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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 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Microstructural evolution and intermetallic formation in Zn-3Mg (wt%) powder mixture processed by high-pressure torsion

Severe plastic deformation (SPD) techniques have been used extensively over the past 40 years for producing strong metals and alloys. High-pressure torsion (HPT) is one of the most promising SPD techniques for achieving high strength through nanoscale grain refinement and phase transformation. In this research, a mixture of pure zinc (Zn) and magnesium (Mg) powders, Zn-3Mg (wt%), was HPT-processed under a pressure of 6 GPa for 1, 5, 10, 20, and 30 turns at room temperature to achieve a high strength biodegradable material. In order to understand the effects of pre-consolidation on the resulting microstructure and hardness, HPT processing was performed on loose powders placed in the die and also on a pre-compacted powder mixture and the resulting HPT disks were characterized by X-ray diffraction, scanning electron microscopy, atom probe tomography, and Vickers microhardness. In both cases, the HPT disk microstructures contained nanoscale grains, and stable and metastable strain-induced intermetallics, but an unusual softening appeared at large shear strains. Grain size, grain morphology, and the formation of different intermetallics were analyzed to explain the unusual hardness distribution, and it was found that an inverse Hall-Petch relationship between hardness and grain size exists. It is suggested that thermally-activated phenomena such as grain boundary sliding contributed to the strain-induced softening of this nano-structured biomaterial due to its low melting point. Finally, the current results are compared with those for HPT-processed cast alloys and hybrids of the same composition.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Crystal growth of new high light yield halide perovskite scintillator RbSrI 3

X-ray radiography systems are a key national security technology enabling non-invasive cargo screening. However, the scintillators currently used in these systems have limitations in performance that prompt research for new suitable materials. Here, in this work, crystal growth, physical properties, and optical and scintillation properties of a new halide perovskite scintillator RbSrI 3 were investigated for the first time. The melting point was determined to be 453 °C. Divalent europium and ytterbium were investigated as potential luminescent activators. Light yield and energy resolution of RbSrI 3 :Eu 5 mol% was 78,700 ph/MeV and 2.8 % at 662 keV, and that of RbSrI 3 :Yb 0.5 mol% was 24,000 ph/MeV and 4.9 % at 662 keV.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Formation mechanism of two-dimensional hexagonal silica on SiO 2 /Si substrate

In this study, owing to their remarkable electronic properties, silica ultrathin films have been utilized as an insulating layer in nanoelectronics systems. Silica films have been epitaxially grown on different substrates using various synthesis methods. Among all fabrication approaches, chemical vapor deposition has long been an advanced method for synthesizing two-dimensional (2D) materials due to its ability to ensure precise stacking control and minimize contamination between layers. This study harnessed the potential of CVD to atomically fabricate thin layered 2D silica on a SiO 2 /Si substrate. Significantly, a unique combination of multiple transition metals and salt as the catalysts aided the formation of 2D silica for the first time. Salt is a crucial catalyst in promoting the evaporation of high-melting-point metal catalysts, resulting in hexagonal nucleation sites on the SiO 2 /Si wafer. By meticulously controlling growth parameters, a distinctive hexagonal structure was obtained. Correspondingly, this work delves into the growth mechanism of 2D silica, as evidenced by experiments involving salt alone and individual transition metals. Group VB transition metals played a prominent role in achieving the hexagonal structure compared to their group IVB counterparts. This research offers insight into the formation and growth mechanism of 2D silica, expanding the understanding of silica nanostructures.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Oxidation of ultrahigh temperature ceramics: kinetics, mechanisms, and applications

Materials capable of oxidizing in a protective manner at ultrahigh (>1700 °C) temperatures are needed to push beyond this barrier defined by SiC. Although possessing attractive mechanical properties and oxidation resistance, SiC-based materials are ultimately temperature limited by the melting point of SiO 2 . The vast array of ultra-high and high temperature ceramic literature indicates the majority of these materials, like borides, carbides, MAX-phases, and high-entropy ceramics, fall woefully short regarding oxidation resistance. However, for specific applications, like low-orbit aeropropulsion, high ballistics coefficient atmospheric re-entry, and hypersonic cruise, there are a few promising materials. Here, oxidation criteria are gathered to build application specific heuristics and are then applied to a multitude of ultra-high temperature ceramics to gauge material efficacy. Discussion of oxidation kinetics, mechanisms and reaction products is offered for each material, identifying strengths, weaknesses, and the remaining gaps in our knowledge.

