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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 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Degradation of electrical resistivity of tungsten following shielded neutron irradiation

A major challenge for heat transfer in nuclear materials is to ensure thermal mobility after high amounts of neutron irradiation. Tungsten is widely selected as a heat transfer material in fusion reactors. In metals, thermal conductivity is dominated by electrons’ ability to transfer energy. Neutron irradiation generates point defects, clusters, and solid transmutation (e.g.rhenium and osmium in tungsten), which inhibit electron motion. The purpose of this work is to quantify the irradiation-induced change in electron mobility and deconvolute transmutation and microstructural effects on observed changes to electron mobility. Single and polycrystalline tungsten were fast neutron irradiated in the High Flux Isotope Reactor at Oak Ridge National Laboratory to doses between 0.2 and 0.7 displacements per atom (dpa) and temperatures from 500 °C to 1000 °C. Grain growth was observed in all samples. Microstructure and transmutation were quantified. The geometric orientation of samples with elongated grains has been shown to affect electrical resistivity. A mathematical model was developed and used to deconvolute solid-solution transmutation, grain, and temperature-dependent lattice effects on resistivity. At ~0.4 dpa at ~590 °C, the combined resistivity degradation due to voids, vacancies, interstitials, and dislocations is estimated to be greater than the contribution from solid solution Re transmutation, which is greater than the contribution from grain boundaries. At doses of ~0.7 dpa at ~750 °C, solid solution Re contributions are greater than all other effects combined. As a result, this work establishes a basis to predict the effects of irradiation temperature and transmutation on thermal properties of tungsten and highlights the importance of irradiation temperature.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Surface grain refinement of casting A380 aluminum alloy by ultrasonic-assisted surface mechanical grinding treatment

A novel surface processing technique, ultrasonic-assisted surface mechnical grinding treatment (U-SMGT), was proposed and applied to modify the surface micriostructure of the casting aluminum alloy (A380). Comprehensive post-process chracterization techniques including 3D profilometry, optical microscope, scanning electric microscopy (SEM), and electron backscatter diffraction (EBSD) were carried out to investigate the surface morphology and microstructure of the processed region (PR). It was found that a surface layer of grain refinement with a thickness of about 50 µm was formed. In the as-received base material (BM), large clusters of dendritic eutectic phase with a size of approximately 20 µm were isolated by α-Al matrix. Nano-indentation tests showed that hardness of the eutectic and α-Al in the as-received BM was 2.15 GPa and 1.2 GPa respectively. After U-SMGT, the large dendritic eutectic phase was broken down into much fine particles with an average size of 0.57 µm. These fine particles were uniformly distributed into the α-Al matrix. Nano-indentation measurement exhibited a hardness of 2.27 GPa within the processed region. The overall high hardness in the processed region is a result of the pinning effect of uniformly distributed fine dendrites and grain refinement of Al-matrix.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Correlated Ion Transport Governed by Dynamic Local Structure in High Concentration and Localized High Concentration Electrolytes

Understanding the dynamics of cluster formation and network percolation provides the mechanistic link between microscopic solvation structure and transport in concentrated electrolytes, including localized high-concentration electrolytes (LHCEs). Although recent studies have shown LHCEs to form micelle-like aggregates at specific compositions, a quantitative understanding of how solvation structures and transport properties depend on salt–solvent–diluent ratios remains limited. Here, we integrate molecular dynamics with Onsager transport analyses to chart the evolution of solvation microstructure and associated ionic transport in LiFSI/DMC/TTE, achieving good agreement with experimental conductivities across composition. We show that micelle formation arises from a dynamic instability of the cation–anion network, and that network percolation is the primary determinant of conductivity. Two compositional thresholds emerge: A critical network concentration (CNC) and a critical micelle concentration (CMC) that delineate transitions from extended percolating networks to micelle-like clusters and then to fragments. These structural transitions rationalize the nonintuitive conductivity decrease at intermediate dilution despite monotonically decreasing viscosity and provide composition-level design rules for LHCEs.

