Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Dislocation”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,675 records · Page 93
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Microstructures and Impact Wear Behavior of Al-Alloyed High-Mn Austenitic Cast Steel After Aging Treatment

The microstructural evolution and impact wear behavior of Al-alloyed ultra-high-Mn austenitic cast steel were investigated after different heat treatments, including quenching (Q) and quenching + aging (Q + A). Another ultra-high-Mn austenitic cast steel without Al was compared under the same experimental condition. After Q + A, Al addition contributed to the precipitation of κ-carbides which improves the yield strength and wear resistance of the austenite matrix. However, the mechanical properties and wear resistance of the Al-free steel were strongly deteriorated due to the precipitation of plate-like M{sub 3}C-type carbides, as well as the worn surface of the Al-free steel had the feature of micro-scratches. After Q + A (2 h), the Al-alloyed steel obtained optimal wear resistance and the main wear feature was strain fatigue. But, excessive aging time (4 h) would lead to a large fragile peeling from the worn surface and the wear resistance was decreased. Also, the worn subsurface microstructures with different heat treatments were observed by transmission electron microscopy (TEM). These results show that the deformation microstructures of the Al-free steel contained stacking faults, dislocation tangle and deformation twins. The deformation character of Al-alloyed steel is dominated by planar glide, and with increase in the aging time, planar dislocation substructures gradually evolved, relating to dislocation pileups, Taylor lattices and micro-bands.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Atomic-scale mechanisms of void strengthening in tungsten

Neutron and heavy-ion irradiation of tungsten produces nanometer-size vacancy voids, gas-filled bubbles and dislocation loops. These defect features can affect mechanical properties and the impact can be quite significant because of their high density. Understanding the basic mechanisms of mechanical properties degradation is necessary for predicting radiation effects. Predictions can be made using discrete dislocation dynamics or/and finite element approaches which, however, need local interaction mechanisms as inputs. Such knowledge can be provided only by atomic-scale modeling. This paper reports the results of an extensive atomic-scale modeling study of the interactions between moving edge dislocations and voids in tungsten. Furthermore, the main focus is on the effects of the void size and ambient temperature. Critical resolved shear stress was calculated for voids up to 9 nm in diameter. Atomistic results are compared with the theoretical approach and with those obtained earlier for voids in body centered cubic (bcc) iron. An important role of the void surface has been revealed.

36 MATERIALS SCIENCE↗
DOE OSTI2011

Structure of Cr Film Epitaxially Grown on MgO (001)

Cr(001) films of thickness ~100 nm were epitaxially grown on MgO (001) using molecular beam epitaxy at both room temperature and 550 °C. The structure of the films was characterized using high resolution transmission electron microscopy, electron diffraction, and electron energy-loss spectroscopy. Room temperature growth leads to highly oriented columnar structured grains in the film, while growth at 550 °C yields high quality single crystal Cr film, possessing a 45 degree rotational epitaxial relationship between the film and the substrate. For the single crystal film, the interface exhibited coherent regions separated by equally-spaced misfit dislocations. When imaged from the MgO[100]//Cr[110] direction, the dislocation spacing is 5.38 nm and 4.64 nm for the films grown at room temperature and 550 °C, respectively, which is slightly smaller than the expected value of 6.25 nm calculated from the differences in lattice spacing based on known lattice constants. The results presented here indicate that the lattice mismatch between the film and the substrate is accommodated mainly by interface misfit dislocations above some critical thickness. The electronic structure of the interface between the film and the substrate is also explored using electron energy-loss spectroscopy.

