i.e., ∼6% and ∼4% larger than existing DFPT-D2 calculations. The computed variation of the standard molar isobaric heat capacity with water content from schoepite (UO3·xH2O, x = 2.25) to dehydrated schoepite (x = 1) is predicted to be essentially linear along isotherms ranging from 100 to 500 K. These findings have important implications for the dehydration of layered uranyl corrosion phases and hygroscopic materials.

Radiolytic corrosion of uranium dioxide induced by He2+ localized irradiation of water: Role of the produced H2O2 distance

The short-range (few μm in water) of the α-emitting from the spent fuel involves that the radiolytic corrosion of this kind of sample occurs at the solid/solution interface. In order to establish the role of localization of H2O2 species produced by the He2+ particle beam in water from the surface, we perform UO2 radiolytic corrosion experiment with different distance between H2O2 production area and UO2 surface. Then, in this work, the radiolytic corrosion of UO2 particles by oxidative species produced by 4He2+ radiolysis of water was investigated in open to air atmosphere. The dose rate, the localization of H2O2 produced by water radiolysis and the grain boundaries present on the surface of the particles were investigated. UO2 corrosion was investigated by in situ (during irradiation) characterization of the solid surface, analysis of H2O2 produced by water radiolysis and quantification of the uranium species released into the solution during irradiation. Characterization of the UO2 particles, surface and volume, was realized by Raman spectroscopy. UV–vis spectrophotometry was used to monitor H2O2 produced by water radiolysis and in parallel the soluble uranium species released into the solution were quantified by inductively coupled plasma mass spectrometry. During the He2+ irradiation of ultra-pure water in contact with the UO2 particles, metastudtite phase was formed on the solid surface indicating an oxidation process of the particles by the oxidative species produced by water radiolysis. This oxidation occurred essentially on the grain boundaries and was accompanied by migration of soluble uranium species (U(VI)) into the irradiated solution. Closer to the surface the localization of H2O2 formation, higher the UO2 oxidation process occurs, whereas the dose rate had no effect on it. Simultaneously, closer to the surface the localization of H2O2 formation lower the H2O2 concentration measured in solution. Moreover, the metastudtite was the only secondary phase formed whatever the irradiation conditions. One hypothesis proposed in this work is the H2O2 may undergo a dismutation reaction leading to the formation of OH at the UO2 surface.

Corrosion of Agr Fuel Pin Steel Under Conditions Relevant to Permanent Disposal

The majority of the UK nuclear reactor fleet is made up of the Advanced Gas-cooled Reactor type (AGR). These utilise fuel rods made from ceramic UO2 fuel encased in a 20/25/Nb stainless steel cladding. Currently, spent fuel rods undergo reprocessing after removal from the reactor, but the option of direct disposal of spent fuel in a Geological Disposal Facility (GDF) is now under examination. Direct disposal is the current policy in several mature nuclear states (USA, Sweden, Spain, and Finland are examples). The vast majority of spent oxide fuel in these states is from Light Water Reactors (LWR), and the disposal of the UK’s gas-cooled reactor fuel may behave differently due to differences in materials and operating temperature. The GDF proposal would involve spent fuel rods being placed in copper or steel canisters and buried several hundred meters underground in tunnels backfilled with clay. It is assumed that over several thousand years groundwater from the environment will penetrate these barriers and come into contact with the fuel surface and steel cladding. The corrosion behaviour of the fuel ceramic and steel cladding is of interest as part of the safety case for the construction of a GDF. The GDF is an anoxic environment, but alpha-radiolysis of water from the activity of the fuel will create a source of near-field peroxide oxidants. Anaerobic corrosion of the cladding is expected to yield Fe2+ and H2 at the surface of the spent AGR fuel, along with species such as Ni(II) and CrO42-. These species are expected to inhibit the dissolution of the UO2 into the groundwater, but the situation may be complicated by the interaction of Fe(II) with radiolytically generated H2O2 via the Fenton cycle. The ceramic UO2 fuel can be oxidised by the species generated by radiolysis, converting the uranium from U(IV), which has very low solubility in water, to U(VI), which is much more soluble. This oxidation will lead to dissolution of the uranium fuel, allowing uranium and fission products to escape the ceramic fuel matrix and enter the groundwater. This contaminated groundwater could pose a danger to life as it is transported through the biosphere. The effect of hydrogen peroxide on the corrosion behaviour of AGR fuel pin steel has been investigated under conditions relevant to permanent disposal in a geological disposal facility. Electrochemical experimental methods including open circuit potentials and linear sweep voltamograms were used to understand the corrosion susceptibility over a range of potentials. Comparisons to other grades of steel and the effect of the high proportion of nickel present examined. Linear polarisation of the steel shows pitting beginning to occur on the steel surface, localised around the NbC precipitates present, at a potential of around -150mV vs. SCE before undergoing severe trans-passive corrosion at around 400mV vs. SCE. The potentials where the onset of pitting is seen are similar to the expected potentials in repository conditions. An increase in concentration of hydrogen peroxide in the electrolyte was found to reduce the region of passivity in the steel, leading to an increase in corrosion potential. The range of potentials expected in a repository was identified. The corrosion products released in this region (Fe, Ni, and Cr species) and their expected concentrations were investigated by ICP OES analysis. The effect of corrosion on the surface of the steel was examined using Electrochemical Impedance Spectroscopy and Scanning Electron Microscopy. Changes to the UO2 corrosion behaviour under the influence of the cladding corrosion products are discussed.