Mohanakrishnan, Rohith Srinivaas [University of Ca↗
DOE OSTI2020

Chapter 21. Atomic-Level Dislocation Dynamics in Irradiated Metals

Primary damage and microstructure evolution in structural nuclear materials operating under conditions of a high flux of energetic atomic particles and high temperature and stress lead to the formation of a high concentration, non-homogeneous distribution of defect clusters in the form of dislocation loops, voids, gas-filled bubbles and radiation-induced precipitates of nanometer scale. They cause changes in many material properties. Being obstacles to dislocation glide, they strongly affect the mechanical properties, with an increase in yield and flow stresses and a reduction in ductility. Atomic-scale computer simulations can provide details of how these effects are influenced by the obstacle structure, applied stress, strain rate and temperature. Processes such as obstacle cutting, transformation, absorption and drag are observed. Some recent results for body-centered and face-centered cubic metals are described in this review and, where appropriate, comparisons are drawn with predictions based on the elasticity theory of crystal defects. Perspectives on how to use this information at higher scales and in particular in mesoscale, dislocation dynamics simulations are also discussed.

Osetskiy, Yury N.↗
DOE OSTI2021

Molecular dynamics simulations of radiation cascade evolution near cellular dislocation structures in additively manufactured stainless steels

The dislocation dense cellular structures observed in additively manufactured (AM) stainless steels can allow these materials to have enhanced strength and ductility compared to conventionally manufactured materials. Effective design of radiation resistant materials often necessitates creating microstructures containing of a high density of dislocations that act as sinks for radiation induced defects. This work uses molecular dynamics simulations to study the impact of pre-existing defects on radiation damage in stainless steel 316L fabricated by the laser powder bed fusion process. The evolution of synthetically generated dislocations in a crystal system in response to multiple radiation collision cascades suggest that dislocation dense regions reduce the probability of surviving point defects forming clusters. This finding seems to result from the dislocation cores’ ability to absorb point defects, thereby limiting the formation of clusters during the recombination phase. However, after successive cascades, the ability of the dislocation entanglement to prevent defect formation is diminished and defect cluster formation trends to an equilibrium. Accordingly, it is hypothesized that in an experimental setting the dislocation cells will initially act as neutral sinks for point defects, potentially delaying the onset of radiation damage effects; but, these structures will be degraded due to radiation-enhanced diffusion and lose their effectiveness at doses above 1.5–2 displacements per atom. Nevertheless, the ability to produce complex and feature-specific microstructures suggests promise for the design of radiation tolerant materials using AM methods.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Development of Corrosion Resistant Coatings for Structural Materials for Liquid Fueled Molten Salts Reactors Applications (Final Report)