(001)↗
DOE OSTI2012

Irradiation effects on microstructure change in nanocrystalline ceria – Phase, lattice stress, grain size and boundaries

With a wide variety of applications in numerous industries, ranging from biomedical to nuclear, ceramics such as ceria are key engineering materials. It is possible to significantly alter the materials functionality and therefore its applications by reducing the grain size to the nanometer size regime, at which point the unique varieties of grain boundaries and associated interfaces begin to dominate the material properties. Nanocrystalline films of cubic ceria deposited onto Si substrates have been irradiated with 3 MeV Au+ ions at temperatures of 300 and 400 K to evaluate their response to irradiation. It was observed that the films remained phase stable. Following a slight stress relief stage at low damage levels, the overall lattice is extremely stable up to high irradiation dose of ~34 displacements per atom. The grains were also observed to undergo a temperature-dependent grain growth process upon ion irradiation. This is attributed to a defect-driven mechanism in which the diffusion of defects from the collision cascade is critical. Formation of dislocations that terminate and stabilize at symmetric grain boundaries may be the limiting factor in the grain growth and overall energy reduction of the system. Utilizing ion modification, possible improvement of the adhesion of thin films and reduction of the probability of detrimental effects of stress-induced problems are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Microstructure and residual elastic strain at graphite nodules in ductile cast iron analyzed by synchrotron X-ray microdiffraction

The microstructure and residual elastic strain at graphite nodules (GNs) in ductile cast iron produced using either a fast or slow cooling rate have been characterized using synchrotron 3D X-ray Laue microdiffraction. The results demonstrate that thermal stress is introduced during cooling and that part of this stress is relaxed by plastic deformation of the polycrystalline ferrite matrix. It is observed that the plastic deformation is accommodated by the formation of dislocations and dislocation boundaries, which are organized in a cell structure. The dislocation density quantified based on the microstructure is most pronounced at the GN/matrix interface around small GNs in the fast cooled sample. Residual elastic strain is also present, which is mainly compressive with a maximum of 6.0–9.9 × 10 -4 near the GNs. Gradients of plastic deformation and elastic strain field around the GNs are observed. The results document for the first time that both the elastic strain field and the plastic strain field averaged over the grains around the GNs is approximately scaling with GN size and not affected by the cooling rate. The experimental data are compared with simulations by a finite element method, and agreement and disagreement are considered in detail.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Nanoscale conditions for ductile void nucleation in copper: vacancy condensation and the growth-limited microstructural state

Ductile rupture or tearing usually involves structural degradation from the nucleation and growth of voids and their coalescence into cracks. Although some materials contain preexisting pores, the first step in failure is often the formation of voids. Because this step can govern both the failure strain and the fracture mechanism, it is critical to understand the mechanisms of void nucleation and the enabling microstructural configurations which give rise to nucleation. To understand the role of dislocations during void nucleation, the present study presents ex-situ cross-sectional observations of interrupted deformation experiments revealing incipient, subsurface voids in a copper material containing copper oxide inclusions. The local microstructural state was evaluated using electron backscatter diffraction (EBSD), electron channeling contrast (ECC), transmission electron microscopy (TEM), and transmission kikuchi diffraction (TKD). Surprisingly, before substantial growth and coalescence had occurred, the deformation process had resulted in the nucleation of a high density of nanoscale (≈50 μm) voids in the deeply deformed neck region where strains were on the order of 1.5. Such a proliferation of nucleation sites immediately suggests that the rupture process is limited by void growth, not nucleation. With regard to void growth, analysis of more than 20 microscale voids suggests that dislocation boundaries facilitate the growth process. The present observations call into question prior assumptions on the role of dislocation pile-ups and provide new context for the formulation of revised ductile rupture models. While the focus of this study is on damage accumulation in a highly ductile metal containing small, well-dispersed particles, these results are also applicable to understanding void nucleation in engineering alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Highly deformable Mg–Al–Ca alloy with Al 2 Ca precipitates

