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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 1,027 records · Page 57
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

A computational study of yttria-stabilized zirconia: II. Cation diffusion

Cubic yttria-stabilized zirconia is widely used in industrial electrochemical devices. While its fast oxygen ion diffusion is well understood, why cation diffusion is much slower—its activation energy (~5 eV) is 10 times that of anion diffusion—remains a mystery. Indeed, all previous computational studies predicted more than 5 eV is needed for forming a cation defect, and another 5 eV for moving one. In contrast, our ab initio calculations have correctly predicted the experimentally observed cation diffusivity. We found Schottky pairs are the dominant defects that provide cation vacancies, and their local environments and migrating path are dictated by packing preferences. As a cation exchanges position with a neighboring vacancy, it passes by an empty interstitial site and severely displaces two oxygen neighbors with shortened Zr-O distances. This causes a short-range repulsion against the migrating cation and a long-range disturbance of the surrounding, which explains why cation diffusion is relatively difficult. In comparison, cubic zirconia's migrating oxygen only minimally disturbs neighboring Zr, which explains why it is a fast oxygen conductor.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Preferential diffusion in concentrated solid solution alloys: NiFe, NiCo and NiCoCr

In single-phase concentrated solid-solution alloys (CSAs), including high entropy alloys (HEAs), remarkable mechanical properties are exhibited, as well as extraordinary corrosion and radiation resistance compared to pure metals and dilute alloys. But, the mechanisms responsible for these properties are unknown in many cases. In this work, we employ ab initio molecular dynamics based on density functional theory to study the diffusion of interstitial atoms in Ni and Ni-based face-centered cubic CSAs including NiFe, NiCo and NiCoCr. We model the defect trajectories over >100 ps and estimate tracer diffusion coefficients, correlation factors and activation energies. Furthermore, we found that the diffusion mass transport in CSAs is not only slower than that in pure components, i.e. sluggish diffusion, but also chemically non-homogeneous. The results obtained here can be used in understanding and predicting the atomic segregation and phase separation in CSAs under irradiation conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Radiation instability of equal channel angular extruded T91 at ultra-high damage levels

Although several studies have shown that nano-grained alloys have higher radiation resistance than coarse-grained variants, with dense grain boundary structures acting as defect sinks to remove damage, the present study shows the failure of severe deformation to suppress void swelling in T91 alloy at ultra-high damage levels. Fe self-ion irradiation up to 1000 peak dpa produced significantly more swelling in extruded T91 than conventional T91 at 475 °C. Dramatic radiation-enhanced grain growth was observed in the extruded T91, manifested in two rapid grain growth stages. One stage occurs at a relatively low damage level, and the other at a higher damage level beyond 750 peak dpa. The second rapid growth stage is correlated with void swelling and starts at the end of the swelling incubation period. The study suggests that the defect sink properties of the growing grain boundaries under a coarsening process are different from that of stable boundary configurations. When void swelling becomes significant and voids act as a biased sink for vacancies, the growing grain boundaries act as a biased sink for interstitials, leading to rapid growth at higher damage levels.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Dislocation loop evolution during in-situ ion irradiation of model FeCrAl alloys

Model FeCrAl alloys of Fe-10%Cr-5%Al, Fe-12%Cr-4.5%Al, Fe-15%Cr-4%Al, and Fe-18%Cr-3%Al (in wt %) were irradiated with 1 MeV Kr++ ions in-situ with transmission electron microscopy to a dose of 2.5 displacements per atom (dpa) at 320 °C. In all cases, the microstructural damage consisted of dislocation loops with ½< 111 > and <100 > Burgers vectors. The proportion of ½< 111 > dislocation loops varied from ~50% in the Fe-10%Cr-5%Al model alloy and the Fe-18Cr%-3%Al model alloy to a peak of ~80% in the model Fe-15%Cr-4.5%Al alloy. The dislocation loop volume density increased with dose for all alloys and showed signs of approaching an upper limit. The total loop populations at 2.5 dpa had a slight (and possibly insignificant) decline as the chromium content was increased from 10 to 15 wt %, but the Fe-18%Cr-3%Al alloy had a dislocation loop population ~50% smaller than the other model alloys. As a result, the largest dislocation loops in each alloy had image sizes of close to 20 nm in the micrographs, and the median diameters for all alloys ranged from 6 to 8 nm. Nature analysis by the inside-outside method indicated most dislocation loops were interstitial type.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Modelling recrystallization textures driven by intragranular fluctuations implemented in the viscoplastic self-consistent formulation

