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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

High density UO2 and high thermal conductivity UO2 composites by spark plasma sintering (SPS)

Embodiments of the invention are directed to a method for production of a nuclear fuel pellet by spark plasma sintering (SPS), wherein a fuel pellet with more than 80% TD or more than 90% TD is formed. The SPS can be performed with the imposition of a controlled uniaxial pressure applied at the maximum temperature of the processing to achieve a very high density, in excess of 95% TD, at temperatures of 850 to 1600° C. The formation of a fuel pellet can be carried out in one hour or less. In an embodiment of the invention, a nuclear fuel pellet comprises UO 2 and a highly thermally conductive material, such as SiC or diamond.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

UO2 microstructural evolutions induced by Ni, Mo, and W dopants for intentional forensics

The concept of tagging nuclear fuel with a chemical barcode to enable forensics analysis across the nuclear fuel cycle is an area of active investigation, particularly to ensure fabrication viability without disrupting current fuel performance. This study explored the feasibility of using Ni, Mo, and W isotopic double-spikes as dopants in UO2 fuel from the perspective of fuel fabrication. Doped UO2 pellets were produced using conventional fuel fabrication processes, including powder mixing, sieving, pressing, and sintering in a reductive atmosphere. Two composition levels, 100 and 1000 ppm, were evaluated for each dopant element with isotopic double-spike configurations. For the Ni system, additional dopant concentrations of 250 and 500 ppm were produced with nonperturbed isotopic ratios. The results demonstrated that successful incorporation of Ni, Mo, and W double-spikes into UO2 pellets occurred with minimal shift in final density or dopant loss during pellet fabrication. Isotopic analysis confirmed the presence of the double-spike signature even when diluted with natural isotopic material in ratio of 1:5 in the fabrication process. Microstructural examinations revealed different impacts on grain size compared with undoped UO2. This study showed that Ni incorporation up to ∼500 ppm promoted moderate grain growth, whereas the Mo and W systems caused grain size reduction at all concentrations. Changes in the UO2 lattice parameter as a function of composition were detected exclusively for Ni up to 500 ppm, indicating that the Ni solid solution was the main factor for the observed grain growth. Insoluble (Mo and W) or supersaturated (Ni > 500 ppm) conditions produced grain size reduction. The Ni-doped pellets in the solution range resulted in a final microstructure within fuel specifications, demonstrating its potential benefits of employing complex dopant systems for potential nuclear forensic applications.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Modeling Non-UO2 Fuel With UNF-ST&DARDS

The U.S. Department of Energy’s Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) provides an easy-to-use interface to analyze irradiated UO2 fuel by allowing all analysis to be performed within the software and being able to store and use dozens of fuel assembly, canister, and cask designs [1]. However, performing these same analyses with non-UO2 fuel, such as UN or U3Si2, requires more user intervention in the process. This work uses UN, UN-ZrO2, and U3Si2 fuel to demonstrate how to perform criticality analyses in the current versions of UNF-ST&DARDS and how a non-UO2 fuel will compare to UO2. This work is part of a larger effort that also includes shielding and thermal analyses, but they will not be discussed.

Ivanusa, Pavlo↗
DOE OSTI2021

Seaborgite, LiNa6K2(UO2)(SO4)5(SO3OH)(H2O), the First Uranyl Mineral Containing Lithium