Current structural alloys, code certified for the MSRs temperature ranges, contain high levels of chromium, making them highly susceptible to molten halide salt corrosion. One potential solution to circumvent the need for code certification of novel alloys, which is expensive and time consuming, is to design claddings that will protect the underlying code certified materials from corrosion damage during operation. In this work, we examined the corrosion of Ni and Cu electroplated, Ni and CuNi weld overlay, Mo-laser clad, and carburized claddings on SS316H for use in molten salt reactor environments. After characterization of the cladded materials (Task 1), static corrosion tests were used to assess corrosion resistance of the claddings in typical molten fluoride salts (Task 2). The corrosion tests were performed in molten FLiNaK at 700°C up to 1000 hours. Pre- and post-corrosion Scanning Electron Microscopy (SEM), energy-dispersive-spectroscopy (EDS), Transmission Electron Microscopy (TEM), X-ray Diffraction (XRD) and glow-discharge-optical-emission-spectroscopy (GDOES) were performed on cladding cross-sections and surfaces to evaluate degradation. The Cu and Ni electroplated samples, as well as the carburized samples, showed excellent corrosion resistance relative to the bare SS316H. To assess high temperature cladding stability, ageing experiments were performed at temperatures up to 900 C in inert atmosphere for the electroplated and carburized samples and a diffusion model was developed to predict long term cladding behavior (Task 3). It was found that the Cu cladding was basically insensitive to the high temperature ageing, except for small secondary phases forming at the interface. On the other hand, the Ni electroplated cladding experienced significant interdiffusion. Nevertheless, the gain in corrosion resistance for a 100μm Ni electroplated cladding is phenomenal, with more than 50% reduction in chromium dissolution from the substrate material for the first 15 years of salt exposure at 700°C. To assess radiation resistance and phase stability, high-temperature 4MeV Ni heavy ion irradiation was performed across the cladding/substrate interface up to 50 displacement-per-atom (DPA) at 500°C and 700°C to assess the phase stability and irradiation behaviors of the electroplated systems (Task 4). The Ni and Cu electroplated systems did not experience void swelling at the contrary to the SS316H substrate due to their nanocrystalline nature. Some level of recrystallization was observed in the cladding, as well as radiation induced segregation and enhanced diffusion. The interface acted as a potent sink for point defects with the presence of a void denuded zone. The mechanical properties of the claddings were assessed using thermal shock, micro-indentation, nano-indentation, and four-point bend testing experiments (Task 5). These experiments were performed on the claddings in as-received, corroded, irradiated, and thermally aged states to determine the effects of typical molten salt reactor environments on cladding integrity. No significant mass loss was observed after repeated thermal shocks. While the electroplated samples softened after high-temperature ageing, irradiation hardening compensate this effect, such that there is little different with the as-received materials. The results of these experiments suggest that the electroplated (copper/nickel) show the greatest promise for application in molten salt reactor development. These claddings prevented any chromium dissolution from occurring during the static corrosion experiments. Additionally, they demonstrated favorable properties for high temperature diffusion, phase stability, interfacial mechanical properties, and irradiation resistance. Weld-overlay cladding are also of great interest and could reach properties similar to the electroplated samples upon optimization (especially with multiple weld passes). The carburized SS316H showed excellent behavior as well, but more work needs to be done to assess their long-term stability. Finally, the Mo-laser clad system was not pursued further due to the manufacturing challenges. It is also worth noting that the Ni cladding systems should behave relatively well in terms of weldability since the Ni-weld overlay microstructure and associated corrosion rate are sound. While the NiCu weld-overlay has even lower corrosion rates than the Ni-weld overlay, more studies on welding of Cu-electroplated systems are necessary since Cu clusters are known to embrittle steels.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Microclusters of Kinked Silicon Nanowires Synthesized by a Recyclable Iodide Process for High-Performance Lithium-Ion Battery Anodes

Silicon shows great promise as a high-capacity anode material for lithium-ion batteries. Nanostructuring silicon minimizes the impact of fracturing during charging and discharging. However, synthesizing nanostructured silicon typically requires complicated procedures with high manufacturing costs. Additionally, these complicated procedures typically have poor secondary particle formation, a requirement to achieve a high tap density. Here, a cost-effective synthesis procedure which generates nearly ideal secondary particle clusters of nanostructured silicon is developed. The cost-effectiveness is a result of the in operando generation of silicon iodide from iodine gas and low-grade silicon microparticle. Decomposition of silicon iodide into crystalline silicon and iodine gas enables recycling of the iodine gas, allowing for near full reutilization of iodine in the following cycles. Additionally, the optimal nanostructures and microstructures of silicon synthesized by the recyclable iodide decomposition reaction enables 83.6% capacity retention over 1000 cycles. The good performance is a result of well-maintained morphology during cycling, enabling reutilization of the solid electrolyte interphase.

25 ENERGY STORAGE↗
DOE OSTI2022

Development of grain-scale slip activity and lattice rotation fields in Inconel 718

Using a combination of in-situ high-resolution digital image correlation (HR-DIC), Heaviside-DIC method (H-DIC), and crystal plasticity finite element (CPFE), we investigate the evolution of intragranular lattice rotations and slip activity during monotonic and cyclic loading in a high performance, polycrystalline face centered cubic material. The CPFE employs a quasi-3D model microstructure, which is a highly resolved mirror representation of the experimental in-situ test sample. In agreement, the measurements and calculations reveal that most grains, regardless of their size and lattice orientation, develop intragranular lattice rotation gradients that span the grain. For a small cluster of grains on the deformed material, we perform HR-DIC analysis of slip lines to demonstrate agreement in the active slip systems and changes in this local slip activity across the individual grains. The combined analysis reveals that deforming grains are divided into sub-granular regions of uniform lattice rotation and these regions are most often associated with only one or two active slip systems. The gradient lines that divide them correspond to changes in the predominant slip system. The model is used to examine the evolution of intragranular lattice rotation in a single fully reversed tension-compression cycle. The calculations indicate that intragranular gradients intensify during the reverse loading path as nearest neighboring regions appear to shed lattice rotation, increasing the lattice rotation in some regions, while shutting down rotation in neighboring regions. So these findings provide insight into the irreversible changes that develop within deforming grains at the scale of the grain, particularly the heterogeneous development of intragranular lattice rotation in early stages of deformation, which could serve as precursors to localization.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Helium causing disappearance of a/2<111> dislocation loops in binary Fe-Cr ferritic alloys

