Uranium Dioxide Matrix Research Articles

Modeling oxygen transport in Cr doped UO2 fuel with the TAF-ID during power transients

This paper proposes an efficient coupling between an oxygen redistribution model and the Thermodynamics for Advanced Fuels-International Database (TAF-ID) to describe the behavior of fission products, actinides, and dopant in a uranium dioxide matrix. The formulation of oxygen migration in the fuel is a function of the oxygen (and uranium) chemical potential gradient. The proposed formulation does not require the introduction of the Soret term and relies solely on the known mobilities of uranium and oxygen in the fuel. This coupling is applied to the simulation of a power ramp on a chromia-doped fuel rod using the PLEIADES/ALCYONE fuel performance code. The simulations successfully reproduce most of the available experimental observations on the fission products (Cs, Mo, Ru) and on the chromium dopant.

Evolution of pressurized xenon bubble and response of uranium dioxide matrix: A molecular dynamics study

To explore the microstructure evolution of uranium dioxide (UO2) due to accumulation of xenon (Xe) atoms at high burnup, the growth of highly pressurized Xe bubbles has been simulated in molecular dynamics (MD) by sequentially inserting Xe into a pre-existing equilibrium Xe bubble in UO2 using five different interatomic potentials at 1500 K. The results reveal that the characteristics of Xe bubbles and the microstructure evolution of UO2 due to the bubble growth can be divided into two stages in terms of the maximum pressure (Pmax) of bubbles. Before the bubble reaches Pmax (stage I), the characteristics of bubbles and the microstructure evolution of UO2 are relatively independent of the interatomic potentials used. The bubble volume and pressure increase with inserting additional Xe atoms, and the Xe atoms in the bubble evolve from an initial gaseous state to a glassy/amorphous state, a nearly fcc solid structure and subsequently a high density amorphous or glassy state. Moreover, the high density, over-pressurized bubbles displace U and O atoms, creating a larger volume to accommodate Xe atoms and correspondingly produce self-interstitial atoms (SIAs) with increasing Xe atoms. Most of the SIAs within stage I are distributed at the center of six {100} facets around the Xe bubble. However, after the bubble pressure reaches Pmax (stage II), the bubble characteristics predicted by MD simulations depend on both the UO2 interatomic potential and the Xe-UO2 potential. However, the microstructure evolution of UO2 is mainly determined by the UO2 potential since it determines the mobility of self-interstitial defects and the stability of the UO2 fluorite structure. ½<110> dislocations (loops) or UO2 phase transitions along with incipient boundaries (or interfaces) appear due to the accumulation of Xe in over-pressurized bubbles.

Direct observations of Pd–Te compound formation within noble metal inclusions in spent nuclear fuel

Although the existence of a five-metal (Mo-Tc-Ru-Rh-Pd) phase – as nanoparticles observed in irradiated nuclear fuel – has been known for more than half a century, the chemical and physical mechanisms controlling the formation and behavior of such particles remain stubbornly elusive. We present in this work new evidence for the presence of a separate nonmetallic phase associated with the metallic particles and containing a significant fraction of Te in addition to Pd. While this new phase potentially complicates the thermodynamic picture of a mixed alloy in equilibrium with the surrounding fuel environment, it also provides new clues in the search for a chemical mechanism for Pd migration through the uranium dioxide matrix and the nucleation behavior of the particles. Fractionation between phases may subsequently affect the mechanical performance of fuels during irradiation and their interactions with the surrounding environment during long-term waste storage.