36 MATERIALS SCIENCE↗
DOE OSTI2023

$In-situ$ synchrotron x-ray diffraction and thermal expansion of TiB2 up to ~3050 °C

There is an increasing interest in understanding the performance and properties of ultra-high temperature ceramics due to their high melting points (<3000 °C) that make them promising for extreme environment applications. In-situ high temperature X-ray diffraction experiments were performed on TiB 2 beads up to ~3050 °C. For these experiments, TiB 2 powders were fabricated into spherical beads via gel casting methods and densified in a high temperature graphite furnace. These sample beads were then levitated in a conical nozzle levitator with reducing atmosphere (3% H 2 -Ar) while being heated using a 400 W CO 2 laser. During levitation a collimated synchrotron X-ray source was used to perform in-situ, temperature-dependent structural characterizations. The anisotropic coefficients of thermal expansion of TiB 2 were characterized as a function of temperature up to ~3050 °C. Elucidation of these properties are critical for the advancement of TiB 2 ceramics and other transition metal di-borides for use in high temperature applications such as hypersonic platforms, nuclear reactors, and atmospheric re-entry.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Application of the core shell model for strengthening polymer filament interfaces

While surface segregation has been intensively studied in thin films, relatively little has been investigated in filament geometries. In contrast to thin films, the filament surface forms a continuum, with interactions both at the free air and internal filament interface, which makes it difficult to achieve an equilibrium configuration. This asymmetry has become particularly apparent with the increasing use of additive manufacturing where entire structures are constructed via layered deposition of filamentous layers from a moving thermal nozzle. In addition, this process is highly non-equilibrated with poor thermal retention and is responsible for insufficient filament–filament interfacial fusion and internal pore formation along the printing direction. These features all contribute to poor mechanical and structural properties, which are exacerbated in semi-crystalline polymers, such as Polylactic Acid (PLA), where the melting point, T m , is much higher than the glass transition, T g . Even though system is liquid the molecules become ordered within the crystal structure and this ordering interferes with interdiffusion, which in turn prevents proper interfacial fusion. Capitalizing on the differential in surface tension, we show that addition of less than 1%, or trace amounts, of longer chain polymers, an oriented core shell structure, will be formed with a self-limiting thickness of three time the radius of gyration, R g , of the tracer chains. The lower T g of the tracers and their orientation makes them poised for interdiffusion when subsequent layers are deposited producing structures with significant improvement of the mechanical properties and internal structural integrity.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Nanoscopic structure of borosilicate glass with additives for nuclear waste vitrification

In this report we investigated the nanoscopic structure of borosilicate glasses as a host for high-level radioactive liquid waste (HLLW) in the presence of the additives Na 2 O and CaO/ZnO with and without Li 2 O. These additives have been used to lower the glass melting point, suppress the macroscopic phase separation, and increase the chemical durability of glasses. Small-angle neutron scattering was used to elucidate the effect of the additives on the nanoscopic structure along with macro- and atomic-scale observations obtained using neutron imaging and X-ray diffraction, respectively. The borosilicate glass made without Na 2 O formed a microphase-separated structure of SiO 2 - and B 2 O 3 -rich domains approximately 100 nm in size, and the material was homogeneous at length scales up to millimeters. With the addition of increasing amounts of Na 2 O, the domain spacing decreased. Introducing CaO/ZnO additives induced inhomogeneities in the glass, such as void structures several nanometers in size, although the inhomogeneity was suppressed by the coaddition of Li 2 O. These results provide insights into the performance of host glasses for HLLW vitrification. Incorporating HLLW into glasses is likely to cause changes in the nanoscopic structure of host glasses similar to those revealed here.