This work presents a recrystallization model driven by intragranular orientation gradients and strain energy fields calculated by means of the viscoplastic self-consistent (VPSC) formulation. The VPSC model is extended for calculation of the coupling between intragranular stress fluctuations with corresponding second moments of lattice spin and misorientation fields in the grains. Access to these quantities allows modelling of transition bands and nucleation kinetics. In the proposed recrystallization model, grain growth is assumed to be proportional to the difference between the stored energy of each grain and that of the effective medium. Recrystallization textures for several cubic metals are simulated, showing good agreement with corresponding experiments. The model reveals the importance of considering appropriate, microstructurally-based and orientation-dependent recrystallization nucleation mechanisms. The recrystallization texture of heavily rolled copper with a strong cube texture component is found to be a consequence of nucleation at transition bands, which is also the cause of the recrystallization textures in compressed iron and drawn copper wire. Finally, in contrast, the recrystallization texture of rolled interstitial-free steel is found to be caused by grain boundary nucleation occurring in grains with the highest strain energy.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Unprecedented irradiation resistance of nanocrystalline tungsten with equiaxed nanocrystalline grains to dislocation loop accumulation

Nanocrystalline metals are often postulated as irradiation tolerant materials due to higher grain boundary densities. The efficiency of these materials in mitigating irradiation damage is still under investigation. Here, we present an in-situ transmission electron microscopy with ion irradiation study on equiaxed 35 nm grained tungsten (NCW-35 nm) and compare its radiation tolerance, in terms of dislocation loop damage, to several other grades of tungsten with different grain sizes at two temperatures (RT and 1073 K). The NCW-35 nm was shown to possess significant higher radiation tolerance in terms of loop damage. As demonstrated by Kinetic Monte Carlo simulations, at least part of the higher radiation tolerance of the small grains is due to higher interstitial storage (at the grain boundaries) and defect recombination (in the grain interiors) in the small grain material. In addition, experimental observations reveal rapid and efficient dislocation loop absorption by the grain boundaries and this is considered the dominant factor for mass transport to the boundaries during irradiation, enabling the remarkable radiation tolerance of 35 nm grained tungsten. Furthermore, this study demonstrates the possibility of attaining high radiation tolerant materials, in terms of dislocation loop damage, by minimizing grain sizes in the nanocrystalline regime.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Microstructure and microchemistry study of irradiation-induced precipitates in proton irradiated ZrNb alloys

Proton irradiation induced Nb redistribution in Zr-xNb alloys (x = 0.4, 0.5, 1.0 wt%) has been investigated using scanning transmission electron microscopy/energy dispersive X-ray spectroscopy (STEM/EDS). Zr-xNb alloys are mainly composed of Zr matrix, native Zr–Nb–Fe phases, and β-Nb precipitates. After 2 MeV proton irradiation at 350 °C, a decrease of Nb content in native precipitates, as well as irradiation-induced precipitation of Nb-rich platelets (135 69 nm long and 27 12 nm wide) were found. Nb-rich platelets and Zr matrix form the Burgers orientation relationship, [$1\bar{1}1$]//[$2\bar{1}\bar{1}0$] and (011)//(0002). The platelets were found to be mostly coherent with the matrix with a few dislocations near the ends of the precipitate. The coherent strain field has been measured in the matrix and platelets by the 4D-STEM technique. Here, the growth of Nb-rich platelets is mainly driven by coherency and dislocation-induced strain fields. Irradiation may both enhance the diffusion and induce segregation of interstitial Nb to the ends of the irradiation induced platelets, further facilitating their growth.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

A phase-field model for hydride formation in polycrystalline metals: Application to δ-hydride in zirconium alloys

