UO2 Samples Research Articles

Leaching of UO2 pellets doped with alpha-emitters (238/239Pu) in synthetic deep Callovian-Oxfordian groundwater

The reactivity of a polycrystalline UO2 surface under alpha irradiation in contact with groundwater is investigated, in the hypothesis of direct disposal of spent fuel in a deep geological repository. Two series of plutonium-doped UO2 samples (specific alpha activity of 18 and 385 MBq·g−1UO2) were leached in a synthetic Callovian-Oxfordian deep groundwater under anoxic conditions (Ar/CO2 3000 ppm, 3.5 bar relative pressure) to assess both the impact of alpha radiolysis of water and the complexing capacity of the groundwater ions on the dissolution of UO2. This study follows a prior one performed in pure and carbonated waters. Firstly, technical developments were necessary for the analyses in the groundwater solution because of its high salt concentrations: quantification limits were determined for the measurement of uranium and radiolytic H2O2 traces in this medium. Secondly, given the very high reactivity of these samples in the presence of air and in order to minimize any prior surface oxidation, a strict experimental protocol was followed, based on high-temperature annealing in Ar + 4% H2 with preleaching cycles. Each type of UO2 pellet was then leached under static conditions for 30 days (anoxic conditions, deep groundwater solutions). Results on the evolution of uranium releases are presented. For the lowest alpha activity (18 MBq·g−1UO2), uranium releases in groundwater were below the quantification limit of 2 × 10−8 mol·L−1 with a kinetic phosphorescence analyzer, even after 30 days. However, for higher alpha activity (385 MBq·g−1UO2) the uranium releases begin to exceed the quantification limit after 14 days of leaching and then increase exponentially. This increase is comparable to results previously obtained in carbonated solutions.

Helium release in uranium dioxide in relation to grain boundaries and free surfaces

Abstract Nuclear reaction analyses (NRA) based on the 3He(2H,4He)1H reaction were previously performed to follow the evolution of implanted 3He in polycrystalline UO2 samples. Experimental results pointed to an enhancement above 800 °C of the diffusion coefficient of helium over several microns in the vicinity of the grain boundaries, with respect to the diffusion coefficient within the grain. This was ascribed to the fact that grain boundaries are probably defect sinks which locally modify the defect concentrations. This study aims at demonstrating the particular effect of grain boundaries on helium migration. To this end, 3He implanted polycrystalline UO2 samples were cracked then annealed at 900 °C. Helium migration in the vicinity of the grain boundaries and near the crack was investigated by means of NRA microanalyses. Helium depletion extends over far larger distances in the vicinity of the grain boundaries than near the crack. Experimental evidence has been collected of the particular effect of grain boundaries on helium migration, which do not act as free surfaces at which helium atoms are simply released.

Relationship between changes in the crystal lattice strain and thermal conductivity of high burnup UO2 pellets

Abstract Two kinds of disk-shaped UO2 samples (4 mm in diameter and 1 mm in thickness) were irradiated in a test reactor up to about 60 and 130 GWd/t, respectively. The microstructures of the samples were investigated by means of optical microscopy, scanning electron microscopy/ electron probe micro-analysis (SEM/EPMA) and micro-X-ray diffractometry. The measured lattice parameters tended to be considerably smaller than the reported values, and the typical cauliflower structure which is often observed in high burnup fuel pellet is hardly seen in these samples. Thermal diffusivities of the samples were also measured by using a laser flash method, and their thermal conductivities were evaluated by multiplying the heat capacity of unirradiated UO2 and sample densities. While the thermal conductivities of sample 2 showed recovery after being annealed at 1500 K, those of sample 4 were not clearly observed even after being annealed at 1500 K. These trends suggest that the amount of accumulated irradiation-induced defects depends on the irradiation condition of each sample. From the comparison of the changes in the lattice parameter and strain energy density before and after the thermal diffusivity measurements, it is likely that the thermal conductivity recovery in the temperature region from 1200 to 1500 K is related to the migration of dislocation.

