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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 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Effects of processing temperature, pressure, and fiber volume fraction on mechanical and morphological behaviors of fully-recyclable uni-directional thermoplastic polymer-fiber-reinforced polymers

This work explores a type of composite called thermoplastic polymer-fiber-reinforced polymers (PFRPs), often referred to as self-reinforced composites (SRCs). A representative PFRP was exemplified using unidirectional (UD) ultra-high-molecular-weight polyethylene (UHMWPE) fibers embedded in a high-density polyethylene (HDPE) matrix. The effects of compression molding temperature and pressure on the mechanical and morphological behaviors of the filament-wound PFRPs with various fiber volume fractions (V f ) were experimentally investigated. The results elucidate the evolution of morphologies and tensile properties of the PFRPs due to thermal melting, fiber misalignment from pressure, and (V f )-induced structural variance, which has not been comprehensively reported yet. The highest specific tensile strength and modulus of the PFRP laminae reach 600 MPa/(g/cm 3 ) and 31 GPa/(g/cm 3 ), respectively. These properties are comparable to glass-/aramid-fiber-reinforced polymers (GFRPs, GFRTPs, AFRPs, and AFRTPs), with PFRPs exhibiting better ductility (specific strain at peak load ≈ 4%/(g/cm 3 )) than other common polymer composites. The motivation for this work was the high recyclability of PFRPs, which can be recycled by melting both the fibers and the matrix, and then reshaped them for re-manufacturing composites to maximize the efficiency in material reuse. This process simplifies the implementation of closed-loop recycling, re-manufacturing, and reuse to support sustainability in composites. This work aims to contribute to advancing thermoplastic PFRPs for their potential applications in various industries.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Electrochemical behavior of SnCl 2 and influence of Cu and Ni ions in molten LiCl−KCl−CaCl 2 eutectic

Reliable transport and thermodynamic data for multivalent ions in complex molten salts are scarce, limiting model fidelity for electrorefining and impurity control. Here, we report a comprehensive electrochemical characterization of SnCl₂ in LiCl–KCl–CaCl₂ (50.5–44.2–5.3 mol%) at 685 K, including the effects of Ni 2+ and Cu + impurities. Using cyclic voltammetry (CV), chronoamperometry (CA), and chronopotentiometry (CP), we quantified Sn 2+ and Ni 2+ diffusion with exceptional agreement across methods: Sn 2+ averaged (1.03 ± 0.10) × 10 −5 cm 2 s −1 , and Ni 2+ averaged (0.75 ± 0.19) × 10 −5 cm 2 s −1 . The tight confidence-interval overlap across CV, CA, and CP strengthens confidence in these values and is uncommon in molten chloride studies. Open-circuit-potential measurements provided standard apparent reduction potentials that closely match LiCl–KCl literature, indicating minimal shift with CaCl₂ present. The Sn 2+ /Sn couple behaves as a reversible two-electron soluble–insoluble process at 685 K; the Sn 4+ /Sn 2+ couple transitions to soluble–soluble behavior near 788 K, which may correlate with the decomposition of surface bound chlorostannates, though direct characterization remains to be established. In mixed systems, Cu+/Cu overlaps Sn 2+ /Sn, limiting Cusingle bondSn electroseparation, whereas the larger potential gap between Ni 2+ /Ni and Sn 2+ /Sn supports selective Ni removal. These internally consistent transport and thermodynamic data establish a validated basis for process modeling and optimization of Sn electrorefining and impurity management in LiCl–KCl–CaCl₂.

Berzins-Delahay↗
DOE OSTI2025

Influence of temperature, oxygen partial pressure, and microstructure on the high-temperature oxidation behavior of the SiC Layer of TRISO particles

Tristructural isotropic (TRISO)-coated fuel particles are designed for use in high-temperature gas-cooled nuclear reactors, featuring a structural SiC layer that may be exposed to oxygen-rich environments over 1000 °C. Surrogate TRISO particles were tested in 0.2–20 kPa O 2 atmospheres to observe the differences in oxidation behavior. Oxide growth mechanisms remained consistent from 1200–1600 °C for each P O$_2$ , with activation energies of 228 ± 7 kJ/mol for 20 kPa O 2 and 188 ± 8 kJ/mol for 0.2 kPa O 2 . At 1600 °C, kinetic analysis revealed a change in oxide growth mechanisms between 0.2 and 6 kPa O2. In 0.2 kPa O 2 , oxidation produced raised oxide nodules on pockets with nanocrystalline SiC. Oxidation mechanisms were determined using Atom probe tomography. Active SiC oxidation occurred in C-rich grain boundaries with low P O$_2$ , leading to SiO 2 buildup in porous nodules. Here, this phenomenon was not observed at any temperature in 20 kPa O 2 environments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Direct observation of annealing-driven recrystallization behavior in magnesium alloy at low strain condition

