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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 415 records · Page 23
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

The transformation of U(VI) and V(V) in carnotite group minerals during dissimilatory respiration by a metal reducing bacterium

Recent results from laboratory and field studies support that dissimilatory metal reducing (DMR) bacteria influence the fate and transport of uranium in anaerobic subsurface environments. To date, most research efforts have focused on the reduction of soluble U(VI) by DMR bacteria to form insoluble uraninite (UO 2 ). Subsurface environments harbor, however, large reservoirs of U(VI) in solid or mineral form. Uranium that is structure-bound in minerals is expected to be more refractory to microbial reduction than soluble U, based on analogy with Fe respiration. The reducibility of U(VI) could impact the fate of U(IV) by controlling mineral precipitation reactions, which has implications for the long-term immobilization of U in subsurface environments. Here, we studied anoxic cultures of Shewanella putrefaciens CN32 incubated with natural carnotite-group minerals by X-ray diffraction, electron microscopy, scanning transmission X-ray microscopy (STXM). Near-edge X-ray absorption fine structure (NEXAFS) spectroscopy measurements at U-N 4,5 , V-L 2,3 , and O-K edges on cultures incubated up to 10 months show that V(V) was reduced to V(IV), whereas U was not reduced. In contrast, V(V) and U(VI) in solution were both completely reduced to lower oxidation states by CN32, as precipitates within the exopolymer surrounding the bacteria. Assays for the toxicity of U and V to CN32 showed that biofilm formation was stimulated at 0.001 M U(VI), and growth was inhibited at concentrations of U(VI) greater than 0.001 M. Vanadium did not inhibit growth or stimulate biofilm formation at any concentration tested. Investigations of the bacteria-mineral and bacteria-metal interface at the nanometer and molecular scales provide new insights into the co-respiration of V and U that help explain their biogeochemical cycling and have implications for subsurface bioremediation of these elements.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

The effect of stress on the migration of He gas bubbles under a thermal gradient in Fe by phase-field modeling

Here a phase-field model is parameterized to study the effect of elastic stresses on the migration of He gas bubbles in Fe under a temperature gradient. Stresses caused by the gas bubble pressure and residual stress in the Fe matrix are considered. The dependence of He bubble migration velocity on the magnitude of the residual stress, average temperature, temperature gradient, and bubble size is measured. In agreement with a theoretical model based on surface diffusion, simulation results demonstrate that He bubbles move towards the high temperature region with velocities in Fe that are orders of magnitude faster than previously reported in UO 2 . It is found that local stresses in the matrix caused by the He bubble have negligible effect on the bubble migration process; however, residual stresses in the Fe matrix, potentially caused by processing or irradiation, can modestly modify bubble kinetics through pressure dependence of the He diffusion coefficients. Compressive residual stress decreases diffusion coefficients for bulk and surface diffusion mechanisms, thus reducing the migration velocity of the gas bubble. In contrast, tensile residual stress increases the diffusion coefficients, resulting in an increase in the gas bubble migration velocity. This pressure dependence is also consistent with a theoretical model. This phase field model lays the foundation for analysis of bubble coalescence-induced fracture in He bubble-containing steels.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Binding of uranyl cations to a Zr-based metal-organic framework by density functional theory

Zr-based metal–organic frameworks (Zr-MOFs) have been widely used as ion adsorbents for the removal or extraction of toxic and/or radionuclide species from aqueous solutions. However, the mechanisms by which uranyl cations (UO 2 2+ ) interact with Zr-MOFs have not been established. In this work, the nature of the bonding of uranyl cations with a Zr-MOF was determined using density functional theory for nineteen structurally distinct candidate complexes. Further, the results showed that in all cases the binding energy was of the order of 1.5 eV, but depended on the specific bonding site. The most stable structure involved coordination of the uranyl cation and two structurally distinct oxygens in the Zr-MOF metal node. It was also found that higher degree of deprotonation in Zr-MOF correlated with higher binding energy between the Zr-MOF and uranyl cations. These insights can aid in the design of Zr-MOFs with optimized features for efficient capture of uranyl cations.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Accelerated and heterogeneous corrosion of Cr-doped uranium nitride fuel pellets