In this work, single and dual-beam self-ion irradiations were performed at 500°C on ultra-high purity Fe14%Cr alloy to ~0.33 displacements-per-atom (dpa), and 0 or 3030 atomic-parts-per-million (appm) helium/dpa, respectively. Using transmission electron microscopy, we reveal that helium can drastically modify the dislocation loop Burgers vector in Fe-Cr alloys. Helium co-implantation caused complete disappearance of a/2<111> type dislocation loops, and the microstructure consisted of only a<100> loops. Conversely, a/2<111> type loops were predominant without He co-implantation. The total loop density remained largely unaffected. The results strikingly contrast literature asserting that helium stabilizes a/2<111> type loops in bcc Fe alloys, based on low temperature irradiations. Collectively analyzing the results with literature suggest that the small positive interaction between helium and self-interstitial atoms (SIA) in Fe predicted by atomistic simulations maybe insufficient to holistically explain the dislocation loop microstructure development in presence of helium. Helium-SIA positive binding inadvertently implies elevated a/2<111> loop fraction and higher loop densities that the present results contradict. Helium induced high cavity density causing a preferential loss of highly glissile <111> clusters, leaving the matrix saturated with <100> type clusters is proposed as a potential mechanism. Further, the in-situ irradiations combined with Burgers vector analysis strengthened the evidence of Cr-induced dislocation loop mobility reduction that appears to stabilize the a/2<111> type loops and causes higher loop densities in Fe-Cr alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2020

A new mechanism for void-cascade interaction from nondestructive depth-resolved atomic-scale measurements of ion irradiation–induced defects in Fe

The nondestructive investigation of single vacancies and vacancy clusters in ion-irradiated samples requires a depth-resolved probe with atomic sensitivity to defects. The recent development of short-pulsed positron beams provides such a probe. Here, we combine depth-resolved Doppler broadening and positron annihilation lifetime spectroscopies to identify vacancy clusters in ion-irradiated Fe and measure their density as a function of depth. Despite large concentrations of dislocations and voids in the pristine samples, positron annihilation measurements uncovered the structure of vacancy clusters and the change in their size and density with irradiation dose. When combined with transmission electron microscopy measurements, the study demonstrates an association between the increase in the density of small vacancy clusters with irradiation and a remarkable reduction in the size of large voids. This, previously unknown, mechanism for the interaction of cascade damage with voids in ion-irradiated materials is a consequence of the high porosity of the initial microstructure.

36 MATERIALS SCIENCE↗
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 OSTI2025

Characterization and Quantification of Radiation-Induced Clusters/Precipitates in RPV Steels Using STEM-EDS and Machine Learning

Over the operational lifespan of a nuclear reactor, reactor pressure vessel (RPV) steels are subjected to significant neutron irradiation, resulting in complex microstructural changes and the consequent degradation of mechanical properties. Various physically motivated correlation models have been developed to predict neutron irradiation-induced embrittlement of RPVs under different irradiation conditions. However, the efficient and accurate characterizations and quantification of radiation-induced clusters in RPVs are still challenging, which will affect the precision of the predictive models for embrittlement of RPV components. In the DOE Visiting Faculty Program (VFP) research work at Oak Ridge National Lab (ORNL), I integrate machine learning to aid Scanning Transmission Electron Microscopy – Energy Dispersive X-ray Spectroscopy (STEM-EDS) analyses, which improve the characterization and quantification of radiation-induced clusters in RPV steels, thereby enabling more accurate predictions of material behavior under irradiation. The surveillance base- and welded- RPV steels were annealed at various temperatures of 340 °C, 450 °C and 500 °C for up to 168 hours, respectively. Afterwards, I have characterized radiation-induced clusters using advanced STEM-EDS techniques and subsequently applying machine learning algorithms to analyze and refine STEM-EDS datasets, enhancing the quantification of clusters compositions and distributions. In the end, an efficient workflow for integrating STEM-EDS data analysis with machine learning to address challenges including noise reduction has been developed. The completion of this VFP work will support bridge critical gaps in the accurate quantification of radiation-induced clusters in RPV steels using STEM-EDS and support the development of more precise models for predicting RPV embrittlement in the Light Water Reactor Sustainability program supported by Department of Energy and enhancing the collaboration between ORNL and Alred University. The outcome of the VFP project will leverage a few research papers submission to peer-reviewed journals in the relevant scientific field and a few oral presentations at national and international conferences.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Microstructural Engineering of Cu-Rich Nanoprecipitate formation in NiCoFeCrCu0.12 High-Entropy Alloy via Severe Plastic Deformation for Enhanced Irradiation Tolerance