Effects of hydrogen and bromide on the corrosion of spent nuclear fuel andγ-irradiated UO2(s) in NaCl brine

AbstractRadiation induced UO2(s) corrosion is studied at elevated hydrogen pressure in NaCl brine containing traces of bromide. Release of Sr, Cs, Tc and actinides was measured in corrosion experiments with spent nuclear fuel pellets in presence of 10−2 mol H2(kg H2O)−1, and 10−4and 10−3 mol Br−(kg H2O)−1, respectively. For comparison, depleted UO2(s) pellets were γ-irradiated in NaCl brine at 10−3 mol H2(kg H2O)−1and 0−10−4 mol Br−(kg H2O)−1, respectively. In the γ-radiolysis experiments a significant increase in the yield of radiolytic products due to Br−is observed. Both, in the γ-radiolysis experiment with Br−and in that without Br−, the UO2(s) sample was oxidized, and the concentration of dissolved uranium was controlled by precipitation of meta-schoepite and clarkeite. In the spent nuclear fuel corrosion experiment under H2overpressure, aqueous concentrations of Tc and Np were in the range of solubilities of Tc(IV) and Np(IV) hydroxides, whereas measured U concentrations were between solubilities of U(VI) and U(IV) phases. The release rate of Sr was significantly increased in the presence of Br−traces. Results of the complementary spent nuclear fuel corrosion andγ-radiolysis experiments allow the conclusion that Br−traces reduce significantly the protective hydrogen effect with respect to the release of certain radionuclides and the yield of radiolytic products.

Growth mechanisms of interstitial loops in α-doped UO2 samples

New experimental size distributions are presented for interstitial-type dislocation loops in (U,Pu)O2 samples after 4 and 7 years of self-irradiation along with a new model. For this model, the work of Hayns has been extended to doped materials and to take into account additional phenomena, namely re-solution and coalescence as applied to gas bubble precipitation. The calculations are compared to the experimental size distributions obtained by means of Transmission Electron Microscopy for two different storage times. The role of re-solution and coalescence is discussed based on this model. This constitutes a basis for modelling the high burn-up structure (HBS) formation which is a consequence of the accumulation of radiation damage in nuclear fuels.

Photoneutron interrogation of low-enriched uranium induced by bremsstrahlung from a 4 MeV linac

Revealing smuggled nuclear material by passive γ-detection is hindered, because the weak radiation can easily be shielded. Neutrons, as penetrate shielding, represent a detection potential, by inducing fission in the nuclear material. A 4MeV linear accelerator was used as a pulsed neutron source for active interrogation of U-bearing material. Produced in heavy water by bremsstrahlung, neutrons subsequently induced fissions in UO2 samples. Delayed fission neutrons were detected in a neutron collar built up by 3He counters in a polyamide container. The counters were gated to be detached from high voltage during the electron pulse. Irradiation-measurement cycles were carried out with a 25Hz pulse repetition rate as optimum setting. The time analyser start-up was externally triggered and synchronised by the electron beam pulse. The response of the system was studied as a function of the intensity of the electron current, the amount of heavy water, U enrichment, and total U content. Sensitivity limit was achieved as 0.5g 235U and/or 30g 238U in a 20s measurement time (500 cycles) with the amount of heavy water of 100g and a mean electron current of 2μA. Because of the long decay time of the prompt (interrogating and fission) neutron pulse, about a half of the time interval (40ms) between pulses is only available for counting delayed neutrons.