The steam oxidation of Cr-doped UN fuel pellets is analyzed during sequential isothermal holds up to 720 °C. In situ neutron diffraction results show how Cr is accommodated in a secondary U 2 CrN 3 phase, leading to the formation of a duplex UN/U 2 CrN 3 microstructure. Under corrosion, the oxidation of the two phases begins at 400 °C for UN and 430 °C for U 2 CrN 3 , respectively. Because the UN phase is preferentially oxidized in the presence of U 2 CrN 3 , addition of Cr in UN based nuclear fuel is found to accelerate the corrosion rate. At 430 °C the oxidation of UN in the UN/U 2 CrN 3 microstructure is ~ 5 times faster than pure UN, increasing to ~19 times faster at 460 °C. The oxidation of U 2 CrN 3 produces UO 2 via the formation of two transient intermediate phases. In situ neutron diffraction enables oxidation processes of UN and U 2 CrN 3 components to be followed separately within the two-phase system.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Thermodynamic properties of metaschoepite predicted from density functional perturbation theory

Density functional perturbation theory (DFPT) calculations of the thermodynamic properties of metaschoepite, (UO 2 ) 8 0 2 (OH) 12 •10H 2 0, are reported. Using a recently revised crystal structure of metaschoepite, the predicted molar entropy and isobaric heat capacity are overall significantly smaller than previous calculations using an earlier orthorhombic crystal structure model. The present DFPT calculations also show large differences between the thermo-dynamic functions of metaschoepite and schoepite, which might reflect the change in phonon properties upon removal of two H 2 0 molecules per formula unit and alteration of the H-bonded interlayer water network from schoepite to metaschoepite.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Spectroscopic investigation of historical uranium glasses

Here, we present spectroscopic investigations of uranium in two historical glasses. A 1920's-1930's Art Deco (Bagley Carnival) green soda-lime-silica uranium glass bowl and a 1930's-1940's (Thomas Webb) yellow full lead crystal uranium glass vase were studied using optical absorption spectroscopy, XRF and SEM-EDX, X-band EPR and High- Energy Resolution Fluorescence-Detected (HERFD)-XANES. Uranium speciation may be different in the two glasses. Uranium occurs as uranyl groups (UO 2 2+ ) with minority reduced uranium species, mostly as U(V), more important in the Webb glass than in the Bagley glass. The differences between the two glasses arise from the fining procedure and result from the presence of other multivalent elements and differences in base glass composition. Electron Paramagnetic Resonance (EPR) spectra show the presence of an incipient signal close to g~2, partially annealed by heating to 550 °C, which we assign to a small concentration of radiation-induced defects caused by uranium decay. The presence of multivalent glass components in addition to U (As + Cu + Fe in the Bagley glass and Sb + Fe in the Webb glass) can trap the electron-hole pairs generated by the presence of uranium. This may explain the weakness of this signal in these glasses that anyway received a limited radiation dose since their fabrication. The low activity of these glasses, close to the background radiation, confirms that there is no danger to exhibiting them in museums.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Preparation of a uranium monocarbide anode and electrochemical characterization in molten LiCl-KCl-UCl 3

Porous uranium carbide (UC) pellets possessing moderate electrical conductivity were synthesized by reaction of UO 2 with graphite at temperatures up to 1550°C under rough vacuum. Conversions as high as 98% were achieved at soak times of 2-4 hours. The electrochemistry of the UC pellets in molten LiCl-KCl-6.5 wt% UCl 3 was explored using a variety of techniques including DC polarization methods, cyclic voltammetry, chronopotentiometry and bulk electrolysis. Here, the electrode reaction for anodic dissolution was found to be kinetically controlled by dissociation of UC to a transition state complex that was hypothesized to consist of a uranium atom partially complexed by chloride ions. Precise measurements of current efficiencies using chronopotentiometry indicated upper limits of 90.9 ± 3.4% and 98.3 +1.7/-3.7% for anode and cathode, respectively, when operating at anodic overpotentials near +300 mV. Bulk electrolysis of a UC pellet performed by passing 98% of the theoretical charge resulted in nearly complete recovery of its uranium content as highly pure metal at the cathode.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Mechanisms and kinetics of citrate-promoted dissolution of a uranyl phosphate mineral