We report a phase-field model for simulating metal hydride formation involving large volume expansion in single- and polycrystals. As an example, we consider δ-hydride formation in α-zirconium (Zr), which involves both displacive crystallographic structural change and hydrogen diffusion process. Thermodynamic Gibbs energy functions are extracted from the available thermodynamic database based on the sublattice model for the interstitial solid solutions. Solute-grain boundary interactions and inhomogeneous elasticity of polycrystals are taken into consideration within the context of diffuse-interface description. The stress-free transformation strains of multiple variants for hcp -Zr (α) to fcc -hydride (δ) transformation are derived based on the well-established orientation relationship between the α and δ phases as well as the corresponding temperature-dependent lattice parameters. In particular, to account for the large volume expansion, we introduced the mixed interfacial coherency concept between those phases—basal planes are coherent and prismatic planes are semi-coherent in computing the strain energy contribution to the thermodynamics. We analyzed the morphological characteristics of hydrides involving multiple structural variants and their interactions with grain boundaries. Moreover, our simulation study allows for the exploration of the possible hydride re-orientation mechanisms when precipitating under applied tensile load, taking into account the variation in the interfacial coherency between hydrides and matrix, their elastic interactions with the applied stress, as well as their morphology-dependent interactions with grain boundaries. Here, the phase-field model presented here is generally applicable to hydride formation in any binary metal–hydrogen systems.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Neutron irradiation-induced microstructure damage in ultra-high temperature ceramic TiC

Titanium carbide (TiC) is an ultra-high temperature ceramic with potential as a structural material candidate for advanced reactor concepts. However, the irradiation tolerance of TiC is not well understood. Here, we reveal the key irradiation damage microstructure degradation processes in TiC using mixed spectrum neutron irradiations at dose of ~2 displacements per atom (dpa) at temperatures of ~220, 620, and 1115 °C, combined with state-of-art microstructure characterization using transmission electron microscopy (TEM) and high-resolution TEM (HRTEM). The helium injection (~65 atomic parts per million) produced by neutron transmutation also occurred in the sample. TiC was observed to form irradiation-induced interstitial-type dislocation loops and He-stabilized cavities. At 220 and 620 °C, the analysis of the electron diffraction patterns, rel-rod imaging and HRTEM revealed that the dislocation loops were faulted Frank loops with Burgers vector b Frank = 1/3<111> lying on {111} planes. A detailed Burgers vector identification performed by the g.b technique revealed that the dislocation loops forming at 1115 °C were unfaulted, edge-type, with Burgers vectors b = a<100> and a/2<110> with corresponding {100} and {110} habit planes. Using continuum mechanics, we estimated the critical radius at which a faulted dislocation loop transitions to a perfect loop to be 9 nm. Further, no amorphization occurred in TiC under-investigated irradiation conditions while macroscopic swelling under point defect swelling regime was observed. He-stabilized cavities were detected at 1115 °C, but not at lower temperature irradiation. These observations indicate the onset of long-range vacancy migration in TiC occurs between 620 and 1115 °C.

36 MATERIALS SCIENCE↗
DOE OSTI2020

In situ microstructural evolution in face-centered and body-centered cubic complex concentrated solid-solution alloys under heavy ion irradiation

This study characterizes the microstructural evolution of single-phase complex concentrated solid-solution alloy (CSA) compositions under heavy ion irradiation with the goal of evaluating mechanisms for CSA radiation tolerance in advanced fission systems. Three such alloys, Cr 18 Fe 27 Mn 27 Ni 28 , Cr 15 Fe 35 Mn 15 Ni 35 , and equimolar NbTaTiV, along with reference materials (pure Ni and E90 for the Cr-FeMnNi family and pure V for NbTaTiV) were irradiated at 50 K and 773 K with 1 MeV Kr ++ ions to various levels of displacements per atom (dpa) using in-situ transmission electron microscopy. Cryogenic irradiation resulted in small defect clusters and faulted dislocation loops as large as 12 nm in face-centered cubic (FCC) CSAs. With thermal diffusion suppressed at cryogenic temperatures, defect densities were lower in all CSAs than in their less compositionally complex reference materials indicating that point defect production is reduced during the displacement cascade stage. High temperature irradiation of the two FCC CSA resulted in the formation of interstitial dislocation loops which by 2 dpa grew to an average size of 27 nm in Cr 18 Fe 27 Mn 27 Ni 28 and 10 nm in Cr 15 Fe 35 Mn 15 Ni 35 . This difference in loop growth kinetics was attributed to the difference in Mn-content due to its effect on the nucleation rate by increasing vacancy mobility or reducing the stacking-fault energy.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Structural and chemical disorder enhance point defect diffusion and atomic transport in Ni 3 Al-based γ' phase