Abstract Seaborgite (IMA2019-087), LiNa6K2(UO2)(SO4)5(SO3OH)(H2O), is a new mineral species from the Blue Lizard mine, Red Canyon, San Juan County, Utah, U.S.A. It is a secondary phase found on gypsum in association with copiapite, ferrinatrite, ivsite, metavoltine, and römerite. Seaborgite occurs in sprays of light-yellow, long flattened prisms or blades, up to about 0.2 mm in length. Crystals are elongated on [100], flattened on {010}, and exhibit the forms {100}, {010}, {001}, and {101}. The mineral is transparent with vitreous luster and very pale-yellow streak. It exhibits bright lime-green fluorescence under a 405 nm laser. The Mohs hardness is ~2½. The mineral has brittle tenacity, curved or conchoidal fracture, and one good cleavage on {100}. The measured density is 2.97(2) g/cm3. The mineral is immediately soluble in H2O at room temperature. The mineral is optically biaxial (–), α = 1.505(2), β = 1.522(2), γ = 1.536(2) (white light); 2Vmeas = 85(1)°; moderate r < ν dispersion; orientation X ^ a ≈ 10°; pleochroic X colorless, Y and Z light green-yellow; X < Y ≈ Z. EPMA and LA-ICP-MS analyses of seaborgite undermeasured its Li, K, and Na. The empirical formula using Li, Na, and K based on the structure refinement is Li1.00Na5.81K2.19(UO2)(SO4)5(SO3OH)(H2O). Seaborgite is triclinic, P1, a = 5.4511(4), b = 14.4870(12), c = 15.8735(15) Å, α = 76.295(5), β = 81.439(6), γ = 85.511(6)°, V = 1203.07(18) Å3, and Z = 2. The structure (R1 = 0.0377 for 1935 I = 2σI) contains [(UO2)2(SO4)8]4– uranyl-sulfate clusters that are linked into a band by bridging LiO4 tetrahedra. The bands are linked through peripheral SO4 tetrahedra forming a thick heteropolyhedral layer. Channels within the layers contain a K site, while an additional K site, six Na sites, and an SO3OH group occupy the space between the heteropolyhedral layers.

Geochemistry & Geophysics↗
DOE OSTI2020

BISON capability and validation for U3Si2, Cr2O3-doped UO2, FeCrAl, and Cr-coated Zircaloy ATF concepts

Development, implementation, and validation of material and behavior models for accident tolerant fuel (ATF) concepts in the BISON fuel performance code began in 2014 in response to the events that occurred at the Fukushima Daichii nuclear power plant in March 2011. Early on, the focus was on U3Si2 fuel and FeCrAl cladding as part of a high-impact problem through the Nuclear Energy and Advanced Model- ing Simulation program [1]. Then, developments for Cr2O3- doped UO2 fuel, as well as SiC-SiC and Cr-coated Zircaloy cladding began based upon industry interests. Near the end of fiscal year 2018 the Consortium for Advanced Simulation of Light Water Reactors took over further Accident Tolerant Fuel (ATF) work in BISON in support of Nuclear Regulatory Commission (NRC) engagement. Discussions with the NRC clarified their list of priority fuel and cladding concepts, which included Cr2O3-doped UO2 and U3Si2 fuels, as well as Cr-coated Zircaloy and FeCrAl claddings. Further, reports similar in form to NUREG/CR-7024 [2] that was developed for traditional light water reactor (LWR) materials, UO2 fuel with Zircaloy claddings were created for the priority fuel and cladding concepts. In this paper select details from the reports entitled "ATF material model development and validation for priority fuel concepts" [3] and "ATF material model development and validation for priority cladding concepts" [4] are presented. First, a brief overview of the available models and their associated uncertainty are provided for the priority fuel and cladding concepts. Then, for the concepts with experimental data, comparisons between simulation and experiment are shown taking into account forward uncertainty quantification (UQ) techniques. For the concepts without publicly available experimental data, parametric studies incorporating forward UQ are completed to verify claims made about the materials’ behavior.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Phase-field simulations of fission gas bubbles in high burnup UO2 during steady-state and LOCA transient conditions

To improve the economics of commercial nuclear energy generation, U.S. utilities are currently seeking licensing approval to operate UO2 fuel to higher burnups. One significant safety issue that must be addressed to obtain approval is the potential for fine fragmentation/pulverization of the fuel during a loss-of-coolant accident (LOCA). It has been hypothesized this is caused by the rapid increase of pressure in fission gas bubbles in the high burnup region of the fuel. To better understand this phenomenon, a novel phase-field model of the fission gas bubble microstructure in UO2 has been developed and implemented in Idaho National Laboratory's Marmot application for phase-field simulation of nuclear materials. Simulations of bubble growth in the high burnup region during steady-state conditions showed that initially overpressurized bubbles decreased in pressure during growth, but still remained above equilibrium pressure. During a LOCA transient, simulations of bubbles in the high burnup region showed that bubble size did not change significantly, and the pressure increase due to the transient was calculated and passed to a phase-field model of fracture.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Development of a Model for Irradiation-Induced Grain Growth in UO2 thin films