Grain-twin interactions are significant in texture modification under thermodynamic driving force. In this study, annealing-driven twinning/detwinning behavior, grain growth, and corresponding texture evolution in a pre-deformed AZ31 magnesium alloy were systematically tracked and investigated via in-situ heating synchrotron X-ray diffraction and quasi in-situ electron backscattered diffraction techniques. A twinning texture is generated in the pre-deformed sample due to the activation of {$10\bar{1}2$} tensile twinning. During annealing, dislocation annihilation occurs between 100 and 280 °C, and recrystallization occurs above 280 °C, manifesting as the initial residual matrix and twins being competitively swallowed by each other, forming a bimodal texture. The recrystallization process is completed by boundary movement, which depends on the energy difference across the boundary. In addition, it is found that the grain boundaries favor movement towards the side with higher stored energy, regardless of the boundary type or the boundary energy.

In-situ observation↗
DOE OSTI2025

Magnetic structure and magnetodielectric behavior of the chiral magnet CoTeMoO 6

We investigate the magnetic structure and magnetodielectric behavior of the chiral magnet CoTeMoO 6 (CTMO) through neutron diffraction, magnetization, and magnetodielectric measurements, complemented by density functional theory (DFT) calculations. Our findings reveal a canted magnetic structure with moments confined to the ab plane, giving rise to weak ferromagnetism under an external magnetic field. Additionally, we observe magnetodielectric coupling that strongly correlates with the magnetic ordering temperature and magnetic structure. Furthermore, these results are discussed in the context of potential mechanisms involving spin-dependent p-d hybridization and spin-phonon coupling.

Canted magnetism↗
DOE OSTI2025

Non-linear elastic-plastic behavior of the invert glass lithium phosphorous oxynitride (LiPON)

Micro and nano-scale mechanical behavior of network/ionic glasses is dictated by their composition and the number of constraints per structural unit. From this perspective, the glass LiPON presents the opportunity for enhancement of the microscale ductility due to its rather unconstrained orthophosphate structure. Here, we use instrumented nanoindentation with different tip geometries to investigate the mechanical response of LiPON glass. The results reveal that the elastic modulus of LiPON is not constant and depends on pressure. With the method utilizing spherical nanoindentation and continuous stiffness measurement (CSM) we determine the yield point of LiPON. We propose the Drucker-Prager type of yield criterion for LiPON and estimate its yield stress in compression as 2.4 GPa. There exists another stress threshold, however, around 490 MPa, at which the elastic deformation becomes non-linear, and this can be mistaken for the yield stress.

Constitutive model↗
DOE OSTI2024

Enhanced Stress Relaxation Behavior via Basal ⟨a⟩ Dislocation Activity in Zircaloy-4 Cladding

Herein this work evaluates the stress relaxation behavior of textured Zircaloy-4 cladding to understand how mechanical anisotropy influences pellet-cladding interactions. Uniaxial and biaxial stress relaxation tests are performed using full-tube axial tension and internal pressurization, respectively, aiming to achieve 0.25%, 1%, and 2% equivalent strains in the cladding samples at a temperature of 300°C. Internal pressure relaxation test results display enhanced stress relaxation compared to axial testing results, particularly for samples loaded beyond yield. Results of electron backscatter diffraction indicate increased deformation microstructure during loading and increased strain homogenization and recovery during relaxation for samples loaded via internal pressurization. Analysis indicates that the increased production and activity of basal ⟨a⟩ dislocations play a significant role in the enhanced relaxation measured in samples subjected to internal pressurization.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Changes in dislocation punching behavior due to hydrogen-seeded helium bubble growth in tungsten

The accumulation of gas atoms in tungsten is a topic of long-standing interest to the plasma-facing materials community due the metal's use as a divertor material in some tokamak fusion reactors. The nucleation and growth of He/H gas bubbles (along with their isotopes) can result from impinging fluxes of these gases which give rise to damage at the W divertor surface. The inclusion of He or H in W has been studied extensively by the community, finding that He bubbles modify the surface through periodic dislocation punching and bursting mechanisms while H bubbles impact the metal through plastic-strain induced material failure. However, the mechanisms which are present during the combined flux of both He and H is not well-studied atomistically. Motivated by this, an atomistic modeling study is conducted using molecular dynamics to assess the behavior of mixed concentration He:H bubbles in W. Here we find that the introduction of H into a growing He bubble results in a dramatic change in the nature and presence of dislocation loops which are typically generated via dislocation punching in over-pressurized He bubbles. Most notably, at high H concentrations, there is a switchover in energetic favorability from glissile 1/2<111> dislocations to sessile <100> dislocations. This thermodynamic crossover could imply a significant reduction in W surface morphology changes than with pure He bubbles and, additionally, we show this to have implications on the trapping of H in the bubbles and their associated dislocations.