Post-irradiation examinations of [formula omitted] composites as part of the Accident Tolerant Fuels Campaign

Enhanced thermal conductivity uranium dioxide composites containing silicon carbide (UO2−SiC) and diamond (UO2-diamond) have been irradiated to low burnup. The conditions of this irradiation test and subsequent postirradiation examinations are discussed. These irradiations evaluate fuel microstructure and potential fuel cladding interaction of UO2 composites, which have been proposed as accident tolerant fuel candidates.Both non-destructive and destructive techniques have been used to evaluate fuel integrity, fission gas release, fission product distribution, burnup, fuel swelling and cladding strain. Examination of the UO2-SiC pellets revealed enhanced cracking when compared to UO2 pellets irradiated under similar conditions. Instability of the SiC whiskers in the uranium dioxide matrix was observed in the pellet central region, where the local temperatures exceeded 1300°C. The microstructure of the UO2-diamond was severely disrupted during irradiation, resulting in local migration of cesium along the fuel stack and increased fission gas release when compared with the expected release from the Vitanza curve at corresponding values of burnup and irradiation temperature.The postirradiation examination results cast doubt on the suitability of these additives to improve UO2 fuel performance in a way that would lead to enhanced accident tolerance.

Powder Leaching Study for Grain Boundary Inventory of Two High Burnup Fuels

ABSTRACTIn the context of safety assessment, the fraction of inventory that is expected to rapidly dissolve when water contacts the spent fuel is called the Instant Release Fraction (IRF). Conceptually, this fraction consists of radionuclides outside of the uranium dioxide matrix and therefore the fraction can be further divided into the radionuclides in the fuel/cladding gap and radionuclides in the grain boundaries. The relative importance of these two fractions is investigated here for two Swedish high burnup fuels through simultaneous grinding and leaching fuel fragments in simplified groundwater for a short period of time. The hypothesis is that this will expose grain boundaries to leaching solution and provide an estimate of the release of the grain boundary inventory upon contact with water. The studied fragments were used in previous leaching experiments and thus pre-washed to remove any pre-oxidized phases. The results showed a significant release of iodine, cesium and rubidium and to a lower extent molybdenum and technetium. The fraction of inventory in the aqueous phase of actinides and lanthanides was 1-2 orders of magnitude lower than for the elements associated to the IRF. Both fuels displayed a very similar behavior and no correlation as a function of burnup or fission gas release was found.

High Useful Yield and Isotopic Analysis of Uranium by Resonance Ionization Mass Spectrometry.

Useful yields from resonance ionization mass spectrometry can be extremely high compared to other mass spectrometry techniques, but uranium analysis shows strong matrix effects arising from the tendency of uranium to form strongly bound oxide molecules that do not dissociate appreciably on energetic ion bombardment. We demonstrate a useful yield of 24% for metallic uranium. Modeling the laser ionization and ion transmission processes shows that the high useful yield is attributable to a high ion fraction achieved by resonance ionization. We quantify the reduction of uranium oxide surface layers by Ar+ and Ga+ sputtering. The useful yield for uranium atoms from a uranium dioxide matrix is 0.4% and rises to 2% when the surface is in sputter equilibrium with the ion beam. The lower useful yield from the oxide is almost entirely due to uranium oxide molecules reducing the neutral atom content of the sputtered flux. We demonstrate rapid isotopic analysis of solid uranium oxide at a precision of <0.5% relative standard deviation using relatively broadband lasers to mitigate spectroscopic fractionation.

Laser-induced breakdown spectroscopy of light water reactor simulated used nuclear fuel: Main oxide phase

The analysis of light water reactor simulated used nuclear fuel using laser-induced breakdown spectroscopy (LIBS) is explored using a simplified version of the main oxide phase. The main oxide phase consists of the actinides, lanthanides, and zirconium. The purpose of this study is to develop a rapid, quantitative technique for measuring zirconium in a uranium dioxide matrix without the need to dissolve the material. A second set of materials including cerium oxide is also analyzed to determine precision and limit of detection (LOD) using LIBS in a complex matrix. Two types of samples are used in this study: binary and ternary oxide pellets. The ternary oxide, (U,Zr,Ce)O2 pellets used in this study are a simplified version the main oxide phase of used nuclear fuel. The binary oxides, (U,Ce)O2 and (U,Zr)O2 are also examined to determine spectral emission lines for Ce and Zr, potential spectral interferences with uranium and baseline LOD values for Ce and Zr in a UO2 matrix. In the spectral range of 200 to 800nm, 33 cerium lines and 25 zirconium lines were identified and shown to have linear correlation values (R2) >0.97 for both the binary and ternary oxides. The cerium LOD in the (U,Ce)O2 matrix ranged from 0.34 to 1.08wt% and 0.94 to 1.22wt% in (U,Ce,Zr)O2 for 33 of Ce emission lines. The zirconium limit of detection in the (U,Zr)O2 matrix ranged from 0.84 to 1.15wt% and 0.99 to 1.10wt% in (U,Ce,Zr)O2 for 25 Zr lines. The effect of multiple elements in the plasma and the impact on the LOD is discussed.