Magnesium (Mg) is the lightest structural metal. However, the poor formability of Mg alloys to great extent limits their applications in making structural parts. Formability is strongly correlated to both high tensile elongation and large work hardening capacity. In this work, we report a new Mg-Al-Ca alloy in which a majority of deformable Al 2 Ca precipitates form while the formation of Laves phases of Mg 17 Al 12 and Mg 2 Ca seems suppressed. Al 2 Ca precipitates imped dislocation motion, leading to large work hardening. Then, Al 2 Ca precipitates deform with dislocations and stacking faults under the enhanced flow stress, which relieve local stress concentration and improve tensile elongation. In addition, solutes Al and Ca suppress twin nucleation while promoting non-basal dislocations in Mg. Our Mg-Al-Ca alloy demonstrates one of the highest combinations of tensile elongation and work hardening capacity among existing Mg alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Improved irradiation resistance of accident-tolerant high-strength FeCrAl alloys with heterogeneous structures

In this work, post–neutron irradiation examination is performed on advanced accident-tolerant fuel (ATF) cladding iron-chromium-aluminum (FeCrAl) alloys with ~10–13at. % Cr, ~10–12 at. % Al, ~1 at. % Mo, and minor alloying elements including Y irradiated to a damage level of 7 displacements per atom (dpa) at irradiation temperatures of 267–282 °C. A compositional dependency of the Cr and Al content is observed on the ratio of sessile and glissile dislocation loops, where the density of a$\langle$100$\rangle$ type loops is somewhat higher than the a/2$\langle$111$\rangle$ type loops. The α' precipitate number density is inversely correlated to the starting Cr concentration of the alloys of interest. The irradiation to a higher dose of 7 dpa results in a higher density of dislocation loops and α' precipitates for the same alloys at a lower irradiation dose, such as 1.8 dpa. In this work, the effect of α' precipitates on the dislocation loop density is discussed, and the presence of α' appears to inhibit the nucleation of loops. Compared with first-generation FeCrAl alloys, these advanced alloys with heterogeneous structure exhibit a lower Cr concentration in α' precipitation at the same dose level; they act as weaker obstacles deviating from the primary hardening contribution from the mature α'. Hence, the overall irradiation-induced hardening decreases; our alloys show improved radiation resistance because of their stronger sink strengths. The results presented in this paper could provide insights for the design and optimization of ATF cladding materials for future fission and space applications.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Plastic homogeneity in nanoscale heterostructured binary and multicomponent metallic eutectics: An overview

Heterostructured materials comprised of relatively soft/hard disparate phases typically exhibit composite strengthening but lack plastic deformability at ambient temperatures. However, heterostructured systems comprised of nanoscale phases can simultaneously enhance yield strength and strain hardening, thereby promoting uniform distribution of plastic flow. In this review, the atomic-scale deformation mechanisms in model systems of eutectic alloys, Al-Al 2 Cu and Al-Si, refined to nanoscales via laser rapid solidification are discussed, and compared with literature on multi-component (high entropy) eutectics such as Ni-Al-Fe-based with Cr and/or Co additions. The nano-lamellar Al-Al 2 Cu structures exhibit unit defect mechanisms not reported in monolithic Al 2 Cu intermetallic: localized shear on {0 1 1} and shear-induced faults on {1 2 1} planes, constrained by closely-spaced dislocation arrays in Al confined by Al/Al 2 Cu interfaces. The unexpected plasticity mechanisms are enabled by slip continuity in nanoscale Al-Al 2 Cu eutectics associated with the orientation relationship and interface habit planes. In nano-fibrous Al-Si eutectic, tensile ductility at strength approaching 600 MPa is observed resulting from dislocation plasticity in the nano-Al channels and cracking in Si nanofibers. Molecular dynamics simulations show that Al dislocations easily cross-slip (screw) or climb (edge) along Al-Si interfaces, making slip transmission difficult. The propagation of nano-cracks is suppressed by surrounding strain hardening Al, retaining good ductility of the sample, in spite of lack of direct slip transmission. Finally, the critical unit mechanisms of slip transmission and interface-enabled plasticity observed in nanoscale eutectic binary systems can also explain the strength-ductility relationship in multi-component eutectics and homogeneously distributed plastic flow with increasing microstructural heterogeneity.