36 MATERIALS SCIENCE↗
DOE OSTI2024

In-situ strain behavior and BISON simulations of Zircaloy cladding subjected to temperature cycling separate-effects tests in a steam environment

Understanding fuel system performance during anticipated transients without scram (ATWSs) in boiling water reactors (BWRs) is necessary for refining current and future safety limits. High-fidelity material models and simulations are fundamental to rigorous assessment of zirconium-based cladding performance. However, experimental thermomechanical data during simulated ATWSs to validate these modes are limited. To provide relevant in-situ data, Zircaloy-4 cladding was subjected to cyclic heating in a steam environment to simulate an out-of-pile BWR ATWS. Digital image correlation was used to capture the cladding strain behavior in-situ for comparison against simulations using the BISON finite element code. Conventional high-temperature models were compared using multiple schemes to gain a better understanding of the applicability of three BISON models to BWR ATWS: (1) the default combination of creep models in BISON, (2) the high-temperature Erbacher model alone, and (3) the low-temperature Limback-Andersson model alone. The cases run with the Limback-Andersson model alone produced the lowest root mean square error (RMSE). The lowest RMSE for the Limback-Andersson model alone was 0.659%, and the highest RMSE reported was 4.22%. A data gap within the model in the temperature regime of interest was also identified, and to account for this gap, the current model in BISON is linearly interpolated between two separate datasets. In conclusion, this evaluation highlights the need to either develop a new model or to improve the existing model to capture transient creep effects resulting from a cyclic temperature transient.

Anticipated transient without SCRAM↗
DOE OSTI2024

Fracture behavior of irradiation induced nanocrystalline UO 2 studied by in-situ mechanical testing in transmission electron microscopy

Uranium Dioxide (UO 2 ) is widely used as a fuel in current light water reactors (LWRs). Upon accumulation of radiation damage, LWR UO 2 fuel pellets start to develop a different microstructure at the pellet periphery when fuel burnup exceeds 45–50 GWd/tHM. The resulting porous, nanocrystalline microstructure is one of the most prominent microstructural changes occurring in such fuel. Its fracture mechanisms, which causes fuel fine fragmentation, could impact safety limits when the cladding breaches. Direct measurements of these properties are challenging, therefore a surrogate obtained via ion irradiation can be used. In this study, multiple microcantilevers were fabricated by focused ion beam from both fresh UO 2 and UO 2 irradiated with 84 MeV Xe 26+ ions to a peak dose of 1357 displacements per atom (dpa). Further, the irradiation produced a pseudo high burnup structure approximately 2 µm below the surface. In-situ nano-mechanical bending tests were conducted to investigate the fracture behavior and the effect of the surrogate UO 2 high burnup structure on local fracture properties. Fresh UO 2 fuel was observed to fracture in transgranular mode without nucleation or movement of dislocations. However, the Xe-irradiated nanocrystalline microcantilevers fractured along the grain boundaries, with no influence from the pre-existing micro-cracks in the microcantilever. Fracture toughness for this type of surrogate high burnup UO 2 structure is reported for the first time in literature. Both the fracture stress and toughness show degradation for UO 2 as a result of Xe-irradiation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Effect of cold forging on the microstructure and corrosion behavior of type 316L stainless steel in molten FLiNaK salt

Here, the effect of cold forging on the microstructure and corrosion behavior of 316L stainless steel (SS) in molten salt was investigated. Static corrosion experiments were performed in FLiNaK (LiF-NaF-KF: 46.5–11.5–42 mol.%) at 600 °C for 50 h in a glove box. The results show that cold forging gives rise to enhanced corrosion of 316L SS in molten fluoride salt due to the increase of crystallographic defects. Based on the potentiodynamic polarization results, the corrosion current density of 50 % cold-forged 316L SS is about 2.1 times larger than that of the as-received 316L SS in molten FLiNaK salt at 600 °C.