The melting behaviour of uranium/neptunium mixed oxides

The melting behaviour in the pseudo-binary system (UO2+NpO2) has been studied experimentally for the first time in this work with the help of laser heating under controlled atmosphere. It has been observed that the solidus and liquidus lines of this system follow an ideal solution behaviour (negligible mixing enthalpy) between the well-established solid/liquid transition temperatures of pure UO2 (3130K) and that recently assessed for NpO2 (T=3070K). Pre- and post-melting material characterizations performed with the help of X-ray diffraction and Raman spectroscopy are also consistent with ideal mixing of the two end members. Such behaviour follows the similar structure and bonding properties of tetravalent uranium and neptunium and the similar melting points of the two oxides. The interest of this investigation is twofold. From a technological viewpoint, it indicates that the incorporation of NpO2 in UO2 fuel or transmutation targets is a viable option to recycle neptunium without inducing any relevant change in the chemical or thermal stability of the uranium dioxide matrix, even up to the melting point. From a more fundamental perspective, it confirms that actinide dioxides, and particularly UO2, tend to mix in a way closer to ideal, the closer are the atomic numbers, 5-f electron shell filling, atomic radii and oxygen potentials of the metals forming the pure dioxides.

Curium analysis in plutonium uranium mixed oxide by x-ray fluorescence and absorption fine structure spectroscopy

Plutonium uranium mixed oxide (MOX) fuels are being used in commercial nuclear reactors. The actinides in these fuels need to be analyzed after irradiation for assessing their behaviour with regards to their environment and the coolant. In this work the study of the local occurrence, speciation and next-neighbour environment of curium (Cm) in the (Pu,U)O2 lattice within an irradiated (60MWdkg−1 average burn-up) MOX sample was performed employing micro-x-ray fluorescence (µ-XRF) and micro-x-ray absorption fine structure (µ-XAFS) spectroscopy. The chemical bonds, valences and stoichiometry of Cm (∼0.7wt% in the rim and ∼0.03wt% in the centre) are determined from the experimental data gained for the irradiated fuel material examined in its centre and peripheral (rim) zones of the fuel. Curium occurrence is also reduced from the centre (hot) to the periphery (colder) because of the condensation of these volatile oxides. In the irradiated sample Cm builds up as Cm3+ species (>90%) within a [CmO8]13− or [CmO7]11− coordination environment and no (<10%) Cm(IV) can be detected in the rim zone. Curium dioxide is reduced because of the redox buffering activity of the uranium dioxide matrix and of its thermodynamic instability.

Americium characterization by X-ray fluorescence and absorption spectroscopy in plutonium uranium mixed oxide

Plutonium uranium mixed oxide (MOX) fuels are currently used in nuclear reactors. The actinides in these fuels need to be analyzed after irradiation for assessing their behaviour with regard to their environment and the coolant. In this work the study of the atomic structure and next-neighbour environment of Am in the (Pu,U)O₂ lattice in an irradiated (60 MW d kg⁻¹) MOX sample was performed employing micro-X-ray fluorescence (µ-XRF) and micro-X-ray absorption fine structure (µ-XAFS) spectroscopy. The chemical bonds, valences and stoichiometry of Am (~0.66 wt%) are determined from the experimental data gained for the irradiated fuel material examined in its peripheral zone (rim) of the fuel. In the irradiated sample Am builds up as Am³⁺ species within an [AmO₈]¹³⁻ coordination environment (e.g. >90%) and no (<10%) Am(IV) or (V) can be detected in the rim zone. The occurrence of americium dioxide is avoided by the redox buffering activity of the uranium dioxide matrix. - Graphical abstract: Americium LIII XAFS spectra recorded for the irradiated MOX sub-sample in the rim zone for a 300 μm×300 μm beam size area investigated over six scans of 4 h. The records remain constant during multi-scan. The analysis of the XAFS signal shows that Ammore » is found as trivalent in the UO₂ matrix. This analytical work shall open the door of very challenging analysis (speciation of fission product and actinides) in irradiated nuclear fuels. - Highlights: • Americium was characterized by microX-ray absorption spectroscopy in irradiated MOX fuel. • The americium redox state as determined from XAS data of irradiated fuel material was Am(III). • In the sample, the Am³⁺ face an AmO₈¹³⁻coordination environment in the (Pu,U)O₂ matrix. • The americium dioxide is reduced by the uranium dioxide matrix.« less