36 MATERIALS SCIENCE↗
DOE OSTI2026

In-situ observations of cyclic deformation in an extruded Mg-2Nd-1Y-0.1Zr-0.1Ca alloy

In this study, the evolution of deformation mechanisms during cyclic loading in an extruded, solution-treated Mg–2Nd–1Y–0.1Zr–0.1Ca alloy was investigated using a combination of in-situ loading, scanning electron microscopy (SEM), electron backscatter diffraction (EBSD), and focused ion beam (FIB) nanofabrication. The initial microstructure exhibited a random crystallographic texture with no preferred grain orientation. Flat, rectangular dog-bone specimens were subjected to load-controlled, fully reversed fatigue for 50 cycles, during which the same region was sequentially mapped to track microstructural changes. After 10 cycles of loading deformation twins were observed. During tensile reloading detwinning or narrowing of those twinned regions occurred. After 20 cycles, detwinning ceased and residual twins remained in the material. SEM imaging revealed numerous surface slip traces after cyclic loading. EBSD-assisted slip trace analysis identified the activation of prismatic and pyramidal < c+a> slip systems during low-cycle fatigue. Site-specific scanning transmission electron microscopy (STEM) further revealed that deformation was also accommodated by basal < a> slip and the dissociation of < c+a> dislocations. Center-of-symmetry (COS) analysis confirmed that the dissociation of < c+a> dislocations resulted in the formation of I₁ intrinsic stacking faults after cyclic loading. These findings provide new insights into the complex interplay of dislocation mechanisms governing fatigue deformation in rare-earth-containing Mg alloys.

Cyclic deformation↗
DOE OSTI2020

The effect of local chemical ordering on Frank-Read source activation in a refractory multi-principal element alloy

In this work, we investigate the operation of Frank-Read (FR) sources in a refractory multi-principal element alloy (MPEA). Simulations of discrete dislocation motion in MPEAs is enabled by the development of a phase field dislocation dynamics model that treats the atomic-scale fluctuations in lattice energies across the glide plane, present due to local ordering in the chemical composition within the nominally random MPEA atomic structure. We consider, through simulation, a range of length scales over which ordering occurs, varying from short-range lengths, a few times dislocation core width, to long-range lengths, an order of magnitude longer than the core. Characteristic of this body-centered cubic MPEA, the simulations also include screw/edge character dependence in glide resistance, as informed by atomic scale simulation. The critical stresses to activate the source for the same source size are found to be statistically distributed, as a direct consequence of the underlying variation in lattice energy. Furthermore, analysis of critical state for activating edge and screw FR sources in the MPEA reveals that FR source operation occurs via a two-step mechanism, involving athermal kink-pair formation, unlike the conventional FR source operation in a material with no composition fluctuations. This mechanism lowers the average critical stress required to activate the FR source and causes the statistical dispersion in critical stress to depend on the range of composition ordering. More importantly, it leads to a more severe dependence of source strength on FR source length than predicted by line tension alone.

36 MATERIALS SCIENCE↗
DOE OSTI2015

X-ray topographic study of growth defects of trans-stilbene crystals grown from solutions

Single crystals of trans-stilbene, C 14 H 12 , with properties suitable for high-energy neutron detection were grown from solution in anisole and toluene by the temperature reduction method with growth rates up to 6 mm/day. From these crystals, slices of appropriate orientation and thickness of 2–4 mm were cut and studied by X-ray diffraction topography applying the Lang method using CuKα radiation. The topographs exhibit growth defects such as liquid inclusions, dislocations, striations, and faulty growth-sector boundaries. These defects occur in the same typical arrangements and geometries as is observed in all kinds of crystals grown on habit faces from solution. Besides growth dislocations originating from inclusions and propagating with the growth front, many plastic glide dislocations in the shape of loops or half-loops emitted from inclusions by stress relaxation are observed. Finally, the glide system underlying this plasticity is discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2022

