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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

A molecular dynamics survey of grain boundary energy in uranium dioxide and cerium dioxide

Uranium dioxide (UO 2 ) is the primary fuel material that is used in current nuclear reactors. As one of the most fundamental material parameters, grain boundary (GB) energy strongly influences many fuel properties, and the influences depend on the characters and properties of individual GBs. Using molecular dynamics simulations, a high throughput survey of GB energy in UO 2 was carried out for the purpose of elucidating the roles of GB geometry such as misorientation and inclination, as well as the bonding nature of UO 2 , in affecting GB energy. GB energies in CeO 2 were calculated as well for comparison with UO2toinvestigate the generality of GB energy anisotropy in fluorite phase oxides. The results show significant GB energy anisotropy in both UO 2 and CeO 2 that is associated with the cubic symmetry of the fluorite structure. More interestingly, the GB anisotropy is found to be dependent not only on the crystal structure but also the ionic bonding. As such, the GB energy anisotropy in fluorite oxides has significant differences compared with that in fcc metals. We report that the data obtained and the increased knowledge on GB anisotropy will facilitate GB engineering for nuclear fuels with improved properties

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Transverse Rupture Strength of Uranium Dioxide

Uranium dioxide (UO 2 ) fuel is used as fuel in light water reactors (LWRs). While the fuel pellet is technically the first engineering barrier for radionuclide release, pellet fracturing at intermediate- to high-burnup values releases fission gases into the fuel rod plenum [1, 2]. Therefore, the true engineering barrier is the fuel cladding, which performs very well in LWR environments [3]. The extreme temperature gradients generated by fission energy and the low thermal conductivity of UO 2 quickly induce radial cracking in UO 2 during operation [4]. Cracks in the fuel provide opportunities for fuel relocation, increased fission gas release, and pellet-cladding mechanical interaction (PCMI) [5]. The ability to predict and engineer the fracture of UO 2 fuel pellets using modern computational tools is therefore a key engineering goal that has been the focus of ongoing experimental and computational efforts [6, 7]. Accurate predictions of fuel pellet cracking during operation requires knowledge of more complex phenomena, but improved understanding of the fundamental fracture behavior of unirradiated UO 2 is first necessary.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Building a DFT+U machine learning interatomic potential for uranium dioxide

Despite uranium dioxide (UO 2 ) being a widely used nuclear fuel, fuel performance models rely extensively on empirical correlations of material behavior, leveraging the historical operating experience of UO 2 . Mechanistic models that consider an atomistic understanding of the processes governing fuel performance (such as fission gas release and creep) will enable a better description of fuel behavior under non-prototypical conditions such as in new reactor concepts or for modified UO 2 fuel compositions. To this end, molecular dynamics simulation is a powerful tool for rapidly predicting physical properties of proposed fuel candidates. However, the reliability of these simulations depends largely on the accuracy of the atomic forces. Traditionally, these forces are computed using either a classical force field (FF) or density functional theory (DFT). While DFT is relatively accurate, the computational cost is burdensome, especially for f-electron elements, such as actinides. By contrast, classical FFs are computationally efficient but are less accurate. For these reasons, we report a new accurate machine learning interatomic potential (MLIP) for UO 2 that provides high-fidelity reproduction of DFT forces at a similar low cost to classical FFs. We employ an active learning approach that autonomously augments the DFT training data set to iteratively refine the MLIP. To further improve the quality of our predictions, we utilize transfer learning to retrain our MLIP to higher-accuracy DFT+U data. We validate our MLIPs by comparing predicted physical properties (e.g., thermal expansion and elastic properties) with those from existing classical FFs and DFT/DFT+U calculations, as well as with experimental data when available.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Effect of Added Gadolinium Oxide on the Thermal Air Oxidation of Uranium Dioxide

To develop a more reliable and stable UO2-based nuclear fuel, the Pacific Northwest National Laboratory (PNNL) investigated modifying fuel with several soluble lanthanides and zirconium. This article provides the results of these studies investigating gadolinium doping at levels up to 10 mass%. The authors characterized and compared commercially- and PNNL-prepared gadolinium-doped UO2 to determine the oxygen-to-metal ratio, elemental distribution, chemical composition, physical appearance, lattice parameters, and grain structure using atomic force microscopy, scanning electron microscopy coupled with energy dispersive spectroscopy, and X-ray diffractometry. After confirming PNNL-prepared UO2 and (Gd,U)O2 were similar to commercially prepared UO2 and (Gd,U)O2, we measured the thermal behavior of these gadolinium-doped UO2 materials to air oxidation using differential scanning calorimetry and thermogravimetric analysis. Addition of gadolinium stabilized the first oxidation product U4¬O9/U3O7 and slowed the subsequent oxidation to U3O8. Comparison of our measured two-step oxidation of UO2 to U3O8 at 325°C to common gas/solid kinetic reaction models found that each oxidation step is best described as a convolution of kinetic behaviors; the gadolinium insertion into the UO2 lattice enforces a significant alteration in mechanism and oxidation rate. Noticeable changes in mechanism become apparent between 1% and 3% gadolinium content.