Analysis of xenon gas inclusions in nuclear fuel using laser ablation ICP-MS

Analysis of fluid and gaseous inclusions in solids have been a major interest in various fields and have been carried out at different pressures, temperatures, and phase conditions. In nuclear fuel, approximately 20% of the fission products (FPs) are gaseous with isotopes of Xe contributing up to 90% to the product gases. However, previous to this work quantitative analysis of Xe inclusions in nuclear fuel samples have not been performed systematically. The method used must incorporate simple sample handling procedures in a shielded environment. This study is the development of a method for the direct determination of the fission gas (FG) products in micro inclusions contained in nuclear fuels using LA-ICP-MS. To determine the concentration of Xe in nuclear fuel, two calibration strategies were investigated. The first strategy was based on the direct injection of a known quantity of a reference gas into the LA-ICP-MS carrier gas system. Further, the ablation of a ‘matrix-matched’ standard of a non-irradiated UO2 sample, implanted with a known amount of 129Xe was also applied. Using these quantification methods, quantitative LA-ICP-MS measurements were performed on high burnup nuclear fuel. This study demonstrates that direct gas injection is most suitable for the quantification of fission gas in micron-sized inclusions. The direct gas addition is simple and linear calibration curves were obtained. Good reproducibility was obtained and matrix effects were within the uncertainty of the measurements. For the quantification of fission gases in nuclear fuel, aerosol particles were filtered before entering the ICP to remove interferences on the Xe isotopes from the solid FP matrix. The first quantitative determinations of the amount of gas in nuclear fuel using the direct injection method for calibration led to sample pressure calculations which were in good agreement with pressures estimated from computer simulations.

Thermal diffusion of chlorine in uranium dioxide

In a nuclear reactor, isotopes such as 35Cl present as impurities in the nuclear fuel are activated by thermal neutron capture. During interim storage or geological disposal of nuclear fuel, the activation products such as 36Cl may be released from the fuel to the geo/biosphere and contribute to the “instant release fraction” as they are likely to migrate in defects and grain boundaries. In order to differentiate diffusion mechanisms due to “athermal” processes during irradiation from thermally activated diffusion, both irradiation and thermal effects must be assessed. This work concerns the measurement of the thermal diffusion coefficient of chlorine in UO2.37Cl was implanted at a 1013 at/cm2 fluence in depleted UO2 samples which were then annealed in the 900−1200 °C temperature range and finally analyzed by Secondary Ion Mass Spectrometry (SIMS) to obtain 37Cl depth profiles. The migration process appears to be rather complex, involving mechanisms such as atomic, grain boundary, directed diffusion along preferential patterns as well as trapping into sinks before successive effusion. However, using a diffusion model based on general equation of transport, apparent diffusion coefficients could be calculated for 1000 and 1100 °C and a mean activation energy of 4.3 eV is proposed. This value is one of the lowest values compared to those found in literature for other radionuclides pointing out a great ability of chlorine to migrate in UO2 at relatively low temperatures. In order to unequivocally determine the diffusion behaviour of both implanted and pristine chlorine before and after thermal annealing, the structural environment of chlorine in UO2 was examined using micro X-ray fluorescence (micro-XRF) and micro X-ray absorption spectroscopy (micro-XAS).

Uranium dioxide interaction with groundwater – Leaching behavior and sorption tests

The solubility of three UO2 samples that were synthesized at different temperatures in reducing atmosphere and having different composition of surface oxidized layer was studied in distilled water and tuff groundwater (J-13). The solubility of the samples was governed by the presence of U(VI) in the surface oxidized layer. The sorption of Np(V) on those samples was accompanied by its partial reduction upon contact with UO2 (samples). For Np(IV) a partial oxidation was observed at pH values between 4.5 and 6.5. The most oxidized sample, with the highest U(VI) content in the surface layer, despite the highest solubility, was more efficient in the sorption of Np(V).

γ-Radiolysis of NaCl Brine in the Presence of UO2(s): Effects of Hydrogen and Bromide