Corrosion Behaviour of Agr Simfuels

In the UK, the majority of spent nuclear fuel (SNF) is from indigenous Advanced Gas-cooled Reactors (AGRs). AGRs, whilst using UO2-based fuel, employ CO2 as coolant and are graphite moderated. Further, the fuel assembly cladding is comprised of 20/25/Nb stainless steel rather than zircalloy as is the case in Pressurised Water Reactors (PWRs). Consequently, AGR fuel has unique characteristics that need to be evaluated before disposal in an expected anoxic UK geological repository. Pure UO2 pellets and, for the first time, AGR SIMFUEL pellets simulating 25 (pellet interior) and 43 GWd/tU burn-up (rim structure) have been fabricated at the UK National Nuclear Laboratory. Using combined electrochemical and Raman spectroscopic studies, we have investigated the effect of the SIMFUEL dopants on the UO2 crystal matrix. We have also studied the effect of H2O2 on the SIMFUEL surface as a simulant for the effect of a radiolysis on intruding groundwater in the near field. Raman microscopy reveals an increase in the lattice damage (500 – 700 cm-1) with increasing burn-up, as the UO2 cubic fluorite lattice structure becomes more distressed and moves towards a tetragonal structure. Cyclic voltammetric studies of the AGR SIMFUELs reveal them to more susceptible to electrochemical oxidation then PWR SIMFUELs; in particular the peak associated with the in-grain UO2 to UO2+x oxidation (onset -0.2 V vs SCE, Ep = +0.14 V) is ~3X larger for AGR SIMFUELs than PWR SIMFUELs of similar simulated burnup. Importantly, analogous studies on PWR SIMFUELs have shown that UO2+x locations act as preferential corrosion sites. The open circuit potential (EOC) for UO2 and AGR SIMFUELs show a logarithmic dependence on [H2O2] in modified simplified groundwater(10 mmol dm-3 NaCl, 2 mmol dm-3 NaHCO3, pH 6.8). In contrast, the EOC of PWR SIMFUELs of ~0.1 V vs SCE in aqueous alkaline solution is near independent of [H2O2], an observation attributed to surface-catalysed H2O2 decomposition by non-stoichiometric U(IV)/U(V) sites. This suggests that, in contrast to the non-complexing alkaline conditions, carbonate complex-ation of U(VI) in groundwater simulant removes the H2O2 decomposition sites from the UO2 surface so resulting in a larger net [H2O2] and thus oxidative stress being induced at the SIMFUEL surface. At [H2O2] = 10 mmol dm-3, close to that encountered in an a dominated near field, H2O2 generates an oxidative stress of 0.05 V vs SCE on 43 GWd/tU AGR SIMFUEL in simplified modified groundwater. This corresponds to the near top of the voltammetric wave for the in-grain UO2 to UO2+x process, suggesting that near field peroxide may be capable of inducing a substantial population of anodically active UO2+x sites on AGR SNF and consequent localised site corrosion at the spent fuel surface. Extended duration Raman studies on the oxidation of UO2 on 25 GWd/tU SIMFUELs in 1 mmol dm-3 H2O2 (EOC = +0.19 V vs SCE) in O2-free deionised water reveal that, over a 1 week period, there is an increase over in the intensity of the peak at 540 cm-1, associated with the formation of defect sites in the UO2 fluorite lattice. Shorter term studies on undoped UO2 samples in 0.01 to 1 mmol dm-3 H2O2 solutions also reveal the in-growth of a peak at ~610 cm-1, a feature that has been attributed to O sublattice distortions due to the generation of UO2+x locations. This again indicates of the potential for near field concentration H2O2 to induce an increased susceptibility to corrosion for AGR SIMFUEL. The ultimate product of H2O2 induced corrosion of either UO2 or AGR SIMFUELs would be expected to be U(VI). The extended duration Raman studies show that a peak appears at 820 cm-1 after 5 hours exposure to 1 mmol dm-3 H2O2. This grows in intensity up to 50 hours exposure and has been assigned to metastudtite (UO2(O2) 2H2O), a uranium peroxide secondary phase formed at the SIMFUEL surface as a result of the reaction between H2O2 and UO2 2+ – the latter necessarily generated as a result of above-expected H2O2–driven corrosion of the AGR SIMFUEL UO2 matrix. Peroxide aside, open circuit potential measurements on 43 GWd/tU simulated burnup AGR SIMFUELs samples in intimate contact with 20/25/Nb stainless steel and in modified simplified groundwater electrolyte find that they exhibit a mixed potential of ~-0.12 V vs SCE. This corresponds to nearly a quarter of the way up the oxidation wave for in-grain UO2 to UO2+x process in the associated voltammogram – implying that UO2/steel coupling can induce UO2 corrosion even in the absence of a peroxide generated oxidative stress.