Uranium Dioxide, Gadolinium Oxide, Thermal Air Oxi↗
NASA NTRS1966

Stabilization of Tungsten-Uranium Dioxide Composites Under Thermal Cycling Conditions

Uranium losses during thermal cycling of tungsten - uranium dioxide composites to 2500° C in flowing purified hydrogen for specimens initially containing 35 volume percent uranium dioxide were reduced by the use of thorium dioxide, calcium oxide, or yttrium oxide as additives in solid solution with the uranium dioxide. The effectiveness of the additives decreased in the order yttrium oxide, calcium oxide, and thorium dioxide. Stabilization of uranium dioxide by calcium oxide or yttrium oxide can be explained in part on the basis of the introduction of oxygen vacancies into the fluorite lattice and the associated lowered partial molar free energy of oxygen. The difference in the effectiveness of calcium oxide and yttrium oxide is discussed, and a possible explanation of the observed behavior is proposed. Of the concentrations of yttrium oxide tested (i.e., 2. 5, 5, and 10 mole percent), 10 mole percent of yttrium oxide was the most effective in pre­ venting loss of uranium from composities.

Gluyas, Richard E.↗
NASA NTRS1965

Fuel-Retention Properties of Tungsten-Uranium Dioxide Composites

Although tungsten-uranium dioxide composites appear very promising, as fuel element materials for high-temperature nuclear rocket reactors, one of the major problems with these materials is the loss of fuel at reactor operating temperatures (near or above 4500 F). Substantial fuel losses occur by vaporization whenever the uranium dioxide is directly exposed to elevated temperature environments and by fuel migration and vaporization when reactor operating conditions that involve thermal cycling are simulated. Several methods of minimizing fuel loss have been evaluated and appear quite promising.

Gedwill, Michael A.↗
DOE OSTI2024

Formation of uranium oxy-carbide and uranium carbide via conversion of polymer covered uranium dioxide by laser-based thermal processing

Conventional formation of carbonaceous uranium compounds requires bulk processing using furnace-based approaches. Here, a methodology employing polymer covered uranium dioxide and laser-based heating is explored to enable rapid, localized formation of carbonaceous uranium compounds. Specifically, heating of poly(methyl methacrylate) covered uranium dioxide powder to high temperatures using laser irradiation in argon and methane gaseous environments was investigated. Decomposition of material and reactions induced by laser irradiation were probed in situ by residual gas analysis using a benchtop mass spectrometer. In this study, to determine the effect on the resultant material phase, three different process parameters were varied: gaseous atmosphere, laser power, and laser irradiation time. Material processed under varying conditions was analyzed using powder X-ray diffraction and scanning electron microscopy. This work realized the conversion of uranium dioxide into uranium oxy-carbide(s) and uranium carbide(s) phases, at over 60 wt.%, via the polymer surface application and laser-based thermal decomposition methodology.

36 MATERIALS SCIENCE↗
DOE OSTI2023

The effects of radiation-induced grain subdivision and dislocations on the fracture properties of uranium dioxide

Microcantilever bending tests were applied on uranium dioxide samples irradiated by 84 MeV Xe ions. Depending on the initial grain sizes, different levels of grain subdivision and dislocation development were observed near the irradiated surface. These radiation damage features were shown to degrade the fracture properties of both samples. The fundamental aspects of the radiation-microstructure-property relationships were discussed in this paper, taking a holistic consideration of the microcantilever bending data, fractography and microstructures. The separate effects of the irradiation-induced grain subdivision and dislocations on the fracture initiation and propagation of uranium dioxide were discussed. In conclusion, it was observed that radiation-induced dislocations and grain subdivision, without the presence of Xe bubbles co-located with the defects, cause reductions in the Young's modulus, fracture strain, and fracture stress, and a transition to intergranular fracture.

36 MATERIALS SCIENCE↗
NASA NTRS1964

Fuel Retention Improvement at High Temperatures in Tungsten-Uranium Dioxide Dispersion Fuel Elements by Plasma-Spray Cladding

An investigation was undertaken to determine the feasibility of depositing integrally bonded plasma-sprayed tungsten coatings onto 80-volume-percent tungsten - 20-volume-percent uranium dioxide composites. These composites were face clad with thin tungsten foil to inhibit uranium dioxide loss at elevated temperatures, but loss at the unclad edges was still significant. By preheating the composite substrates to approximately 3700 degrees F in a nitrogen environment, metallurgically bonded tungsten coatings could be obtained directly by plasma spraying. Furthermore, even though these coatings were thin and somewhat porous, they greatly inhibited the loss of uranium dioxide. For example, a specimen that was face clad but had no edge cladding lost 5.8 percent uranium dioxide after 2 hours at 4750 dgrees F in flowing hydrogen. A similar specimen with plasma-spray-coated edges, however, lost only 0.75 percent uranium dioxide under the same testing conditions.