The ordered L1 2 γ' Ni 3 Al intermetallic inclusions are routinely used as strengthening constituents in Ni-based superalloys and recent high-entropy alloys. As there are different metallic elements relative to the alloy matrix, the composition of γ' phase inclusions usually suffers some variations. In this work, we studied the elemental partition and its influence on atomic transport properties in a model (NiCo) 3 Al γ' phase with chemical and structural disorder. Combining the Monte Carlo and molecular static techniques, we first determined the elemental distribution in such compositionally-complexed γ' phase. The results suggest that Co tends to segregate, which profoundly influences defect energetics and diffusion properties. By modeling thermally activated diffusion by molecular dynamics, we found that chemical disorder induced by Co can enhance vacancy diffusion while significantly suppress diffusion via interstitial atom migration, as compared to the stoichiometric γ' Ni 3 Al. We further reveal that these observations are closely related to the defect energetics in these systems. Furthermore, our results explain the effects of disorder, long- and short-range orders on the diffusion properties in multicomponent intermetallic alloys, which helps to understand the stability and, thus, γ'-phase-assisted strengthening at different conditions including annealing and irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Mechanisms of Ti 3 Al precipitation in hcp α -Ti

Here, nucleation and growth of Ti3Al α( 2 ) ordered domains in alpha-Ti-Al-X alloys were characterised using a combination of transmission electron microscopy, atom probe tomography and small angle X-ray scattering. Model alloys based on Ti-7Al (wt.%) and containing O, V and Mo were aged at 550 °C for times up to 120d and the resulting precipitate dispersions were observed at intermediate points. Precipitates grew to around 30nm in size, with a volume fraction of 6-10% depending on tertiary solutes. Interstitial O was found to increase the equilibrium volume fraction of α 2 , while V and Mo showed relatively little influence. Addition of any of the solutes in this study, but most prominently Mo, was found to increase nucleation density and decrease precipitate size and possibly coarsening rate. Coarsening can be described by the Lifshitz-Slyozov-Wagner model, suggesting a matrix diffusion-controlled coarsening mechanism (rather than control by interfacial coherency). Solutionising temperature was found to affect nucleation number density with an activation energy of E f = 1.5 +/- 0.4 eV, supporting the hypothesis that vacancy concentration affects α 2 nucleation. The observation that all solutes increase nucleation number density is also consistent with a vacancy-controlled nucleation mechanism.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Unveiling the interaction of nanopatterned void superlattices with irradiation cascades

Nanopatterned microstructures in materials can have a profound impact on materials’ physical and chemical properties. While voids are typically considered as detrimental defects in irradiated materials, the patterning of nanoscale voids causes the formation of void superlattices and provides a highly efficient mechanism for gas storage. Despite the important applications of nanopatterned defect superlattices, how they degrade under irradiation remains unclear. Here we provide direct observation of the evolution of void superlattices under irradiation and elucidate the interaction of void superlattices with irradiation cascades. We reveal that the instability of void superlattices under irradiation is caused by heterogenous void shrinkage and demonstrate the imperative role of mixed 1D/3D diffusion of self-interstitial atoms and injected inert gas atoms on void shrinkage and void superlattice instability. Understanding the degradation mechanisms of nanopatterned microstructures is essential to designing damage-tolerant materials and broadening their applications in extreme environments.