AbstractConcentrated NaCl solution was γ-irradiated in autoclaves under a pressure of 25 MPa. A set of experiments were conducted in 6 mol (kg H2O)-1NaCl solution in presence of UO2(s) pellets; in a second set of experiments γ-radiolysis of the NaCl brine was studied without UO2(s). Hydrogen, oxygen and chlorate were formed as lonγ-lived radiolysis products. Due to the high external pressure, all radiolysis products remained dissolved. H2and O2reached steady state concentrations in the range of 5·10-3to 6·10-2mol (kg H2O)-1corresponding to gas partial pres-sure of ∼2 to ∼20 MPa. Radiolytic formation of hydrogen and oxygen increases with the concentration of bromide added to solution. Both, in the presence of bromide, resulting in a relatively high radiolytic yield, and in the absence of bromide surfaces of the UO2(s) samples were oxi-dized, and concentration of dissolved uranium reached the solubility limit of schoepite / NaUO2O(OH)(cr) transition. At the end of the experiments the pellets were covered by a surface layer of a secondary solid phase having a composition close to Na2U2O7. The experimental results demonstrate that bromide counteracts an H2inhibition effect on radiolysis gas production, even at a concentration ratio of [H2] / [Br-] > 100. The present observations are related to the competitive reactions of OH radicals with H2, Br-and Cl-. A similar competition of hydrogen and bromide, controlling the yield of γ-radiolysis products, is expected for solutions of lower Cl-concentration

Mechanisms governing the release of radionuclides from spent nuclear fuel in geological repository: major outcomes of the European Project SFS

ABSTRACTEuropean Commission supported a wide research project entitled “Spent Fuel Stability under repository conditions” (SFS) within the 5th FWP, the aim of which was to develop a common understanding of the radionuclides release from spent nuclear fuel in geological disposal and build a RN release model in order to assess the fuel performance. This project achieved by the end of 2004 focuses both on the Instant Release Fraction (IRF) model and the Matrix Alteration Model (MAM).A new IRF model was developed based on the anticipated performances of the various fuel microstructures (gap, rim, grains boundaries) and the potential diffusion of RN before the canister breaching. However, this model lets the choice to the end-user about the degree of conservativeness to consider.In addition, fuel alteration has been demonstrated to be linked to the production of radiolytic oxidants by water radiolysis at the fuel interface, the oxidation of the fuel interface by radiolytic oxidants and the subsequent release of uranium under the influence of aqueous ligands. A large set of experimental data was therefore acquired in order (i) to upgrade the current radiolytic kinetic scheme, (ii) to experimentally correlate the fuel alteration rate and the fuel specific alpha activity by performing experiments on alpha doped samples, (iii) to experimentally test the potential inhibitor effect of hydrogen on fuel dissolution. Based on these results, a new MAM was developed, which was also calibrated using the experiments on inactive UO2 samples. This model was finally applied to representative granitic, salt and clayey environment to predict spent fuel long-term fuel performance.

Effects of alpha self-irradiation on actinide-doped spent fuel surrogate matrix

ABSTRACTDuring the initial period after disposal, in which no mass transfer occurs with environmental materials, spent fuel undergoes modifications mainly arising from the effects of α self-irradiation damage. To simulate this aging phenomenon and estimate the consequences on the matrix, an experimental study was carried out in which old UO2 samples doped with short-lived actinides were characterized and analyzed. Changes that occur rapidly at the beginning of the aging process were identified in two lots of actinide-doped samples. Volume and microscopic swelling increased, reaching a maximum relative level of 1.5%. At the same damage level, microscopic swelling remained below 0.9%, suggesting that helium formation and atomic defects have a role in this regard. Similarly, the sample microhardness rapidly increased by up to more than 12%. Characterizing 238PuO2 sources showed at very high integrated and instantaneous doses, irradiation damage and helium accumulation at the grain boundaries resulted in grain decohesion or even powder formation. Helium mobility appeared to be modified by the presence of defects, and local implantation experiments were devised in the CEA's Pierre Süe Laboratory at Saclay to quantify this thermally enhanced diffusion component. The physical condition of the sintered PuO2 pellets after 30 years of storage revealed a purely athermal diffusion mechanism, depending on the sample α activity.

Testing of uranium dioxide enriched with 233U under reducing conditions

ABSTRACTExperiments have been conducted to determine an upper limit to the dissolution rate of UO2 under reducing conditions at great depths in granitic bedrock. The aqueous phase was a dilute, synthetic groundwater that simulates fresh groundwater conditions, so-called modified Allard groundwater. Four dissolution tests of long duration were performed. Test duration ranged from 52 to 140 days. Reducing conditions were produced with metallic iron. Average dissolution rates for the UO2 samples under the test conditions were preliminarily calculated based on the total recovery of U at the end of the tests. The dissolution rates ranged from 0.12 to 0.14 ppm/yr for the sum of the total release from the third and fourth cycles of testing (269 days). There was no indication of an effect of alpha radiolysis on the dissolution rate results for UO2 samples with 0, 5, and 10% 233U.