316L stainless steel↗
DOE OSTI2025

MX precipitate behavior in an irradiated advanced Fe-9Cr steel: Helium sequestration and cavity swelling performance

This work is the third and final part in an initial series on addressing the behavior of MX precipitate stability in an advanced Fe-9Cr reduced activation ferritic/martensitic (RAFM) alloy under fusion-relevant ion irradiation conditions. Here, the helium trapping properties of MX precipitates are investigated across varying damage levels (15–100 dpa), temperatures (400–600 °C), and helium doses (10–25 appm He/dpa) using sophisticated dual ion beam experiments and electron microscopy. Results indicate that individual MX precipitates efficiently sequester helium in the form of nanoscale bubbles at the precipitate-matrix interfaces near the peak swelling temperature (∼5 bubbles/precipitate at 500 °C). Swelling was primarily due to matrix cavities. The Fe-9Cr alloy reached 2% swelling by 100 dpa, suggesting a shift to steady-state swelling around 50 dpa at 500 °C. Furthermore, MX precipitate dissolution beginning at 15 dpa did not coincide with this onset of steady-state swelling.

Dual ion irradiation↗
DOE OSTI2025

Mechanical properties, strain hardening, and fracture behavior of ultrasonic additively manufactured Zircaloy-4 after low-temperature neutron irradiation

Ultrasonic additive manufacturing (UAM) is a solid-state, layer-by-layer advanced manufacturing process that has the potential to create custom spatially controlled composites with embedded wires and sensors for nuclear component manufacture. For this work, to assess the feasibility of using UAM for nuclear-relevant materials research, the technique was used to produce a 3.5-mm-thick Zircaloy-4 plate for irradiation testing. The UAM Zircaloy-4 specimens were irradiated in the High Flux Isotope Reactor at a target irradiation temperature of 117 °C to 2.9 displacements per atom (dpa) to assess differences in irradiation-hardening behavior as a function of alloy processing path. The UAM and reference baseplate (BP) materials increased in yield strength by 372±27 MPa and 346±21 MPa, respectively, and both suffered significant reductions in uniform and total elongation attributed to irradiation hardening at low-temperature. Although the materials had similar nanoscale defect structures, including nanoscale black dot/loop features and strain-induced dislocation channels, the UAM material’s processing-related defects resulted in accelerated strain localization and failure as demonstrated by lower post-irradiation uniform elongation of UAM specimens (0.5 %) compared to BP (1.5 %) material. The UAM material also showed considerable anisotropy in mechanical response due to crack propagation along weld boundaries, resulting in differences in strength & ductility when tested parallel and perpendicular to the prior UAM build orientation. Therefore, although the fundamental irradiation response of UAM-processed Zircaloy-4 was phenomenologically comparable to that of BP reference material, additional optimization of the UAM processing is needed to produce irradiation-resistant and nuclear-relevant materials.

Digital image correlation↗
DOE OSTI2026

Molecular dynamics simulations of reflection and sputtering behavior of boron under deuterium ion irradiation

Boronization is a commonly used method of wall conditioning in fusion reactors. The application of boron films to the plasma-facing materials results in enhanced plasma performance due to the reduction of intrinsic impurities. This is primarily driven by a reduction in oxygen content that is chemically trapped in the boron film. The reactive nature of these boron films also raises questions concerning interactions with hydrogen isotopes. In this work, boron-deuterium interactions were studied using molecular dynamics (MD). Reactive force field potentials were used to model the chemical interactions between B and D. An amorphous boron substrate was irradiated by D atoms at varying incident energies, 10 eV < E i < 150 eV and angles, 0° < α < 85°. The reflection probability was calculated and compared to results from the commonly-used binary collision approximation (BCA) method. This comparison found that the BCA underestimated the reflection probability at E i < 35 eV and α > 45 ∘ . The source of this discrepancy was found to be the surface binding energy model. The BCA calculation with an isotropic surface binding energy model was more closely aligned to the MD result. This, in combination with a correction function based on the MD results allows for corrections to the reflection probability of deuterium impinging on boron surfaces. The sputtering of the substrate material was also studied. While this study did not contain sufficient events to quantitatively describe the sputtering behavior, some qualitative results emerged: namely, chemical sputtering of B and D-containing molecules (BD, BD 2 , BD 3 ) at low ( < 20 eV) incident deuterium energies. This result suggests that chemical sputtering could be a significant factor in limiting boron coating lifetime when exposed to lower ion energies, such as those in detached plasmas. The results show that chemical interactions should be taken into account when modeling the interactions between D ions and B surfaces.