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

Viscosity anomaly of a metallic glass-forming liquid under high pressure

Viscosity, as a critical property closely associated with the glass-forming ability of a liquid, has been extensively studied with varying temperatures. However, its pressure dependence has not been well explored yet due to experimental difficulties. Here, we measured the viscosity of a metallic glass-forming liquid, Zr 46 Cu 37.6 Ag 8.4 Al 8 , at pressures up to 6.1 GPa above the melting points by falling-sphere viscometry with ultrafast synchrotron x-ray imaging. Overall, the viscosity increases with pressure, while surprisingly, there is an abrupt drop between 3.2 GPa and 3.7 GPa, indicating the possible existence of a pressure-induced liquid-to-liquid transition. Pressure could change the short- and medium-range orders in the multicomponent glass-forming liquid, as suggested by the different crystalline outcomes after cooling to room temperature at high- and low-pressure ranges. In conclusion, our work extended the viscosity investigation of metallic glass-forming liquids to the high-pressure regime, which will expand our understanding of liquid-liquid transitions and metallic glass formation.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Fabrication and thermophysical properties of UO 2 -UB 2 and UO 2 -UB 4 composites sintered via spark plasma sintering

Uranium dioxide (UO 2 ) composites with uranium diboride (UB 2 ) and uranium tetraboride (UB 4 ) have been proposed as advanced fuel candidates due to their high thermal conductivity, high melting point, high fissile density and their ability to incorporate a built-in burnable poison by tailoring the targeted 10 B/ 11 B ratio. As such, it is important to assess the fabrication, and thermal and micromechanical properties of such composites. In this work, UO 2 -UB 2 and UO 2 -UB 4 samples with boride phase fractions of 5, 15 and 30 wt% were fabricated to high densities (above 95 % theoretical density) via spark plasma sintering (SPS). This enabled sintering at relatively low temperatures and short timescales. SPS also aided in maintaining the target phase fractions of the samples as reactions between the constituent phases were suppressed due to the short timescales and reducing environment during sintering. Here, thermal diffusivity measurements from 299 to 1273 K were conducted through laser flash analysis (LFA). The diffusivity increased as a function of boride weight fraction, and UB 2 additions increased the thermal diffusivity of the composites more than UB 4 additions. Assessment of the LFA results indicated in-situ reactions between the UO 2 and boride phases that suppress the thermal diffusivity occur above 800 K for all samples. Oxidation of the boride phase was proposed as the underlying reaction. This was supported by thermodynamic assessments from the literature, as well as microstructural, crystallographic, and nanoindentation characterization performed on these samples.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Characterization of the molten salt FMgNaK through ab initio molecular dynamics and experimental density measurements

The selection of a salt species for use in a molten salt reactor MSR is a key part of any MSR design. However, many salts have sparse or no thermophysical property data sets, especially those with higher melting points. One such salt is the eutectic mixture of (NaF)0.345(KF)0.59(MgF 2 )0.065, or FMgNaK, and is explored here through ab initio molecular dynamics (AIMD) simulation (1023-1273K) as well as experimental measurement of the liquid density using the Archimedean method (973-1223K). Predicted densities from AIMD are similar to experimentally-measured densities, though with more noise, suggesting that the small scale necessitated by AIMD simulations may be problematic for simulating FMgNaK in particular. Here the coefficient of thermal expansion is predicted from simulation, and salt structure is characterized. Mg-F-Mg chaining is observed in the salt network, though the low concentration of Mg inhibits chaining on the scale that is observed in (LiF) 0.67 (BeF 2 ) 0.33 , or FLiBe.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Postirradiation characterization of palladium as an additive for fuel cladding chemical interaction mitigation in metallic fuel