Effect of the cascade energy on defect production in uranium dioxide

The primary damage induced by a displacement cascade in a pure uranium dioxide matrix was investigated using classical molecular dynamics simulations. Cascades were initiated by accelerating a uranium primary knock-on atom (PKA) to a kinetic energy ranging from 1keV to 80keV inside a perfect UO2 lattice at low temperature (300K and 700K). There is little effect of temperature in the temperature range studied. Following the cascade event, the damage level, defined as the total number of defects irrespective of whether they form clusters or not, is proportional to the initial kinetic energy of the PKA, in agreement with the literature relating to other materials. The linear dependence of damage upon initial PKA energy results from the formation of subcascades at high energy and constitutes a simple law which can be applied to any material and used in order to extrapolate molecular dynamics results to high energy PKAs. The nature of irradiation induced defects has also been studied as a function of the cascade energy.

Plutonium–uranium mixed oxide characterization by coupling micro-X-ray diffraction and absorption investigations

Plutonium–uranium mixed oxide (MOX) fuels are currently used in nuclear reactors. The potential differences of metal redox state and microstructural developments of the matrix before and after irradiation are commonly analysed by electron probe microanalysis. In this work the structure and next-neighbor atomic environments of Pu and U oxide features within unirradiated homogeneous MOX and irradiated (60 MW d kg −1) MOX samples was analysed by micro-X-ray fluorescence (μ-XRF), micro-X-ray diffraction (μ-XRD) and micro-X-ray absorption fine structure (μ-XAFS) spectroscopy. The grain properties, chemical bonding, valences and stoichiometry of Pu and U are determined from the experimental data gained for the unirradiated as well as for irradiated fuel material examined in the center of the fuel as well as in its peripheral zone (rim). The formation of sub-grains is observed as well as their development from the center to the rim (polygonization). In the irradiated sample Pu remains tetravalent (>95%) and no (<5%) Pu(V) or Pu(VI) can be detected while the fuel could undergo slight oxidation in the rim zone. Any slight potential plutonium oxidation is buffered by the uranium dioxide matrix while locally fuel cladding interaction could also affect the redox of the fuel.

On the solution and migration of single Xe atoms in uranium dioxide – An interatomic potentials study

Atomic scale simulation techniques based on empirical potentials have been considered in the present work to get insight on the behaviour of single Xe atoms in the uranium dioxide matrix. In view of the high activation energies commonly observed for Xe migration, this work has focused on the so-called “static calculations” (i.e. energy minimization based calculation) of incorporation and migration energies of Xe in UO 2, using empirical interatomic potentials to describe atom interactions. A detailed study of these results enables to determine the solution and the migration properties of Xe in the different stoichiometry regimes, and can be applied as well for the in-pile behaviour of xenon.