In this work, we develop a model of irradiation-induced grain growth in UO2 using the MARMOT mesoscale nuclear materials simulation tool. We couple the existing thermally activated grain growth model with a heat conduction model that includes random heat sources representing thermal spikes. We compare the results with the irradiation data on UO2 thin films.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Meso-scale modeling of UO2 nuclear fuel to high burnup

To improve the economics of light water reactors for commercial nuclear energy generation, utility operators are seeking to obtain regulatory approval to run UO2 fuel to higher levels of burnup. One potential impediment to obtaining this approval is the phenomenon of fuel fragmentation, relocation, and dispersal (FFRD). FFRD can result when fuel experiences a rapid temperature transient, such as that occurring during a Loss Of Coolant Accident (LOCA). FFRD has historically been most associated with the rim region in UO2 fuel pellets, where the phenomenon of fragmentation is also referred to as pulverization due to the small size of the fragments. More recent evidence suggests that the so-called “dark zone” (due to its appearance in micrographs) that can be observed in the mid-radial regions of high burnup fuel is also susceptible to FFRD. Although empirical fuel performance models have been developed that can adequately predict pulverization in the rim region under typical LWR conditions, a scientific understanding of what underlies fuel restructuring and subsequent FFRD is lacking even in the rim region, and no models are currently available for the behavior the dark zone. To address these challenges, the U.S. Department of Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has employed a multi-scale modeling approach to improve scientific understanding and develop new fuel performance models. In this talk, I will focus on meso-scale efforts, which form a crucial link between atomic-scale and engineering-scale models. Phase-field modeling combined with cluster dynamics is used to predict the restructuring process in the rim region. Phase-field fracture modeling, informed by atomistic simulations, is used to predict the onset of pulverization in the rim region. Combining these techniques together allows the extent of rim pulverization to be predicted. The formation and evolution of the dark zone has also been simulated with the phase-field method, using an improved approach to vacancy source term parameterization. The work shows the important impact of microstructure on fuel performance.

fracture↗
DOE OSTI2025

In-situ ion irradiation of a spent UO2 fuel: evolution of fission products and nanograins with radiation dose

This study investigates the behavior of fission gas bubbles, five metal precipitates (5MPs) (Mo, Ru, Rh, Tc, Pd) and nanograins in spent uranium dioxide (UO2) fuel under various ion irradiation doses, temperatures, and flux conditions. Utilizing in-situ ion irradiation and advanced transmission electron microscopy, we analyzed the evolution of fission gas bubbles, 5MPs and nanograins in UO2 samples from the Belgium Reactor 3 (BR-3). Our findings reveal significant shrinkage of fission gas bubbles and 5MPs with increasing irradiation dose, accompanied by a decrease in pair density. The study also reveals nanograin coarsening under irradiation. We demonstrate that ion irradiation induces a homogeneous re-solution process where individual atoms are ejected from bubbles and precipitates, leading to their dissolution and subsequent re-precipitation in the matrix. This study provides critical insights into the dynamic behavior of fission products and nanograins under irradiation, facilitating the development of predictive models and contributing to the optimization of nuclear fuel performance and safety.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Phase-field simulations of fission gas bubbles in high burnup UO2 to inform engineering-scale fuel performance modeling

To improve the economics of commercial nuclear energy generation, U.S. utilities are currently seeking licensing approval to operate UO2 fuel to higher burnups. One significant safety issue that must be addressed to obtain approval is the potential for fine fragmentation/pulverization of the fuel during a loss-of-coolant accident (LOCA). The cause of pulverization has been hypothesized to be the rapid increase of pressure in fission gas bubbles in the high burnup region of the fuel during a LOCA. To better understand this phenomenon, a novel phase-field model of the fission gas bubble microstructure in UO2 has been developed and implemented in Idaho National Laboratory (INL)'s Marmot application for phase-field simulation of nuclear materials. Simulations of the bubble response to steady-state and transient conditions were conducted. Simulation results were used to inform a mechanistic model of pulverization in BISON, INL’s fuel performance simulation code.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Mesoscale Modeling of the Effects of Accelerated Burnup on UO2 Microstructural Evolution