This work describes the microstructural and elemental characterization of irradiated metallic fuels containing palladium as an additive. The use of additives has been proposed to control Fuel-Cladding Chemical Interaction (FCCI) and thus to promote higher fuel utilization (i.e., higher burnup). In this work, Pd has been investigated as a potential additive to metallic fuel to bind lanthanides, impeding their migration and attack on the cladding. The influence of Pd on the microstructure, chemistry and performance of metallic fuel has been characterized via scanning electron microscopy for two metallic fuel designs—namely, annular and solid fuel. Pd was observed to play an important role in the chemistry of the fuel. Indeed, the addition of Pd leads to the formation of new phases. Pd was detected to combine not only with the lanthanides, as intended, but also with Zr, a main element of the fuel matrix. While Pd proved to be effective in preventing lanthanide migration and their attack on the cladding, the Pd-Zr compound may potentially lead to other unexpected fuel-performance issues, such as the formation of low-melting point phases and increased unalloyed U available for FCCI interaction with Fe in the cladding. Even the increase of Zr to 13wt%. did not completely mitigate this adverse phenomenon generated by the Pd-Zr interaction. Furthermore, the efficacy of using this additive needs further investigation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Molecular dynamics simulations of radiation response of $\mathrm{LiAlO_2}$ and $\mathrm{LiAl_5O_8}$

Here, we report the findings of a computational investigation of defect production and migration in two lithium aluminate ceramics: LiAlO 2 and LiAl 5 O 8 . Although the first ceramic, LiAlO 2 has been widely investigated computationally, especially via molecular dynamics (MD), there have been no MD studies done on LiAl 5 O 8 up to date, due to the lack of interatomic potentials for this system. This work implemented an existing set of Li-Al-O coulombic and Buckingham potential parameters for the LiAl 5 O 8 system from the literature, followed by validating it based on experimental values of density, crystallinity, bond lengths and melting point. The Li + diffusion in the two ceramics was examined and was found to be one order of magnitude slower in LiAl 5 O 8 (D 0 = 3.17 x 10 -11 m 2 /s) as compared to that in LiAlO 2 (D 0 = 4.02 x 10 -10 m 2 /s) at 600 K. The lithium vacancy migration barrier of 4.15 eV in LiAl 5 O 8 was more than three times that in LiAlO 2 (1.31 eV) possibly due to stronger ordering in the Li-O 6 octahedra of LiAl 5 O 8 than in Li-O 4 tetrahedra of LiAlO 2 . The Li displacement threshold energy (E d ) in LiAl 5 O 8 was found to have a median value of 68 eV which is much higher than that in LiAlO 2 (40 eV). The greater E d for Li signifies a lower likelihood for defect formation and clustering and thus lower tendency for amorphization in LiAl 5 O 8 . The presented results show the difference in defect dynamics in the two ceramics that may help us understand the susceptibility of the two ceramics to amorphization caused by irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Ab initio study of tungsten-based alloys under fusion power-plant conditions

Tungsten (W) is considered a leading candidate for structural and functional materials in future fusion energy devices. The most attractive properties of tungsten for magnetic and inertial fusion energy reactors are its high melting point, high thermal conductivity, low sputtering yield, and low long-term disposal radioactive footprint. However, tungsten also presents a very low fracture toughness, primarily associated with intergranular failure and bulk plasticity, limiting its applications. In recent years, several families of tungsten-based alloys have been explored to overcome the aforementioned limitations of pure tungsten. These include tungsten-based high-entropy alloys (W-HEAs) and tungsten-based Self-passivating Metal Alloys with Reduced Thermo-oxidation or “SMART alloys” (W-SAs). Given their proximity to the plasma, it is crucial to understand how the exposure of these candidate plasma-facing materials (PFMs) to the neutron fluxes expected in fusion reactors impacts their material behavior over time. In this work, we present a computational approach that combines inventory codes and first-principles DFT electronic structure calculations to understand the behavior of transmuting tungsten-based PFMs. In particular, we calculate the changes in the chemical composition, production uncertainties, the elastic and ductility properties, and the density of states for five tungsten-based PFMs when exposed to EU-DEMO fusion first wall conditions for ten years.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