This study demonstrates a defect-engineering approach for controlling Cu-rich precipitates in FeNiCrCoCu0.2 high-entropy alloys (Cu-HEAs), delivering a novel pathway for next-generation nuclear reactor materials with superior irradiation resistance. This work establishes that severe plastic deformation (SPD) processing via Shear Assisted Processing and Extrusion (ShAPE) and Friction Stir Layer Deposition (FSLD) creates dense dislocation networks and subgrain boundaries that fundamentally alter precipitation behavior under identical thermal treatments. Atom probe tomography (APT) indicates that SPD produces a metastable, atomically homogeneous solid solution that, upon moderate heat treatment (500°C/10 hour), develops remarkedly stronger Cu clustering than the as-cast counterpart. High-temperature exposure (800°C/100 h) produces near-pure Cu precipitates (~90 at% Cu) with significantly enhanced defect-sink efficacy in SPD-processed alloys: precipitate sizes of 50-60 nm and number densities of 2.7-3.8 × 10¹7 m?³, compared to 89 nm and 0.44 × 10¹7 m?³ in as-cast materials. Collectively, the findings establish defect-mediated precipitation control as a scalable, high-impact route to tailor sink density and distribution in HEAs, enabling microstructures optimized for irradiation tolerance and mechanical robustness in nuclear reactor environments.

Meher, Subhashish↗
DOE OSTI2024

A spatially-resolved model of neutron-irradiated tungsten coupling stochastic cluster dynamics and finite deformation plasticity

Structural materials used in nuclear reactors face severe degradation in mechanical properties, such as hardening and embrittlement. At the microscopic scale, this occurs due to creation and accumulation of irradiation-induced defects and their interaction with system dislocations. Although techniques exist which can model evolution of irradiation defects, for instance kinetic transport theory-based models, their interaction with mechanical deformation of the bulk material has not been investigated extensively. In this work, we demonstrate a novel spatially-resolved multiscale coupling between microscopic irradiation defect evolution, modeled using Stochastic Cluster Dynamics (SCD) and macroscopic mechanical deformation modeled using a finite-deformation plasticity model. SCD is used to determine the statistically averaged defect cluster spacing, dependent on operating conditions such as irradiation dose and temperature. This acts as an initial condition that governs the critical resolved shear stress of dislocation glide in the macroscopic plasticity model. This framework is used to predict mechanical behavior in post-mortem test of irradiated Tungsten samples, which has found its importance as structural material used in nuclear reactors. The results obtained using the coupled approach are in good agreement with experimental data of uniaxial tension tests. The model is able to capture the effect of temperature and irradiation dose on the material hardening. Two methods are proposed to estimate hardness – using Tabor's Law relating uniaxial yield stress to hardness and from flat-punch simulations. The results are in reasonable agreement with hardness data from micro-indentation experiments of irradiated Tungsten samples. Finally, the model is also able to reveal microstructural details such as spatial variation in defect density and local stress.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Revealing the Sodium Storage Mechanisms in Hard Carbon Pores