A concurrent irradiation-mechanics multiscale coupling model

The vast majority of our current knowledge regarding the basic mechanisms controlling irradiation effect on mechanical properties is based almost entirely on the results of post-irradiation experiments or theoretical models. However, the concurrent effects of irradiation, mechanical stress, and thermal damage on the failure phenomena of materials and components remain largely unexplored due to its internal multiscale-multiphysics coupling nature. Here, we present here a concurrent irradiation-mechanics multiscale coupling model. The concurrent evolutions of nanoscale irradiation defect clusters, microscale dislocation configurations, and mechanical responses are well captured based on coupling cluster dynamics, discrete dislocation dynamics, and the finite element methods using an effective time marching scheme. Model predictions of defect densities and size are in general agreement with experimental observations. Irradiation hardening is shown to take place also in samples undergoing concurrent irradiation-mechanical loading, similar to samples tested post-irradiation. However, the occurrence of plastic flow localization and dislocation channel formation is not accompanied with apparent yield drop (softening) under concurrent irradiation-mechanical loading conditions, which is different from the post-irradiation case.

42 ENGINEERING↗
DOE OSTI2011

In situ transmission electron microscopy of electron-beam induced damage process in nuclear grade graphite

Atomic level processes involved in the swelling and crack-closing in nuclear grade graphite under electron irradiation have been observed in real-time using transmission electron microscopy. Noise-filtered lattice images show the formation of vacancy loops, interstitial loops and resulting dislocations with unprecedented clarity. The dislocation dipoles formed via vacancy loops were found to undergo climb resulting in extra basal planes. Concurrent EELS studies showed a reduction in the atomic density because of the breakage of hexagonal carbon rings. The formation of new basal planes via dislocation climb in addition to the bending/breaking of basal planes leads to swelling and closing of micro-cracks.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Effect of heavy ion irradiation on microstructural evolution in CF8 cast austenitic stainless steel

The microstructural evolution in ferrite and austenitic in cast austenitic stainless steel (CASS) CF8, as received or thermally aged at 400 °C for 10,000 h, was followed under TEM with in situ irradiation of 1 MeV Kr ions at 300 and 350 °C to a fluence of 1.9 × 10 15 ions/cm 2 (~3 dpa) at the IVEM-Tandem Facility. For the unaged CF8, the irradiation-induced dislocation loops appeared at a much lower dose in the austenite than in the ferrite. At the end dose, the austenite formed a well-developed dislocation network microstructure, while the ferrite exhibited an extended dislocation structure as line segments. Compared to the unaged CF8, the aged specimen appeared to have lower rate of damage accumulation. The rate of microstructural evolution under irradiation in the ferrite was significantly lower in the aged specimen than in the unaged. Finally, we attributed this difference to the different initial microstructures in the unaged and aged specimens, which implies that thermal aging and irradiation are not independent but interconnected damage processes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Effects of neutron irradiation of Ti 3 SiC 2 and Ti 3 AlC 2 in the 121–1085 °C temperature range

Herein we report on the formation of defects in response to neutron irradiation of polycrystalline Ti 3 SiC 2 and Ti 3 AlC 2 samples exposed to doses of 0.14±0.01, 1.6±0.1, and 3.4±0.1 displacements per atom (dpa) at irradiation temperatures of 121±12, 735±6 and 1085±68 °C. After irradiation to 0.14 dpa at 121 °C and 735 °C, black spots are observed in both Ti 3 SiC 2 and Ti 3 AlC 2 . After irradiation to 1.6 and 3.4 dpa at 735 °C, basal dislocation loops, with a Burgers vector of b = ½ [0001] are observed in Ti 3 SiC 2 , with loop diameters of 21±6 and 30±8 nm for 1.6 dpa and 3.4 dpa, respectively. In Ti3AlC2, larger dislocation loops, 75±34 nm in diameter are observed after 3.4 dpa at 735 °C, in addition to stacking faults. Impurity particles of TiC, as well as stacking fault TiC platelets in the MAX phases, are seen to form extensive dislocation loops under all conditions. Voids are observed at grain boundaries and within stacking faults after 3.4 dpa irradiation, with extensive void formation in the TiC regions at 1085 °C. Remarkably, denuded zones on the order of 1 µm are observed in Ti 3 SiC 2 after irradiation to 3.4 dpa at 735 °C. Small grains, 3-5 µm in diameter, are damage free after irradiation at 1085 °C at this dose. The presence of the A-layer in the MAX phases is seen to provide enhanced irradiation tolerance. Based on these results, and up to 3.41 dpa, Ti 3 SiC 2 remains a promising candidate for high temperature nuclear applications.