Technetium and Molybdenum in Oxide Spent Nuclear Fuel: Impact on Release Estimates

AbstractTechnetium-99 99Tc) is an important radionuclide in repository models owing to its relatively long half-life and the high aqueous solubility of compounds where it is inthe heptavalent state. The vast majority of the 99Tc inventory presently slatedfor disposal is contained in oxide commercial spent nuclear fuel (CSNF), where it is divided (along with its cohort, molybdenum) betweenexsolved, intermetallic “epsilon” particles and isolated, likely oxidized, atoms distributed in the uranium dioxide matrix. We present recent evidence from synchrotron x-ray fluorescence spectroscopy, electron microscopy, and fuel dissolution testing on the likely oxidation state, coordination environment, and physical disposition of technetium and molybdenum in CSNF. Effects of the relative proportioningof technetium and molybdenum among the metallic and oxidized states in CSNF, andtheir distribution in or near grain boundaries and gaps on release during CSNF corrosion testing are discussed.

Thermal diffusion of Helium and volatil fission products in UO2 and zirconolite nuclear ceramics

AbstractThe behaviour and diffusion mechanisms of helium in nuclear ceramics, such as uranium dioxide spent fuel matrix and zirconolite for the specific conditioning of minor actinides, significantly impact the possible evolution of those matrices in interim storage or disposal conditions. In the framework of spent fuel storage studies, the additional diffusion of gas and fission products in uranium dioxidematrix is also an essential aspect of the R&amp;D. Specific experimental studies have been conducted, devoted to thermal diffusion under 1000 C. Data processing methods, lead to helium diffusion coefficient and associated activation energy of 1.05 eV in the zirconolite and 2 eV in UO2. Comparativelywith the uranium dioxide matrix, the helium diffusion coefficient in zirconolite is 1 to 100 million times higher; this parameter will have to be taken into account to dimension the waste form. Diffusion coefficients measurements between 800 C and 1000 C, investigated by SIMS, showed a very slow diffusion of volatile fission products Xe, I, Te and Cs, with coefficients two or three order of magnitude lower than for helium.

Leaching behaviour of UO2 containing α-emitting actinides

After a few hundred years in a geologic repository, α-emissions will dominate the radiation field in and around spent nuclear fuel. In the event of a failure of the spent fuel container, the dissolution of the uranium dioxide matrix in contact with groundwater could be enhanced by α-radiolysis, which could create oxidizing conditions at the fuel surface. The radioactivity (hence the radiolysis) of the irradiated fuel available nowadays is characterized by strong β- and γ-contributions, which are not representative of aged fuel in a repository. A possible way to single out the effects of α-radiolysis on the dissolution behaviour of irradiated fuel is to study unirradiated UO2 doped with a short-lived α-emitter. UO2 containing ~0.1 and ~10 wt.% of 238PuO2 was fabricated and leached. Previous results of static (batch and sequential) leaching tests on monoliths at room temperature in demineralized water under anoxic atmosphere showed that the amounts of U released were higher in the case of UO2 containing 238Pu than in the case of undoped UO2. This work presents results of leaching experiments performed under similar conditions on materials of the same composition, but crushed to give samples with two different particle size distributions to investigate the effects of different surface areas. The U release data, normalized to the geometric surface area, showed enhanced U dissolution from α-doped UO2 also in the case of crushed materials, except for the initial release after 1 h of leaching. Under these experimental conditions, the concentration of U measured in the leachates for the two doped materials did not show a clear dependence on α-activity.

Refractory Oxide-Metal Composites: Scanning Electron Microscopy and X-ray Diffraction of Uranium Dioxide-Tungsten

Unidirectional solidification of a melt, consisting of uranium dioxide and 5 to 15 percent by weight of tungsten, formed well-ordered arrangements of tungsten fibers or platelets in the uranium dioxide matrix. Fiber shape and orientation relations of the components indicate the development of several dominant growth modes.

The emission-spectrographic determination of impurities in uranium-233 dioxide after pre-concentration of the rare-earth elements by ion exchange

Twenty-nine impurity elements were measure directly in uranium-233 dioxide by the carrier-distillation method. Enhancement of some of the impurity elements (magnesium, bismuth, antimony, lead, manganese, zinc, cadmium, and indium) in the uranium dioxide matrix is shown. The precision of the carrier- distillation method was improved by using internal standard control. Rare-earth elements were determined by a spectrographic method after concentration by ion exchange, giving a total of 45 impurity elements determined for each sample. The uranium-233 is recoverable after the analysis. (auth)