The formation of uranyl phosphate precipitates is a remediation strategy to limit the mobility of uranium in contaminated soils. However, exposure to organic ligands, like the plant exudate citrate, can remobilize the uranium. The purpose of this study is to provide a more thorough comprehension of citrate-promoted dissolution of a uranyl phosphate mineral, chernikovite [(H 3 O)(UO 2 )(PO 4 )·3H 2 O], by determining the extent of uranium release from chernikovite at a wide range of citrate concentrations. Here we have quantified the kinetics of dissolution and proposed potential mechanisms of chernikovite dissolution to gain a better understanding of the fate of uranyl phosphate precipitates in the environment. Batch dissolution and continuously stirred tank reactor (CSTR) experiments indicate that increasing citrate concentrations from 0.1 mM to 50 mM increases the concentration of dissolved uranium. However, at citrate concentrations of 10 mM and greater, the effectiveness of the ligand to enhance the dissolution decreases. Density functional theory modeling and Raman spectroscopy indicate a strong interaction between citrate and the uranyl ions at the surface of the uranyl phosphate. The interaction between citrate and uranyl suggests the formation of an alteration layer of uranyl-citrate surface complexes, which could impede uranium release at citrate concentrations of 10 mM or higher. Together, these results indicate that the citrate-promoted dissolution of chernikovite is a fast reaction that is hindered by a combination of surface saturation and a secondary-phase precipitation reaction at higher concentrations of citrate and reaction times greater than or equal to 72 min based on the hydraulic residence time of CSTR experiments. These findings improve our understanding of the stability of uranyl phosphate for use in sequestration of uranium groundwater plumes at contaminated sites.

58 GEOSCIENCES↗
DOE OSTI2022

U(VI) binding onto electrospun polymers functionalized with phosphonate surfactants

We previously observed that phosphonate functionalized electrospun nanofibers can uptake U(VI), making them promising materials for sensing and water treatment applications. Here, we investigate the optimal fabrication of these materials and their mechanism of U(VI) binding under the influence of environmentally relevant ions (e.g., Ca 2+ and CO 3 2- ). We found that U(VI) uptake was greatest on polyacrylonitrile (PAN) functionalized with longer-chain phosphonate surfactants (e.g., hexa- and octadecyl phosphonate; HDPA and ODPA, respectively), which were better retained in the nanofiber after surface segregation. Subsequent uptake experiments to better understand specific solid-liquid interfacial interactions were carried out using 5 mg of HDPA-functionalized PAN mats with 10 μM U at pH 6.8 in four systems with different combinations of solutions containing 5 mM calcium (Ca 2+ ) and 5 mM bicarbonate (HCO 3 -). U uptake was similar in control solutions containing no Ca 2+ and HCO 3 - (resulting in 19 ± 3% U uptake), and in those containing only 5 mM Ca2+ (resulting in 20 ± 3% U uptake). A decrease in U uptake (10 ± 4% U uptake) was observed in experiments with HCO 3 -, indicating that UO 2 -CO 3 complexes may increase uranium solubility. Results from shell-by-shell EXAFS fitting, aqueous extractions, and surface-enhanced Raman scattering (SERS) indicate that U is bound to phosphonate as a monodentate inner sphere surface complex to one of the hydroxyls in the phosphonate functional groups. New knowledge derived from this study on material fabrication and solid-liquid interfacial interactions will help to advance technologies for use in the in-situ detection and treatment of U in water.

42 ENGINEERING↗
DOE OSTI2021

Multicomponent pyrochlore solid solutions with uranium incorporation – A new perspective of materials design for nuclear applications

Multicomponent pyrochlore solid solutions with and without uranium incorporation were fabricated and their thermal-mechanical properties were characterized. Multicomponent pyrochlore solid solutions without uranium exhibit comparable thermal conductivity and higher mechanical strength compared to baseline single component rare-earth titanate pyrochlore (A 2 Ti 2 O 7 ). Uranium incorporation reduces hardness as compared with single component compositions. High entropy pyrochlore with uranium displays the highest thermal conductivity within multicomponent pyrochlore solid solutions with significantly better mechanical properties than UO 2 . The measured thermal conductivity correlates well with A-site cation mixing entropy and a modified size disorder parameter, and thus the size disorder and mixing entropy could be good indicators for predicting thermal conductivity of multicomponent pyrochlore solid solutions. This work opens up the possibility of designing multicomponent oxide solid solutions by controlling their chemical disorder/mixing entropy to achieve acceptable thermal-mechanical properties, desired radiation and corrosion performance for potential nuclear waste form and inert matrix fuel applications.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Accumulation mechanisms for contaminants on weak-base hybrid ion exchange resins