Boron↗
DOE OSTI2024

Understanding ion-selective Li/Na metal plating behavior in hybrid Li-Na battery

This study investigates ion-selective Li/Na metal plating behavior in hybrid Li-Na battery systems, revealing the critical role of electrolyte solvents in these processes. Using a hybrid battery design with a LiFePO 4 cathode, Na metal anode, and NaPF 6 -based electrolytes, we observed contrasting effects of carbonate- and ether-based electrolyte solvents. While ether-based electrolytes showed expected Na plating/stripping, carbonate-based electrolytes surprisingly favored a Li-dominant plating/stripping reaction despite the Na-rich environment. X-ray photoelectron spectroscopy revealed that this selectivity is linked to the composition of the solid electrolyte interphase (SEI) layer, with carbonate electrolytes forming Li-based inorganic-rich SEI layers that facilitate Li-ion diffusion while screening Na ions. In conclusion, these findings challenge the conventional understanding of metal plating in multi-ion environments and offer insights for designing future hybrid battery systems.

25 ENERGY STORAGE↗
DOE OSTI2025

Microstructure and mechanical behavior of laser remelted amorphous Al-Ni-La welds

To explore the feasibility of using laser powder bed fusion systems to print bulk amorphous Al alloys, a single autogenous weld produced by laser remelting on a cast Al-5La-9Ni (at.%) alloy was studied for its microstructure and mechanical behavior. The solidification rates experienced by material within the welds were high enough to produce entirely amorphous regions within the welds. Welds were characterized using SEM, STEM, and APT and were found to contain Ni-rich amorphous clusters. Micropillar compression tests were used to assess mechanical properties of the welds and found that all of the amorphous regions exhibited yield strengths around 1 GPa. In conclusion, these results are discussed in the context of diffusivity of Ni and rare earth elements in liquid, glass forming ability, thermodynamic driving force for phase formation during solidification, free volume concentrations in bulk metallic glasses, and cluster-related softening.

APT↗
DOE OSTI2024

Effect of iodides on thermal behavior and phase partitioning in LiCl-KCl

Liquid-fueled molten salt reactors (MSRs) are designed to operate with fissile materials and, ultimately, fission products dissolved in the primary molten salt coolant. Understanding the speciation and transport of iodine—a high-yield fission product—is essential because this element’s accidental release poses significant environmental concerns due to its capacity to be readily absorbed by the human thyroid gland. Here, we report the impact of iodide species (LiI and KI) on phase transitions, phase distribution, and phase stability in LiCl–KCl-eutectic salt mixtures. The study employed a combination of computational and experimental techniques, including thermodynamic FactSage calculations, differential scanning calorimetry, and high-temperature X-ray diffraction. The results indicate that the presence of iodide (10–25 wt%) significantly alters the melting behavior of the LiCl–KCleutectic system. Adding 10 wt% LiI has a more-pronounced effect than 10 wt% KI, as LiI converts to KI, leading to formation of LiCl, thereby, altering the LiCl-KCl ratio which significantly affects the melting temperature of the mixture. Furthermore, the evolution of crystalline structure, solid-fraction composition, and the dynamics of mixed-halide solid–liquid partitioning as a function of temperature indicate the potential for selective iodide separation from chloride-salt mixtures via solid–liquid separation techniques. Overall, the presented findings provide valuable insights that are beneficial for the design and operation of MSRs, as well as for the safe handling and effective processing of used nuclear fuel using advanced pyrochemical techniques.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Influence of post deposition annealing on the microstructural evolution and tensile behavior of austenitic stainless-steel alloy 709 made by laser powder bed fusion

Structural materials used in the next generation nuclear reactors or other advanced energy systems are often required to withstand higher operating temperatures for a longer service life. Alloy 709 (A709), a Nb-containing and N-stabilized Fe-25Ni-20Cr austenitic stainless steel, has been recommended for high-temperature reactor constructions due to the better performance over the baseline Type 316H stainless steel. Here, in the current manuscript, for the first time, the microstructure evolution and tensile behavior of an A709 made by laser powder bed fusion (LPBF) is discussed. The as deposited and heat treated LPBF alloys are compared with a reference wrought alloy. Owing to the chemical segregation and a higher number of dislocations, the LPBF materials show a significantly higher density of precipitates formed during high temperature treatments when compared to the wrought samples. This is reflected when looking at the tensile properties, both at room and elevated temperatures, where LPBF samples show higher strengths compared to wrought materials.

36 MATERIALS SCIENCE↗