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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↗
NASA NTRS2015

Methodology for Producing a Uniform Distribution of UO2 in a Tungsten Matrix

Current work at NASA's Marshall Space Flight Center (MSFC) is focused on the development CERMET fuel materials for Nuclear Thermal Propulsion (NTP). The CERMETs consist of uranium dioxide (UO2) fuel particles embedded in a tungsten (W) metal matrix. Initial testing of W-UO2 samples fabricated from fine angular powders performed reasonably well, but suffered from significant fuel loss during repeated thermal cycling due to agglomeration of the UO2 (1). The blended powder mixtures resulted in a non-uniform dispersion of the UO2 particles in the tungsten matrix, which allows rapid vaporization of the interconnected UO2 from the sample edges into the bulk material. Also, the angular powders create areas of stress concentrations due to thermal expansion mismatch, which eventually cracks the tungsten matrix. Evenly coating spherical UO2 particles with chemical vapor deposited (CVD) tungsten prior to consolidation was previously demonstrated to provide improved performance. However, the CVD processing technology is expensive and not currently available. In order to reduce cost and enhance performance, a powder coating process has been developed at MSFC to produce a uniform distribution of the spherical UO2 particles in a tungsten matrix. The method involves utilization of a polyethylene binder during mixing which leads to fine tungsten powders clinging to the larger UO2 spherical particles. This process was developed using HfO2 as a surrogate for UO2. Enough powder was mixed to make 8 discs (2cm diameter x 8mm thickness) using spark plasma sintering. A uniaxial pressure of 50 MPa was used at four different temperatures (2 samples at each temperature). The first two samples were heated to 1400C and 1500C respectively for 5 minutes. Densities for these samples were less than 85% of theoretical, so the time at temperature was increased to 20 minutes for the remaining samples. The highest densities were achieved for the two samples sintered at 1700C (approx. 92% of theoretical). Scanning electron microscopy (SEM) of the mixed powders and the sintered samples along with energy dispersive x-ray analysis was obtained. The SEM of the powders clearly show the fine W powder adhered to the larger HfO2 particles and a uniform distribution of HfO2 particles in a tungsten matrix upon densification. Vicker's Microhardness testing was also performed on all samples using 0.5, 1.0 and 2.0 kg loads. Five indents were made at each load level. All indents were placed in the tungsten matrix to assist as a proxy in measuring densification. The highest hardness value was obtained for the 1700C specimens. The hardness average for these samples was 312.14 MPa. This powder processing method has been applied to W/UO2 powders with the SEM of the powders appearing similar to the W/HfO2 powder images.

Tucker, Dennis S.↗
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↗
NASA NTRS1971

Out-of-core evaluations of a nonfueled and a UO2-fueled cylindrical thermionic converter.

Two cylindrical thermionic energy converters similar to those in a flashlight thermionic fuel element were tested electrically at the Jet Propulsion Laboratory. One converter was not fueled, but the other was fueled with UO2 imbedded in six pencil-lead-size holes in the emitter, which was made of rhenium. The nonfueled converters showed no change in its performance during 4000 hr of testing, whereas the output current of the UO2-fueled converter degraded 15% during 2400 hr of testing at 2000 K. Measurements on the UO2-fueled converter showed an increase in the collector work function and an increase in the bare emitter work function. The degradation appears to be caused primarily by foreign deposits on the collector, probably uranium that diffused from the UO2 fuel through the emitter.

Shimada, K.↗
NASA NTRS2013

Advances in the Development of a WCl6 CVD System for Coating UO2 Powders with Tungsten

W-UO2 CERMET fuels are under development to enable Nuclear Thermal Propulsion (NTP) for deep space exploration. Research efforts with an emphasis on fuel fabrication, testing, and identification of potential risks is underway. One primary risk is fuel loss due to CTE mismatch between W and UO2 and the grain boundary structure of W particles resulting in higher thermal stresses. Mechanical failure can result in significant reduction of the UO2 by hot hydrogen. Fuel loss can be mitigated if the UO2 particles are coated with a layer of high density tungsten before the consolidation process. This paper discusses the work to date, results, and advances of a fluidized bed chemical vapor deposition (CVD) system that utilizes the H2-WCl6 reduction process. Keywords: Space, Nuclear, Thermal, Propulsion, Fuel, CERMET, CVD, Tungsten, Uranium