Accelerating the nuclear fuel qualification process will rely on some combination of advanced modeling and simulation techniques with accelerated irradiation testing and separate effects experiments to enable the development of new fuel concepts in a shorter time frame. One of the key challenges to successfully leveraging accelerated irradiation tests will be understanding the artifacts that may be introduced with accelerated accumulation of dose and/or burnup. This work presents phase field (MARMOT) simulations of the evolution of representative 2D UO2 microstructures up to 40 MWd/kgU. Simulations were performed under both commercial light water reactor fuel conditions as well as those that would be expected for highly accelerated (~10x) burnup conditions similar to those used in the MiniFuel irradiations in Oak Ridge National Laboratory’s High Flux Isotope Reactor. The phase field model was coupled with a discrete nucleation algorithm to model re- structuring at high burnup. The effect of the different fission rates in both microstructures was investigated at two temperatures: 650?C and 800?C. The lower temperature simulations both showed an onset of restructuring at nearly 60 MWd/kgU. More extensive restructuring was obtained in the MiniFuel microstructure compared with that of the PWR fuel. At 800?C, no restructuring was obtained as a result of the thermally activated diffusion of Xe atoms and U vacancies to fission gas bubbles, which reduces the nucleation driving force. These results highlight the importance of using modeling and simulation tools to inform the environmental conditions during targeted accelerated irradiation tests to extract the most useful fuel performance data.

accelerated fuel qualification, Phase Field, Restr↗
DOE OSTI2024

Bobfinchite, Na[(UO2)8O3(OH)11]·10H2O, a new Na-bearing member of the schoepite family

Abstract The new mineral bobfinchite (IMA2020-082), Na[(UO2)8O3(OH)11]·10H2O, was found in the Burro mine, Slick Rock district, San Miguel County, Colorado, U.S.A., where it occurs as an oxidation product of uraninite on asphaltite matrix in intimate association with gypsum, natrozippeite, metaschoepite, and uranopilite. Bobfinchite crystals are transparent to translucent, yellow, lozenge-shaped disks up to 0.3 mm wide. Crystals are flattened on [100] and exhibit the forms {100}, {011}, {021}, {021}, and {011}. Bobfinchite has a pale-yellow streak and emits very dim yellow fluorescence under 365 nm ultraviolet illumination. The crystals are brittle with very good {100} cleavage and irregular, stepped fracture. The Mohs hardness is ca. 2 based on scratch tests. The calculated density is 5.044 g/cm3 based on the empirical formula and 5.036 g/cm3 for the ideal formula. Bobfinchite is optically biaxial (–), with α = 1.690(5), β = 1.7205(5), and γ = 1.730(5) (white light). The measured 2V, estimated from the interference figure, is 55(5)° and the calculated value is 59.1°. Dispersion is moderate, r > v; orientation: X = a, Y = b, Z = c; pleochroism: X nearly colorless, Y yellow, Z yellow; X < Y ≈Z. Electron microprobe analysis provided the empirical formula (Na0.99Pb0.02)[(UO2)7.99O3(OH)11]·10H2O. The five strongest X-ray powder diffraction lines are [dobs in Å(I)(hkl)]: 7.34(100)(200), 3.59(50)(024), 3.23(60)(224), 3.18(36)(240), and 2.01(23)(624,551,208,640,346). Bobfinchite is orthorhombic, Pbcn, a = 14.6249(9), b = 14.0389(10), c = 16.6923(10) Å, V = 3427.2(4) Å3, and Z = 4. The structure of bobfinchite (R1 = 0.0330 for 3770 I > 4σI) is built from uranyl oxide-hydroxide sheets that adopt the fourmarierite topology, with interlayer Na+ and H2O groups. Both the sheet and interlayer topology mimic those observed in natural and synthetic Na-metaschoepites studied previously, and as seen in other uranyl oxide hydrate minerals, charge balance is achieved at specific sites in the sheet through the substitution O2– ↔ (OH)–.