36 MATERIALS SCIENCE↗
DOE OSTI2022

The influence of nitrogen and nitrides on the structure and properties of proton irradiated ferritic/martensitic steel

The 12Cr1MoWV (wt%) ferritic/martensitic steel HT9 is a candidate material for fuel cladding in advanced nuclear reactors, such as the Versatile Test Reactor currently under development. As such, understanding the relationship between microstructure and mechanical properties in the context of irradiation environments for these steels is critical. N content, and more specifically interstitial N, has been hypothesized to be detrimental to irradiated properties at lower temperatures (less than 0.3T m ) to a total of 6 dpa; however, in this work at a dose of 1 dpa the irradiated microstructure was improved with added N, leading to less irradiation hardening. Three variants of HT9 were irradiated with 1.5 MeV protons to a dose of 1 dpa at 300°C. The HT9 variants included Low (10 ppm), Mid (190 ppm), and High (440 ppm) N alloys that were otherwise nearly identical. Changing the N content had a variety of effects on the irradiated defect structures. As N content increased, the average dislocation loop diameter decreased, while the number density of loops increased. Additionally, extensive Ni clustering was observed on dislocations and interfaces. The Mid and High N specimens exhibited significantly less hardening (ΔHV≃100) relative to the Low N specimen (ΔHV≃160). The decrease in hardening is attributed to vanadium carbonitride acting as a sink for Ni clusters that would otherwise form on dislocations. Under the irradiation conditions used, these results suggest increasing the N content in HT9 may have a desirable effect on the irradiated structure and properties at the dose studied, as well as the swelling resistance at higher doses. In other words, N content appears to be a powerful tool for tailoring the self-interstitial atom cluster mobility in F/M steels for different temperature and dose applications.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Mechanistic nuclear fuel performance modeling of uranium nitride

Uranium mononitride (UN) is a nuclear fuel candidate for advanced reactor designs and an alternative being considered for light water reactors due to its higher thermal conductivity and uranium density than UO 2 . As with any nuclear fuel, swelling and fission gas release are important factors for safety, while also being some of the hardest phenomena to predict with a high degree of confidence. Getting a grasp on the gas swelling behavior and release is crucial to lower the barrier for UN utilization. An accelerated swelling rate at high temperatures observed experimentally, sometimes referred to as “breakaway swelling,” further complicates the prediction of fuel performance of UN. A mechanistic model has been developed using a multiscale approach to describe the intragranular and intergranular fission gas behavior. Lower-length-scale calculations have been employed to inform models of the gas and self-diffusion behavior, resolution rate, and bubble shape. Leveraging previous work on high burnup UO 2 , two populations of intragranular bubbles are considered; small bulk bubbles and larger bubbles located along dislocations. The dislocation bubbles were found to be crucial to the overall swelling behavior, and the breakaway swelling transition was associated with the transition in the gas atom diffusion mechanism from an irradiation-induced athermal diffusion regime at lower temperatures to an intrinsic thermal equilibrium regime at higher temperatures, accelerating the growth of the dislocation bubbles. Similarly, the threshold for fission gas release was associated with the grain boundary vacancy diffusivity surpassing the gas atom diffusivity at sufficiently high temperatures, allowing the over-pressurized grain boundary bubble to grow in size and interconnect. Using thermo-mechanical models with the fission gas model, two integral fuel pin assessment cases were simulated. Finally, this work demonstrates the ability of a multiscale approach to accelerate the understanding of advanced fuel forms when experimental data is limited.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