Simulation of Mixed-Oxide and Low-Enriched Uranium Fuel Burnup in a Pressurized Water Reactor and Validation Against Destructive Analysis Results

This work examines the capabilities of simulation codes to predict the concentration of nuclides in spent reactor fuel, in particular mixed-oxide (MOX) fuel, via comparisons with destructive radiochemical analyses performed on irradiated samples. We report on three MOX samples irradiated in a pressurized water reactor (PWR) and two UO2 samples irradiated in a different PWR. Actinide and fission-product concentrations were measured and were compared with concentration values obtained from simulation studies. The actinides include isotopes of uranium, neptunium, plutonium, americium, and curium. The fission products include isotopes of cesium, neodymium, samarium, europium, and gadolinium as well as 90Sr, 95Mo, 99Tc, 101Ru, 106Ru, 103Rh, 109Ag, 125Sb, 129I, and 144Ce. For many of the actinides, the predictions are quite good when compared with the measured values; but concentrations of some tend to be overpredicted. The cesium and neodymium, and some samarium concentrations, are well predicted, but some of the other fission products show variable results. The sensitivity of some of the results to sample-burnup estimates is discussed.

Variation of Solubility, Biokinetics and Dose Coefficient of Industrial Uranium Oxides According to the Specific Surface Area

The in vitro solubility, absorption to blood, lung retention and dose coefficient of industrial UO2 samples were studied as a function of the specific surface area (SSA) of the particles. An in vitro study has been carried out on two samples of industrial UO4 to compare the results with those obtained with UO2. Ten UO2 samples supplied by different fuel factories or research laboratories, presented specific surface areas from 1.00 to 4.45 m2.g-1. The wide range of values of SSA was due to the different conditions of fabrication. Dissolution tests in cell culture medium made on these ten samples have shown that the solubility increased 2.5-fold when the SSA increased 1.7-fold. The same tendency has been found for UO4, a soluble compound, and for U3O8, a moderately soluble compound. Four in vivo experiments carried out on rats by intratracheal instillation of dust suspensions of UO2, have highlighted the decrease in lung retention and the increase of absorption to blood with the SSA. The experimental absorption parameters calculated from the in vivo data allowed specific dose coefficients to be obtained which decreased from 6.6 to 4.3 µSv.Bq-1 when the SSA increased from 1.60 to 3.08 m2.g-1. Thus, the medical monitoring of workers at the workplace has to take into account any change in the fabrication process of the uranium compound which can affect the physiochemical properties and consequently the dose coefficient.

Measurement of uranium isotope ratios in solid samples using laser ablation and diode laser-excited atomic fluorescence spectrometry

A diode laser is used for the selective excitation of 235U and 238U in a laser-induced plasma applying Nd:YAG laser pulses to UO2 samples. The diode laser is rapidly scanned immediately following each laser sampling and the resonance atomic fluorescence spectrum for both isotopes is obtained on a pulse-to-pulse basis. Time-integrated measurements, with the diode laser fixed at either isotope, were also made. Optimum signal-to-noise was obtained at a distance of 0.8 cm from the sample surface, a pressure of 0.9 mbar and a Nd:YAG laser pulse energy of 0.5 mJ (880 MW cm−2). Three samples with 0.204, 0.407 and 0.714% 235U were measured. For example, for the UO2 pellet with the natural uranium isotopic composition (99.281% 238U and 0.714% 235U), the accuracy and precision were 7% and 5% (460 shots), respectively, limited by the continuum emission background from the laser-induced plasma.