The Anodic Reactions on Simulated Spent Fuel (SIMFUEL) in H2O2 Solutions: Effect of Carbonate/Bicarbonate

The anodic behavior of simulated nuclear fuel (SIMFUEL) in solutions containing H2O2 and HCO3−/CO32− has been studied electrochemically and using surface analytical techniques, in particular X-ray photoelectron spectroscopy. Two anodic reactions are possible, the oxidative dissolution of UO2 and H2O2 oxidation. The rates of both reactions are controlled by the chemical release of UVI surface species, and can both be increased by the addition of HCO3−/CO32−. Under anodic conditions the dominant reaction is H2O2 oxidation, although UO2 dissolution may also be accelerated by the formation of a uranyl peroxocarbonate complex. Similarly, under open circuit (corrosion) conditions both UO2 corrosion and H2O2 decomposition are also controlled by the rate of release of UVI surface species which blocks access of H2O2 to the underlying conductive UIV1–2xUV2xO2+x surface. Presently, the balance between these two reactions is not known.

Density-functional theory investigation of oxidative corrosion of UO2

Corrosion and weathering of uranium dioxide (UO2) is a serious concern in a broad range of technological and environmental systems. Oxidation of UO2 can compromise the integrity of nuclear fuel rods, as well as result in the bioavailability of uranium in contaminated ground water at mines, mills, and nuclear waste storage facilities. How the oxidation proceeds, however, is not well understood. In this work, density-functional theory and ab initio thermodynamics are used to delineate the initial stages of surface and subsurface oxidation of UO2 at the (111) surface as a function of temperature and oxygen pressure. Initially, chemisorption of oxygen on the clean stoichiometric surface results in formation of highly stable triple-bonded uranyl groups and oxidation of the topmost uranium atoms to U6+ at a minimal p(O2) near 0K. Once the surface is saturated with uranyl groups and the oxygen chemical potential increases above −1.0eV, subsurface oxidation becomes thermodynamically favored. The degree of oxidation of the subsurface uranium atoms is determined by quantifying the transfer of electrons from the localized U 5f bands to those dominated by the delocalized O 2p bands as oxygen atoms occupy octahedral interstitial sites in the UO2 lattice. Occupation of the octahedral site nearest the surface results in an expansion of the lattice, whereas movement of the oxygen interstitial to deeper sites results in a net contraction.

Incorporation of cerium and neodymium in uranyl phases

The potential for incorporating rare earth elements (REE) into/onto crystalline compounds has been evaluated by precipitating uranyl phases from aqueous solutions containing either cerium or neodymium. These REEs serve both as monitors for evaluating the potential repository behavior of REE radionuclides, and as surrogate elements for actinides (e.g., Ce4+ and Nd3+ for Pu4+ and Am3+, respectively). The present experiments examined the behavior of REE in the presence of ianthinite ([U24+(UO2)4O6(OH)4(H2O)4](H2O)5), becquerelite (Ca(UO2)6O4(OH)6(H2O)8), and other uranyl hydroxide compounds commonly noted as alteration products during the corrosion of UO2, spent nuclear fuel, and naturally occurring uraninite. The results of these experiments demonstrate that significant quantities of both cerium (Kd=1020) and neodymium (Kd=840) are incorporated within the uranium alteration phases and suggest that ionic substitution and/or adsorption to the uranyl phases can play a key role in the limiting the mobility of REE (and by analogy, actinide elements) in a nuclear waste repository.

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.