Grisaffe, Salvatore J.↗
DOE OSTI2022

Capturing the ground state of uranium dioxide from first principles: Crystal distortion, magnetic structure, and phonons

Uranium dioxide (UO 2 ) remains a formidable challenge for first-principles approaches due to the complex interplay among spin-orbit coupling, Mott physics, magnetic ordering, and crystal distortions. Here we use DFT+U to explore UO 2 at zero temperature, incorporating all the aforementioned phenomena. The technical challenge is to navigate the many metastable electronic states produced by DFT+U, which is accomplished using f-orbital occupation matrix control to search for the ground state. We restrict our search to the high-symmetry ferromagnetic phase, including spin-orbit coupling, which produces a previously unreported occupation matrix. This newfound occupation matrix is then used as an initialization to explore the broken symmetry phases. We find the oxygen cage distortion of the 3k antiferromagnetic state to be in excellent agreement with experiments, and both the spin-orbit coupling and the Hubbard U are critical ingredients. Further, we demonstrate that only select phonon modes have a strong dependence on the Hubbard U, whereas magnetic ordering has only a small influence overall. We perform measurements of the phonon dispersion curves using inelastic neutron scattering, and our calculations show good agreement when using reasonable values of U. The quantitative success of DFT+U warrants exploration of thermal transport and other observables within this level of theory.

36 MATERIALS SCIENCE↗
NASA NTRS2013

Design of a Uranium Dioxide Spheroidization System

The plasma spheroidization system (PSS) is the first process in the development of tungsten-uranium dioxide (W-UO2) fuel cermets. The PSS process improves particle spherocity and surface morphology for coating by chemical vapor deposition (CVD) process. Angular fully dense particles melt in an argon-hydrogen plasma jet at between 32-36 kW, and become spherical due to surface tension. Surrogate CeO2 powder was used in place of UO2 for system and process parameter development. Particles range in size from 100 - 50 microns in diameter. Student s t-test and hypothesis testing of two proportions statistical methods were applied to characterize and compare the spherocity of pre and post process powders. Particle spherocity was determined by irregularity parameter. Processed powders show great than 800% increase in the number of spherical particles over the stock powder with the mean spherocity only mildly improved. It is recommended that powders be processed two-three times in order to reach the desired spherocity, and that process parameters be optimized for a more narrow particles size range. Keywords: spherocity, spheroidization, plasma, uranium-dioxide, cermet, nuclear, propulsion

Cavender, Daniel P.↗
DOE OSTI2021

Hyperstoichiometric Uranium Dioxides: Rapid Synthesis and Irradiation-Induced Structural Changes

Uranium dioxide (UO 2 ), the primary fuel for commercial nuclear reactors, incorporates excess oxygen forming a series of hyperstoichiometric oxides. Thin layers of these oxides, such as UO 2.12 , form readily on the fuel surface and influence its properties, performance, and potentially geologic disposal. This work reports a rapid and straightforward combustion process in uranyl nitrate–glycine–water solutions to prepare UO 2.12 nanomaterials and thin films. We also report on the investigation of the structural changes induced in the material by irradiation. Despite the simple processing aspects, the combustion synthesis of UO 2.12 has a sophisticated chemical mechanism involving several exothermic steps. Raman spectroscopy and single-crystal X-ray diffraction (XRD) measurements reveal the formation of a complex compound containing the uranyl moiety, glycine, H 2 O, and NO 3 – groups in reactive solutions and dried combustion precursors. Combustion diagnostic methods, gas-phase mass spectroscopy, differential scanning calorimetry (DSC), and extracted activation energies from DSC measurements show that the rate-limiting step of the process is the reaction of ammonia with nitrogen oxides formed from the decomposition of glycine and uranyl nitrate, respectively. However, the exothermic decomposition of the complex compound determines the maximum temperature of the process. In situ transmission electron microscopy (TEM) imaging and electron diffraction measurements show that the decomposition of the complex compound directly produces UO 2 . The incorporation of oxygen at the cooling stage of the combustion process is responsible for the formation of UO 2.12 . Spin coating of the solutions and brief annealing at 670 K allow the deposition of uniform films of UO 2.12 with thicknesses up to 300 nm on an aluminum substrate. Irradiation of films with Ar 2+ ions (1.7 MeV energy, a fluence of up to 1 × 10 17 ions/cm 2 ) shows unusual defect-simulated grain growth and enhanced chemical mixing of UO 2.12 with the substrate due to the high uranium ion diffusion in films. As a result, the method described in this work allows the preparation of actinide oxide targets for fundamental nuclear science research and studies associated with stockpile stewardship.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