Mechanism of hexavalent chromium removal (Cr(VI) as CrO 4 2- ) by the weak-base ion exchange (IX) resin ResinTech® SIR-700-HP (SIR-700) from simulated groundwater is assessed in the presence of radioactive contaminants iodine-129 (as IO 3 - ), uranium (U as uranyl UO 2 2+ ), and technetium-99 (as TcO 4 - ), and common environmental anions sulfate (SO 4 2- ) and chloride (Cl-). Batch tests using the acid sulfate form of SIR-700 demonstrated Cr(VI) and U(VI) removal exceeded 97%, except in the presence of high SO 4 2- concentrations (536 mg/L) where Cr(VI) and U(VI) removal decreased to ≥ 80%. However, Cr(VI) removal notably improved with co-mingled U(VI) that complexes with SO 4 2- at the protonated amine sites. These U–SO 4 2- complexes are integral to U(VI) removal, as confirmed by the decrease in U(VI) removal (<40%) when the acid chloride form of SIR-700 was used instead. Solid phase characterization revealed that CrO 4 2- is removed by IX with SO 4 2- complexes and/or reduced to amorphous Cr(III)(OH) 3 at secondary alcohol sites. Tc(VII)O 4 - and I(V)O 3 - also undergo chemical reduction, following a similar removal mechanism. Oxyanion removal preference is determined by the anion reduction potential (CrO 4 2 - >TcO 4 - >IO 3 - ), geometry, and charge density. For these reasons, 39% and 69% of TcO 4 - and 17% and 39% of IO 3 - are removed in the presence and absence of Cr(VI), respectively.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2026

Coupling geophysical, geological, geochemical and mineralogical assessments to examine preferential contaminant transport pathways in interbedded fractured bedrock

This study shows that a multi-faceted approach, combining borehole geophysical logging and surface seismic P-wave first-arrival tomography with confirmatory coring, well installation, and chemical and mineralogical analysis, is effective for identifying difficult-to-locate preferential contaminant transport pathways in deeper fractured bedrock. Seismic tomography detected porous 10–20 m wide elongated fractured conduits that allow acidic groundwater contaminated with uranium (U) and nitrate (NO 3 - ) to migrate within interbedded shale-limestone bedrock over 1000 m from a former disposal facility (S-3 Ponds site) located at the DOE Y-12 National Security Complex in Tennessee (USA). Conventional drilling techniques would easily miss these conduits because they are oriented parallel with fractured bedding planes. Synchrotron analysis of aquifer solids revealed that > 95 % of the U is hexavalent (U VI ). This uranyl (UO 2 2+ ) species is coordinated with carbonate, iron oxide, silicate and phosphate minerals within cemented white to yellow precipitates, which contain U concentrations as high as ∼21.6 % by weight fraction. Identifying the presence of these mineral phases, enables a further understanding of the potential effectiveness of remediation actions. The combination of methodologies presented here can also be applied to other explorations, such as the detection of water supply.

Fate and transport↗
DOE OSTI2020

High temperature steam oxidation dynamics of U 3 Si 2 with alloying additions: Al, Cr, and Y

Uranium silicides are considered for advanced technology reactor fuels due to their enhanced thermal conductivity and high uranium density (U 3 Si and U 3 Si 2 ) compared to traditional UO 2 . Susceptibility to oxidation and wash out, in the event of a cladding breech, could limit the potential for deployment of silicides as accident tolerant fuels. Mitigating the water reaction for U 3 Si 2 could enable its use as an accident tolerant, high uranium density fuel or as a composite fuel constituent. Reported here is the impact of alloying additions of Al, Cr and Y on the high temperature, steam oxidation response of U 3 Si 2 . In addition to the thermogravimetric response, as melted microstructures, phase compositions and post oxidation analysis are also presented. Here, the investigation shows steam oxidation dynamics are altered, from non-alloyed U 3 Si 2 , under thermally ramped conditions. However, additional alloy development of these fuel forms is necessary for further consideration as candidate accident tolerant fuels in water-cooled reactor designs.

36 MATERIALS SCIENCE↗
DOE OSTI2020

In-pile thermal conductivity of uranium dioxide at low burnup

In-pile thermal conductivity of uranium dioxide (UO 2 ) was investigated at low burnup levels (<0.2 kWd/kg-UO 2 ) to elucidate the concurrent effects of the fission damage and the thermal recovery. In-pile experiments were performed in the Halden reactor. The fuel centerline and the cladding temperatures were measured during the experiments, and the power level of specially designed test rodlets were also monitored. The uncertainty of the pellet-cladding gap's thermal resistivity was minimized with a liquid-metal bond, while comparable measurements were made in reference test rodlets without liquid metal bonding. The experimental data were analyzed using the inverse heat transfer approach. The heat conduction equation was solved by applying the measured temperature of the cladding as a boundary condition, and the calculated fuel centerline temperature was compared with the measured fuel centerline temperature to determine the relative matrix resistivity. Results showed that the in-pile matrix resistivity, the phonon-lattice interaction terms of the thermal conductivity, was ~1.5–2.5 times higher than its unirradiated value due to fission-induced damage. Finally, the in-pile annealing caused a significant recovery of the matrix resistivity, and the amount of recovery increased from ~20 to ~33%, while the annealing temperature increased from 700 to 1,000 °C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Phase-field simulations of intergranular fission gas bubble behavior in U 3 Si 2 nuclear fuel