Metallic ion production with electron cyclotron resonance ion source Caprice

In cooperation with CEN-Grenoble we developed a high temperature oven for the production of beams from rare and expensive nickel isotopes (62Ni and 64Ni) with an electron cyclotron resonance (ECR) ion source Caprice-14 GHz. For the first time, nickel has been produced in this way. For an experiment at the Unilac we have provided a stable Ni9+ beam for 40 days with a mean intensity of 25 eμA. The nickel consumption was a few mg per hour and ten ovens have been used for that period. Surprising and promising results were obtained with this oven for the production of uranium beams by using UO2 samples. These results are compared with the ‘‘plasma heating’’ method. Complementary investigations about the already studied ‘‘sputtering’’ method are also presented.

Premelting transition in uranium dioxide

Thermal analysis of the cooling curves of small samples of UO2±x —initially laser heated (in a high-pressure autoclave to inhibit evaporation) on a subsecond time scale to temperatures just below their melting points [T m(x)]—reveals, in the case of nominally stoichiometric UO2.00, a significant, λ-like, heat capacity [C p(T)] peak near 2670 K; the cooling curves of samples exposed to a reducing environment, on the other hand, exhibit undercooling, characteristic of a first-order phase transition, while under oxidizing conditions it is found that the premelting transition readily disappears. These findings confirm Bredig's original prediction of a premelting transition in this material, in common with that found in other (nonactinide) fluorites near 0.85T m. A simple model is presented in terms of which the observed behavior can be rationalized. The model is based on the hypothesis that the premelting transition is due to Frenkel disordering of the oxygen sublattice—a process which is rendered cooperative by attractive interactions between complementary Frenkel defects (oxygen interstitials and vacancies); these interactions are treated in a “mean-field” approximation. The quantitative degree of maximum disorder (realized just above the transition) is, on the other hand, controlled by repulsive interactions between like defects—the inclusion of which, solely through their effect on the configurational entropy, satisfactorily reproduces the values inferred from recent high-temperature neutron diffraction experiments. Assuming that the phase transition in stoichiometric UO2.00 is of second order, the model predicts a divergent heat capacity, C v, which approximates well to the experimental (λ-like) C p peak. Crucial to reproducing the observed behavior away from stoichiometry is the introduction of a (linear) dependence of the nonconfigurational partial entropy of formation on the prevailing concentration of intrinsic Frenkel defects in UO2±x; interestingly, it is found that the line of calculated (but unrealized) second-order transitions in UO2+x intersects the U4O9 phase boundary near to where a high-temperature diffuse order-disorder transition has been observed in the oxygen superlattice, suggesting that the second-order, λ-transition in UO2.00 is the stoichiometric counterpart of this transition in U4O9.

Luminescence Study of UO2+2 Binding to Immobilized Datura Innoxia Biomaterial

The spectrally and temporally resolved UO2+2 fluorescence signals have been measured for solid samples of UO2+2 -Datura, UO2+2-(immobilized Datura), and UO2+2-(silicate polymer) at liquid nitrogen temperature. The binding capacity of UO2+2 to Datura innoxia cell material has been enhanced significantly at pH 5 and 6 when immobilized in a polysilicate matrix. New binding sites having a greater binding strength with a lower availability have been observed for the binding of UO2+2 after immobilization of the cultured biomaterial. However, chemical alteration of the cell material resulting from this immobilization process has not been observed. The chemical environment of the binding sites responsible for the binding of UO2+2 in immobilized and free D. innoxia cell material has been demonstrated to be different. This difference results from the immobilization of the cell material within a polysilicate matrix.

Kinetically Controlled Dissolution of UO2(s) Under Oxidizing Conditions. A Combined Dissolution-Okidation Model.

ABSTRACTThe release of uranium from three samples of UO2(S) with different particle sizes (100–300 μm, 900–1100 μm, pellet) has been studied as a function of time. In all cases, the same pH, ionic medium, temperature, and oxygen partial pressure were used.Two distinctive trends were observed in the three cases, with a relatively fast initial uranium release followed, after 10–15 days, by a slower dissolution rate. The experiments were continued during 200 days; no change in the second dissolution rate was noticed.The uranium released as a function of time has been successfully fitted with a mathematical expression which combines an oxidation-dissolution mechanism.