A Comparison and Abstraction of the Commercial Spent Fuel Corrosion Experiments in the U.S.

ABSTRACTThis paper compares the results of three different fuel corrosion experiments by taking the existing corrosion data reported for various temperatures and recalculating corrosion rates at a single temperature of 25°C using a temperature dependent model developed elsewhere. Three types of light water reactor fuel corrosion tests (sometimes called dissolution or alteration tests) were performed in support of Yucca Mountain Project. The tests used three water contact modes and various fuel burnups. All measurements were adjusted for temperature differences and then compared. Five different isotopes (cesium, technetium, iodine, strontium, and, in the flow-through tests, uranium) were considered as a measure of corrosion. The data used represent over ten years of experiments with about nine different fuel types. Most experiments were with repository type fluids, containing dissolved constituents such as carbonate, calcium and silicon. The results show that all of the experiments predict similar fuel corrosion rates. Small differences in the isotope release rates are observed and incorporated in the abstracted uncertainty. Water contact mode (flow-through, batch, or drip) does not seem to be very important although the drip tests introduced larger variations. In developing a corrosion abstraction, all of the isotope measurements were considered equally. The distribution of release rates was used directly to develop the uncertainty. The mean corrosion rate was 1.8 × 10-4 fraction/year at 25°C (5%-95% range = 5.7 × 10-5 to 1.7 × 10-3). Using the derived corrosion rate for 25°C, and considering rapid axial splitting of the cladding, the CSNF fuel rod is expected to corrode in less than 2,000 years. The abstraction uses all the available experiments performed with water containing carbonates, silicon, or calcium and irradiated fuel to produce a corrosion rate distribution. Sensitivity studies using this corrosion rate abstraction in the TSPA-SR analysis show very small changes in dose (3%) in response to changes in the UO2 corrosion abstraction.

Effect of β-Radiation on the Non Irradiated UO2(s) Dissolution

ABSTRACTIn this work, we studied the effect of β radiation (Sr90 source with an activity of 7 mCi) on the dissolution of non irradiated UO2, as a chemicalanalogue of the spent nuclear fuel, in different leaching solutions.The experiments were carried out using a specifically designed continuous flow-through reactor and in nitrogen atmosphere to avoid oxygen contamination. The solid used was a non irradiated uranium dioxide with a particle size of 100–320 μm, which specific surface are was determined by the BET method. The experiments were carried out in NaCl media at different pH values. Both pH and redox potential of the solutions were continuously monitored. In all the cases, blank experiments were performed in parallel.Dissolution rates obtained under the effect of β radiation were compared with dissolution rates determined in the presence of hydrogen peroxide (the main oxidizing species radiolitically formed by the β radiation according to the CHEMSIMUL code) and with electrochemically determined UO2 corrosion rates found in the literature.

Investigations into speciation: Sorption and corrosion of actinides in systems of high ionic strength

It is a well known fact that Germany still favours the direct disposal of spent fuel elements in salt diapirs. Besides this spent fuel larger amounts of UO2, originating from enrichment plants as UF 6, transferred into (NH4)2UO2(CO3) 3 (AUC)and decomposed to UO 2 shall also be deposited in these salt domes: A hypothetical accident which must be taken into consideration is the intrusion of brine into these waste repositories. Therefore the corrosion of UO2 had to be investigated under various practical conditions and under the influence of a variety of parameters in order to obtain physicochemical data for modelling the dissolution and migration of the radionuclides contained in the various waste systems. Potentiostatic and potentiodynamic measurements which the corrosion rates were determined from had to be carried out not only by use of electrochemical but also by radiochemical methods in different systems at various temperatures and relevant potentials applied. Since many uranium species in the brines originating from the dissolution of the deposited material can easily be adsorbed at the corrosion products of cement, which is widely used for solidification of radioactive waste, investigations into the adsorption of these species on the cementitious materials had also to be performed.

The Corrosion of Spent UO2 Fuel in Synthetic Groundwater

AbstractLeaching of high burnup BWR fuel for up to 3 years showed that both U and Pu attain saturation rapidly at pH 8.1, giving values of 1–2 mg/l and 1 μg/l respectively. The leaching rate for Sr-90 decreased from about 10−5/d to 10−7/d but was always higher than the rates for U, Pu, Cm, Ce, Eu and Ru. Congruent dissolution was only attained at pH values of about 4.The hypothesis that alpha radiolytic decomposition of water is a driving force for UO2 corrosion even under reducing conditions has been examined in leaching tests on low burnup (low alpha dose-rate) fuel. No significant effect of alpha radiolysis under the experimental conditions was detected.