A laboratory-scale process for producing dilithium beryllium tetrafluoride (FLiBe) with dissolved uranium tetrafluoride

Flibe Energy, Incorporated (FEI)'s conceptual Lithium Fluoride Thorium Reactor (LFTR) incorporates a chemical processing facility aimed at recovering uranium and other valuable volatile radionuclides while managing harmful radionuclides from the used fuel. The fuel utilized in this reactor is a combination of dilithium beryllium tetrafluoride (Li 2 BeF 4 or FLiBe) and uranium tetrafluoride (UF 4 ), (FLiBe/U). FEI's plan involves extracting the uranium and other valuable volatile fluoride-forming radionuclides using nitrogen trifluoride (NF 3 ). To facilitate laboratory-scale testing of uranium extraction using NF 3 and address the toxicity and physical hazards associated with beryllium and beryllium fluoride (BeF 2 ), we used a two-step process to prepare the simulated fuel salt. The first step entailed thermally decomposing ammonium beryllium tetrafluoride [(NH 4 ) 2 BeF 4 ] (ABeF) through a nominal 3-step process, combined with appropriate amounts of lithium fluoride (LiF) and UF 4 , resulting in the formation of beryllium fluoride (BeF 2 ). In the second step, the mixture was repeatedly melted and frozen at the melting point of FLiBe to prepare the eutectic FLiBe with dissolved UF 4 . Although the concept appears straightforward, the production of FLiBe/U involved various challenges. These challenges included transporting the gaseous decomposition products of ABeF, hydrogen fluoride (HF) and ammonia (NH 3 ), while preventing the formation of ammonium fluoride (NH 4 F). Additionally, it was necessary to control the reaction between the higher-than-anticipated water content in the commercial ABeF with NH 3 , HF, and the condensed NH 4 F, protect UF 4 from forming an unknown black compound, select suitable structural materials to mitigate fluoride corrosion, address the risks associated with beryllium toxicity through equipment design and operational protocols, and monitor process conditions. This article provides an account of the thermal decomposition chemistry observed in the commercial ABeF, describes the FLiBe/U production apparatus, describes the experiences and process refinements developed to prepare FLiBe/U, and presents our characterizations of prepared FLiBe/U.

Ammonium beryllium fluoride thermal decomposition↗
DOE OSTI2025

First-principles investigation of cerium and neodymium diffusion in BCC chromium and vanadium via vacancy-mediated transport

Lanthanide transport plays a crucial role in the performance and longevity of metallic nuclear fuels. This study examines the diffusion behavior of Ce and Nd—two major fission products—in body-centered cubic (BCC) Cr and V, which are potential liner or coating materials for mitigating fuel-cladding chemical interactions (FCCI). Using density functional theory (DFT) calculations and self-consistent mean-field (SCMF) analysis, the vacancy-mediated diffusion coefficients are evaluated. Our findings reveal that Ce and Nd act as oversized solutes and are strongly bound to vacancies in BCC Cr and V, with diffusivities in Cr significantly lower than in V and in hexagonal closed-packed (HCP) Zr, as investigated in our previous work. The activation energies for Ce and Nd diffusion are 3.39 and 3.32 eV, respectively, in BCC Cr, and 2.56 and 2.33 eV, respectively, in BCC V. Analysis of vacancy drag and partial diffusion coefficient ratios indicates a strong tendency for lanthanide enrichment at vacancy sinks in BCC Cr, and to a lesser extent in BCC V, with this effect persisting up to the melting point in Cr and remaining substantial for Nd in V at high temperatures. Under irradiation, the increase in vacancy concentration is expected to enhance lanthanide transport, potentially accelerating interactions at liner-cladding interfaces. Although BCC Cr exhibits relatively low lanthanide diffusivities under equilibrium conditions, the expected segregation tendencies under irradiation suggest that Zr liners may be a more favorable option. Further investigations using rate theory, cluster dynamics, and phase-field modeling are required to quantitatively assess the performance of these materials in reactor environments.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