Abstract Hard carbon (HC) is the most promising anode for the commercialization of sodium‐ion batteries (NIBs); however, a general mechanism for sodium storage in HC remains unclear, obstructing the development of highly efficient anodes for NIBs. To elucidate the mechanism of sodium storage in the pores, operando synchrotron small‐angle X‐ray scattering, wide‐angle X‐ray scattering, X‐ray absorption near edge structure, Raman spectroscopy, and galvanostatic measurements are combined. The multimodal approach provides mechanistic insights into the sodium pore‐filling process for different HC microstructures including the pore sizes that are preferentially filled, the extent to which different pore sizes are filled, and how the defect concentration influences pore filling. It is observed that sodium in the larger pores has an increased pseudo‐metallic sodium character consistent with larger sodium clusters. Furthermore, it is shown that the HCs prepared at higher pyrolysis temperatures have a larger capacity from sodium stored in the pores and that sodium intercalation between graphene layers occurs simultaneously with the pore filling in the plateau region. Opportunities are outlined to improve the performance of HC anodes by fully utilizing the pores for sodium storage, helping to pave the way for the commercialization of sodium ion batteries.

Kitsu Iglesias, Luis↗
DOE OSTI2021

Thermal gradient effect on helium and self-interstitial transport in tungsten

First-wall materials in a fusion reactor are expected to withstand harsh conditions, with high heat and particle fluxes that modify the materials microstructure. These fluxes will create strong gradients of temperature and concentration of diverse species. Besides the He ash and the hydrogenic species, neutron particles generated in the fusion reaction will collide with the material creating intrinsic defects, such as vacancies, self-interstitials atoms (SIAs), and clusters of such point defects. These defects and the He atoms will then migrate in the presence of the aforementioned gradients. In this study, we use nonequilibrium molecular dynamics to analyze the transport of He and SIAs in the presence of a thermal gradient in tungsten. We observe that, in all cases, the defects and impurity atoms tend to migrate toward the hot regions of the tungsten sample. The resulting species concentration profiles are exponential distributions, rising toward the hot regions of the sample, in agreement with irreversible thermodynamics analysis. For both He atoms and SIAs, we find that the resulting species flux is directed opposite to the heat flux, indicating that species transport is governed by a Soret effect (thermal-gradient-driven diffusion) characterized by a negative heat of transport that drives species diffusion uphill (from the cooler to the hot regions of the sample). Here, we demonstrate that the steady-state species profiles obtained accounting for the Soret effect vary significantly from those where temperature-gradient-driven transport is not considered and discuss the implications of such a Soret effect on the response to plasma exposure of plasma-facing tungsten.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Automated identification of deformation twin systems in Mg WE43 from SEM DIC

In this study, the application of machine learning and computer vision approaches to microscale deformation data for the automated identification of deformation twinning systems, and their associated twin area, is introduced. Deformation data was obtained during in-situ SEM compression testing of a WE43 Mg alloy using digital image correlation (DIC) modified for use with electron microscopy. A 5.7 mm × 3.4 mm area of interest was analyzed, generating ~100 million data points of deformation. Experimental twin trace directions were determined by applying k-means clustering, morphological thinning, and Hough transforms to the deformation data. The identification of deformation twinning was achieved by consideration of the twin trace direction, its strain value, and its evolution. The deformation map was divided into areas corresponding to individual active twin systems, enabling the analysis of microstructure dependence of deformation behavior and twinning activity. The performance of the proposed twinning identification approach was evaluated by accuracy, precision, and sensitivity metrics. The effect of the tuning parameters used in the algorithm on the performance is also discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2026

First-principles insights into chromium-induced oxide phases in NiO

The high-temperature oxidation of Ni–Cr alloys leads to complex oxide scales comprising Ni(Cr)O solid solutions, NiCr 2 O 4 spinel, and Cr 2 O 3 corundum within the NiO matrix. Understanding the atomic-scale mechanisms of Cr segregation and precipitate formation is crucial for enhancing oxidation resistance. Here, we employ density functional theory calculations to investigate Cr behavior on NiO(100), (110), and (111) surfaces and in the bulk. Our results reveal that isolated Cr atoms preferentially segregate to the surfaces, stabilizing Ni(Cr)O solid solutions via strong Cr–O bonding, whereas Cr pairs and clusters favor subsurface migration and bulk aggregation, promoting nucleation of NiCr 2 O 4 and Cr 2 O 3 phases. These findings elucidate a size-dependent segregation mechanism linking Cr coordination environments to oxide phase evolution. This atomic-scale insight informs strategies to tailor oxide microstructures and enhance the high-temperature oxidation resistance of Ni–Cr alloys.

Chromium Segregation↗