U 3 Si 2 is a potential accident-tolerant fuel that shows promise due to its high thermal conductivity and higher uranium density relative to UO 2 . However, its swelling and fission gas release behavior in light water reactor (LWR) conditions is relatively unknown. To provide mechanistic insight and determine parameters for engineering-scale fuel performance modeling of pellet-form U 3 Si 2 , phase-field simulations of the growth, interconnection, and venting of intergranular fission gas bubbles were performed. The fractional coverage of the grain boundary and the fraction of bubble area that is vented were calculated as a function of time. From the simulation data, the fractional grain boundary coverage at saturation, an important parameter needed in engineering-scale modeling of swelling and fission gas release, was determined. Multiple simulations were run to determine the uncertainty in the calculated value. Finally, the effect of model assumptions and input parameters that are not well known was evaluated. Simulation results are compared to related theoretical and computational work. Based on the simulation results, a value of 0.60 for the fractional grain boundary coverage at saturation is recommended for U 3 Si 2 fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Controlled current-rate AC flash sintering of uranium dioxide

Uranium dioxide (UO2) pellets with controlled microstructures were densified up to 93.4% of their theoretical density in less than 25 minutes at a furnace temperature of 873 K, utilizing controlled current-rate alternating current (AC) flash sintering (FS). Using this AC-FS method it was possible to control the sintering rate and thermal gradients, resulting in dense pellets with no appreciable hourglassing and good mechanical integrity. Moreover, the apparent sintering activation energy for FS and for conventionally sintered samples was estimated to be 108 kJ mol -1 and 380 kJ mol -1 , respectively using the master sintering curve method. The apparent activation energy for FS was remarkably close to those reported in the literature for spark plasma sintering of UO 2 . Furthermore, both these field assisted sintering methods utilize fast heating rates and electrical effects that are likely enhancing the grain boundary diffusion mechanism. Finally controlled current-rate AC-FS has been demonstrated as a technological advancement, capable of producing ceramic nuclear fuels in a fraction of the conventional processing time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Development of a U-19Pu-10Zr fuel performance benchmark case based on the IFR-1 experiment

Metallic nuclear fuels are subject to research and development for use in advanced reactors. Robust, accurate metallic fuel performance models are important for the design, safety analysis, and licensing of these reactors. However, metallic fuel performance models require additional development; they are not as mature as UO 2 fuel performance models. A benchmark case based on the IFR-1 experiment was developed to better gauge the accuracy of existing models, identify models for high-priority development, and potentially quantify any future improvements made by further model development.This work collected publicly available information on the IFR-1 experiment and used it to develop the benchmark case. Fuel behavior during the IFR-1 irradiation was simulated by using the fuel performance code BISON, and the predicted results were compared with postirradiation examination data from the IFR-1 experiment. Furthermore, a sensitivity study and tuning studies were performed as a preliminary investigation into the causes of inaccurate temperature and dimensional change predictions.The benchmark predicted reasonably accurate values for the burnup and fission gas release. There was error in the predicted temperatures, which could be explained by uncertainty in the input parameters and legacy temperatures. BISON over-predicted dimensional changes in the fuel and cladding. The sensitivity study showed that the dimensional changes were most sensitive to the fuel swelling anisotropy and the cladding void swelling model. Future benchmark and model development should focus on cladding swelling behaviors to improve dimensional change predictions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Coupled physics simulation of fracture in nuclear fuel pellets induced by resistive heating

While it is well known that the UO 2 fuel used in light-water reactors cracks early in its life in the reactor, it is very difficult to collect in-reactor data regarding fuel cracking initiation and propagation to validate fracture models in fuel performance codes. Most fracture data comes from the post-irradiation examination of fuel that has been subjected to long and complex power histories. Replicating the thermal gradients experienced in this fuel to study crack initiation and propagation in a laboratory environment is challenging. This is primarily because the fuel is heated volumetrically by fission, which is difficult to replicate. One way to approximate this volumetric heating out of the reactor is to use electrical resistance (or Joule heating). An electrical conductivity model has been developed and coupled with the existing thermomechanical modeling capability in the BISON nuclear fuel performance code. This will permit simulation of experiments that use resistive heating for the purpose of validating fracture models in BISON. The modeling approach developed here is applied to simulate both historical and current resistive heating experiments on ceramic nuclear fuel.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