36 MATERIALS SCIENCE↗
DOE OSTI2022

A combined theoretical-experimental investigation of thermal transport in low-dose irradiated thorium dioxide

During reactor operation, nuclear fuels are subject to extreme temperature and irradiation conditions which can significantly degrade the fuel's thermal transport properties. The reduction in thermal conductivity of the fuel as a result of irradiation-induced lattice defects is arguably the most important fuel performance metric in regard to reactor efficiency and safety. Because thorium dioxide (ThO2) is suitable as a model system for more complex materials such as UO2 and its mixed oxides, we present a theoretical investigation of thermal conductivity reduction seen in defect-bearing thorium dioxide and compare directly to experimental measurements. Phonon-mediated thermal transport of the fuel is modeled by a solution to the Boltzmann transport equation (BTE) for phonons. A cluster dynamics (CD) model for lattice defect evolution during irradiation predicts defect densities which are used as input to the BTE for modeling phonon-defect scatterings. Phonon scatterings by lattice defects include those from point defects and vacancy clusters and interstitial clusters of various sizes. The CD model is benchmarked against structural defect characterization of irradiated thorium dioxide using electron microscopy. Thermal conductivity predicted by the BTE model is compared to measured values for irradiated thorium dioxide specimens below room temperature to isolate effects of phonon-defect scattering from intrinsic 3-phonon processes, which dominate at higher temperatures. The computed conductivity values are in partial agreement at temperatures close to room temperature while slight deviations are observed at the lowest measured temperatures, suggesting that implemented phonon-defect scattering cross-section expressions may not be adequate for low temperatures. The presented work provides a necessary investigation of the influence of irradiation induced defects on fuel performance and represents a first step toward a full characterization of phonon mediated thermal transport in irradiated materials with complex defect microstructure.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Enhanced resistance to helium irradiations through unusual interaction between high-entropy-alloy and helium

Finding high performance plasma-facing materials (PFMs) is one of the most important and challenging tasks for realizing the commercial application of fusion reactors. Herein, we found the CrMoTaWV high entropy alloy (HEA) is highly resistant to low-energy and high-flux He plasma exposure. The nanochannel HEA film has 20 times higher initial fluence for the formation of fuzz and a remarkable 8.9 times slower fuzz growth rate than those of W. Combining the in-situ TEM observation and the Molecular dynamics (MD) simulation of the He bubble growth process, a new mechanism for the enhanced radiation resistance in HEA with the unusual interaction between HEA and He is found, where, differing from traditional metal, bubble growth in HEA leads to non-directional emission of interstitial atoms while HEA greatly suppress the growth of He bubbles. Additionally, the special nanochannel structure further rise the radiation resistance through releasing He out of the HEA film and reducing the He concentration. This new nanochannel refractory HEA material presents a promising choice as the PFMs with excellent performance and a much longer serving lifetime for future commercial fusion reactors.

36 MATERIALS SCIENCE↗
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

Mechanisms for high creep resistance in alumina forming austenitic (AFA) alloys

Castable alumina forming austenitic (AFA) alloys have demonstrated superior creep life and oxidation resistance at temperatures exceeding 800⁰C. Despite the success in the applicability of these alloys in extreme environments, there is a limited understanding of the deformation modes and the influence of each alloying element guiding the alloy design strategies that could further enhance the creep strength of these AFA alloys, particularly at temperatures at and above 900⁰C. In this study, we reveal the mechanism underpinning the superior creep performance of castable AFA alloys that involves suppressing primary carbide formation through minor compositional modification. This approach results in a three-fold increase in creep strength at 900⁰C and 50 MPa. Here, by employing integrated characterization techniques, we analyzed the microstructures of two AFA alloys, both before and after the creep process. We discovered that the suppression of primary carbides permits the in-situ clustering of now-available interstitial elements such as C, Si, and O during high-temperature creep. This improved solid solution strengthening and reduced stacking fault energy of the alloy. Moreover, it also enabled controlled secondary carbide formation during testing, further improving the creep resistance. These findings underline the important interplay between alloy composition, microstructure, and creep properties, and offer a promising design strategy for developing economical high-temperature Fe-based alloys suitable for advanced applications.

36 MATERIALS SCIENCE↗