Ab-initio informed cluster dynamics simulation of self- and Xe diffusivity in uranium mononitride under irradiation

Uranium mononitride (UN) is one of the ceramic nuclear fuel alternatives to oxide fuels considered for light water reactor and advanced reactor designs, as it presents significant advantages such as high uranium density (better economics) and high thermal conductivity and melting point (increased safety). Self- and fission gas diffusivities need to be better understood, given that they influence key fuel performance phenomena like swelling and fission gas release. Recently, radiation enhanced diffusivity was investigated in UN by means of cluster dynamics simulations relying on empirical potential-based parameterizations, the reliability of which highly depends on the interatomic potential accuracy. Here, in this work, we refine this approach by determining, using ab-initio calculations, the properties of defect clusters containing vacancies, self-interstitials and Xe impurities. We also consider larger clusters than previous studies. The obtained dataset (formation enthalpies, entropies, and migration barriers) is used to parameterize a cluster dynamics model of mobile clusters, and to calculate the defect cluster concentrations under irradiation. This gives us access to the radiation enhanced self- and fission gas diffusivities. Although the resulting diffusivities are close to the values reported in the literature, we find important qualitative differences in the diffusion mechanisms. Capturing the correct mechanisms is crucial to properly describe the chemistry and fission rate dependence of the model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Density functional theory calculations of the mixing enthalpy of ternary uranium carbide compounds

The high melting point of uranium-zirconium carbides (U,Zr)C makes them an ideal fuel for nuclear thermal propulsion (NTP) reactors. Gaps remain in the current understanding of the U-Zr-C system due to the difficulty of conducting thermodynamic experiments at NTP operation conditions. Density functional theory calculations using the Hubbard U model (DFT+U) were performed using orbital matrix occupation (OMC) to obtain the mixing enthalpy for UC and ZrC for (U,Zr)C ternary compounds. Similarly, DFT+U calculations were also carried out for the (U,Nb)C and (U,Ta)C systems. In conclusion, the DFT results are envisioned to be used in thermodynamic assessments of the uranium carbide systems based on the CALPHAD approach to supplement the lack of experimental data for the mixing thermodynamics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Hot Hydrogen Exposure of U x Zr 1-x C y Nuclear Fuel: The Influence of Composition and Density

Refractory carbide nuclear fuel has been one of the most promising fuel candidates for space nuclear propulsion due to its high melting point, temperature stability, and compatibility in a hot hydrogen environment. In this study, U x Zr 1-x C y fuel was produced by means of a carbothermic reduction process in different UC compositions including 5,10, 20, and 30 at.% UC in the fuel compound. The powder feedstock was consolidated via direct current sintering with densities up to 97% of the theoretical density. The samples with different U x Zr 1-x C y compositions were exposed to hot hydrogen at 2600 K for a cumulative time of 300 min. The samples were characterized by SEM, XRD, density, and measured for mass losses. The high-density samples displayed improved performance in hot hydrogen by minimizing porous sites and reducing areas of direct contact with hydrogen gas, leading to reduced mass losses. Variations in sample density proved to induce large changes in mass loss rates, increasing them up to 90%. The compositions with higher UC content reported the largest mass losses in the study. The loss of uranium occurred primarily at the surfaces exposed to the hot hydrogen where changes in the lattice constant confirmed losses exceeding 50% of the initial UC content in higher compositions, specifically to U 0.3 Zr 0.7 C y . XRD analyses revealed the presence of UH 3 in U 0.3 Zr 0.7 C y suggesting that metallic uranium formed inside the sample as a product of carbon losses. High-density U x Zr x-1 C y fuel with UC concentrations at or below 20 at.% UC exhibited stability and negligible density changes in a high temperature hydrogen environment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