Geochemistry & Geophysics↗
DOE OSTI2022

Multi-scale simulation of high burnup UO2 nuclear fuel during loss-of-coolant accident conditions

To improve the economics of commercial nuclear energy generation, U.S. utilities are currently seeking licensing approval to operate UO2 fuel to higher burnups. One significant safety issue that must be addressed to obtain approval is the potential for fine fragmentation/pulverization of the fuel during a loss-of-coolant accident (LOCA). It has been hypothesized that pulverization is caused by the rapid increase of pressure in fission gas bubbles in the high burnup region of the fuel during a LOCA. To better understand this phenomenon, a novel phase-field model of the fission gas bubble microstructure in UO2 has been developed and implemented in Idaho National Laboratory (INL)'s Marmot application for phase-field simulation of nuclear materials. Simulations of the bubble response to steady-state and transient conditions were conducted, and the results were used to inform a mechanistic model of pulverization in BISON, INL’s fuel performance simulation code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Materials Data on UO2 by Materials Project

UO2 is Fluorite structured and crystallizes in the monoclinic P2_1/c space group. The structure is three-dimensional. there are two inequivalent U4+ sites. In the first U4+ site, U4+ is bonded in a body-centered cubic geometry to eight O2- atoms. There are a spread of U–O bond distances ranging from 2.33–2.36 Å. In the second U4+ site, U4+ is bonded in a body-centered cubic geometry to eight O2- atoms. There are a spread of U–O bond distances ranging from 2.34–2.36 Å. There are four inequivalent O2- sites. In the first O2- site, O2- is bonded to four U4+ atoms to form a mixture of edge and corner-sharing OU4 tetrahedra. In the second O2- site, O2- is bonded to four equivalent U4+ atoms to form a mixture of edge and corner-sharing OU4 tetrahedra. In the third O2- site, O2- is bonded to four U4+ atoms to form a mixture of edge and corner-sharing OU4 tetrahedra. In the fourth O2- site, O2- is bonded to four equivalent U4+ atoms to form a mixture of edge and corner-sharing OU4 tetrahedra.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on UO2 by Materials Project

UO2 is Hydrophilite structured and crystallizes in the tetragonal P4_2/mnm space group. The structure is three-dimensional. U4+ is bonded to six equivalent O2- atoms to form a mixture of edge and corner-sharing UO6 octahedra. The corner-sharing octahedral tilt angles are 52°. There are two shorter (2.25 Å) and four longer (2.27 Å) U–O bond lengths. O2- is bonded in a trigonal planar geometry to three equivalent U4+ atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Cracking in first ramp to power of standard versus high thermal conductivity UO2 pellets via internal nuclear heating

Advanced fuel designs that incorporate thinner fuel UO2 pellets interspaced by high thermal conductivity inserts have been proposed, with the primary goals of reducing peak centerline temperatures and temperature gradients across fuel pellets and enhancing heat transfer from the fuel to the coolant. An initial series of experiments has been performed on this design, including laboratory experiments and a series of experiments using the Idaho National Laboratory (INL) Transient Reactor Test (TREAT) Facility, the latter of which compared thermal gradient driven fracture of standard pellet designs with that in the proposed advanced fuel design. Although reducing fracture is not the primary objective of the new fuel design, the lower thermal gradients are expected to reduce fracture, so it can serve as an indicator of the thermal behavior of this fuel in the reactor. The in-reactor tests were conducted at multiple linear heat generation rates and confirm the expected result that fracture in both the standard and advanced fuel pellets occurs during the first ramp to power in standard light-water reactor conditions. Post-irradiation examination of the experiment material was performed and included quantification of the extent of fracture in the fuel pellets. It was found that the advanced-design pellets reduce the extent of fracture in a statistically significant way. This confirms the expected behavior predicted by two-dimensional axisymmetric models of this experiment. This study is an important first experimental confirmation of the efficacy of the proposed inserts for achieving their desired effect on the thermal behavior of the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