Mireles, Omar R.↗
NASA NTRS2014

Development Status of a CVD System to Deposit Tungsten onto UO2 Powder via the WCI6 Process

Nuclear Thermal Propulsion (NTP) is under development for deep space exploration. NTP's high specific impulse (> 850 second) enables a large range of destinations, shorter trip durations, and improved reliability. W-60vol%UO2 CERMET fuel development efforts emphasize fabrication, performance testing and process optimization to meet service life requirements. Fuel elements must be able to survive operation in excess of 2850 K, exposure to flowing hydrogen (H2), vibration, acoustic, and radiation conditions. CTE mismatch between W and UO2 result in high thermal stresses and lead to mechanical failure as a result UO2 reduction by hot hydrogen (H2) [1]. Improved powder metallurgy fabrication process control and mitigated fuel loss can be attained by coating UO2 starting powders within a layer of high density tungsten [2]. This paper discusses the advances of a fluidized bed chemical vapor deposition (CVD) system that utilizes the H2-WCl6 reduction process.

Mireles, O. R.↗
NASA NTRS2015

Optimization of a Wcl6 CVD System to Coat UO2 Powder with Tungsten

In order to achieve deep space exploration via Nuclear Thermal Propulsion (NTP), Marshall Space Flight Center (MSFC) is developing W-UO2 CERMET fuel elements, with focus on fabrication, testing, and process optimization. A risk of fuel loss is present due to the CTE mismatch between tungsten and UO2 in the W-60vol%UO2 fuel element, leading to high thermal stresses. This fuel loss can be reduced by coating the spherical UO2 particles with tungsten via H2/WCl6 reduction in a fluidized bed CVD system. Since the latest incarnation of the inverted reactor was completed, various minor modifications to the system design were completed, including an inverted frit sublimer. In order to optimize the parameters to achieve the desired tungsten coating thickness, a number of trials using surrogate HfO2 powder were performed. The furnace temperature was varied between 930 C and 1000degC, and the sublimer temperature was varied between 140 C and 200 C. Each trial lasted 73-82 minutes, with one lasting 205 minutes. A total of 13 trials were performed over the course of three months, two of which were re-coatings of previous trials. The powder samples were weighed before and after coating to roughly determine mass gain, and Scanning Electron Microscope (SEM) data was also obtained. Initial mass results indicated that the rate of layer deposition was lower than desired in all of the trials. SEM confirmed that while a uniform coating was obtained, the average coating thickness was 9.1% of the goal. The two re-coating trials did increase the thickness of the tungsten layer, but only to an average 14.3% of the goal. Therefore, the number of CVD runs required to fully coat one batch of material with the current configuration is not feasible for high production rates. Therefore, the system will be modified to operate with a negative pressure environment. This will allow for better gas mixing and more efficient heating of the substrate material, yielding greater tungsten coating per trial.

Belancik, Grace A.↗
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↗
NASA NTRS1971

Molybdenum-UO2 cerment irradiation at 1145 K

Two molybdenum-UO2 cermet fuel pins were fission heated in a helium-cooled loop at a temperature of 1145 K and to a total burnup of 5.3 % of the U-235. After irradiation the fuel pins were measured to check dimensional stability, punctured at the plenums to determine fission gas release, and examined metallographically to determine the effect of irradiation. Burnup was determined in several sections of the fuel pin. The results of the postirradiation examination indicated: (1) There was no visible change in the fuel pins on irradiation under the above conditions. (2) The maximum swelling of the fuel pins was less than 1%. (3) There was no migration of UO2 and no visible interaction between the molybdenum and the UO2. (4) Approximately 12% of the fission gas formed was released from the cermet cone into the gas plenum.

Mcdonald, G.↗
NASA NTRS1966

Basic Behavior and Properties of W-UO2 Cermets

Thermal cycling studies on cermets prepared from tungsten and UO2 powder mixtures and cermets prepared using tungsten coated UO2 particles are discussed. The cermets were tested in hydrogen at high temperatures to determine the effects of temperature, thermal cycling, and hydrogen pressure. Basic mechanism studies of tungsten­-UO2 interactions in thermally cycled cermets are also described.

Baker, R. J.↗
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↗