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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 OSTI2025

Effect of reversible dislocation-based deformation on nanoparticle strain at failure

Molecular dynamics simulations are used to isolate and quantify the reversible and irreversible mechanisms that contribute to deformation in platinum nanoparticles under compression. Quantitative analysis reveals how the nucleation and entanglement of dislocations can lead to reversible dislocation-based deformation. Simulations run at different temperatures and loading conditions show that the formation of entangled dislocations is more likely at higher temperatures and is facilitated by loading orientations where dislocations nucleate on intersecting slip planes. The presence of entangled dislocations increases the strain at failure due to the ability of those dislocations to accommodate strain reversibly. The results are corroborated by the observation of similar entangled dislocation loops during in situ compression experiments on nanoparticles of the same material. Altogether, these findings provide insight into the role of dislocations in both reversible and irreversible deformation and their implications for nanoparticle stability and properties.

36 MATERIALS SCIENCE↗
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 OSTI2022

Discrete dislocation dynamics for crystal RVEs. Part 1: Periodic network kinematics

A novel implementation of the dislocation flux boundary condition in discrete dislocation dynamics is presented. The continuity of the individual dislocation loops in a periodic representative crystal volume (RVE) is enforced across the boundary of the RVE with the help of a dual topological description for representing dislocation line kinematics in two equivalent spaces representing the deforming crystal, the RVE and the unbounded crystal spaces. The former describes the motion of the dislocations in the simulated crystal RVE whereas the latter represents the motion of dislocations in an infinite space containing all replicas of the RVE. A mapping between the two spaces forms the basis of the implementation of flux boundary condition. The implementation details are discussed in the context of statistical homogeneity of bulk crystals undergoing macroscopically homogeneous plastic deformation. In this case, the boundary nodes associated with dislocation segments bear no relevance in the motion of the dislocations. Finally, some test cases are presented and discussed to establish the proposed approach.

42 ENGINEERING↗
DOE OSTI2021

Dissociated prismatic loop punching by bubble growth in FCC metals

Abstract Materials performance can be significantly degraded due to bubble generation. In this work, the bubble growth process is elaborated in Cu by atomistic modeling to bridge the gap of experimental observations. Upon continuous He implantation, bubble growth is accommodated first by nucleation of dislocation network from bubble surface, then formation of dissociated prismatic dislocation loop (DPDL), and final DPDL emission in $$\langle 110\rangle$$ ⟨ 110 ⟩ directions. As the DPDL is found capable of collecting He atoms, this process is likely to assist the formation of self-organized bubble superlattice, which has been reported from experiments. Moreover, the pressurized bubble in solid state manifests the shape of an imperfect octahedron, built by Cu $$\{111\}$$ { 111 } surfaces, consistent with experiments. These atomistic details integrating experimental work fill the gap of mechanistic understanding of athermal bubble growth in Cu. Importantly, by associating with nanoindentation testings, DPDL punching by bubble growth arguably applies to various FCC (face-centered cubic) metals such as Au, Ag, Ni, and Al.

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 OSTI2022

Orientation-selected micro-pillar compression of additively manufactured 316L stainless steels: Comparison of as-manufactured, annealed, and proton-irradiated variants

In this work, irradiation response and deformation mechanisms of additively manufactured (AM) 316L stainless steel were studied by atomic scale characterization and micro-pillar compression. The AM 316L stainless steels were fabricated by direct energy deposition, a laser-based additive manufacturing process. Irradiation with 2 MeV protons at 360°C was performed to create ~1.8 displacements-per-atom (dpa) damage in AM 316L. Deformation behaviors of the as-manufactured, annealed, and proton-irradiated variants were studied, focusing on the effects of manufacturing-induced pores, residual stress, and irradiation-introduced defects (dislocation loops and voids). Micro-pillars were prepared from grains of pre-selected orientation, avoiding contributions of grain boundaries and allowing determination of resolved shear stress on {111} glide planes. Transmission electron microscopy was used to characterize the pre- and post-deformation microstructure. It was found that in the as-manufactured alloy variant, moving dislocations were the major deformation carrier, with noticeable blocking by fabrication-induced pores, In the annealed variant, hardness was reduced, and deformation was also accomplished by dislocation gliding. In the proton-irradiated variant, significant twinning was observed. Comparing measured resolved shear stress and predicted critical stress for dislocation dissociation, we conclude that irradiation hardening became high enough to activate twinning. Therefore, the deformation mechanism changes from dislocation gliding to twinning. The study is important for both processing optimization and performance evaluation of AM alloys for reactor applications.

36 MATERIALS SCIENCE↗
DOE OSTI2020

The combined effects of hydrogen and aging condition on the deformation and fracture behavior of a precipitation-hardened nickel-base superalloy

The effect of hydrogen (H) on the deformation behavior of Monel K-500 in various isothermal heat treatment conditions (non-aged, under-aged, peak-aged, and over-aged) was assessed via uniaxial mechanical testing. H-charged and non-charged specimens were strained to failure to facilitate a comparison of ductility, fracture surface morphology, strength, and work hardening behavior. For all examined heat treatment conditions, H charging leads to a significant reduction in ductility, which is accompanied by a consistent change in fracture surface morphology from ductile microvoid coalescence to brittle intergranular fracture. While H charging led to a systematic enhancement in the yield strength of all heat treatments, the three age-hardened conditions exhibited a more than 2-fold increase relative to the non-aged heat treatment. This suggests that H modifies the dislocation–precipitate interactions, which also manifest themselves through changes in work hardening metrics related to the dislocation storage and recovery rates. In particular, the H-charged peak-aged specimen exhibited a significant increase in initial hardening (dislocation storage) rate relative to the H-charged under-aged specimen. Transmission electron microscopy of these samples revealed the onset of widespread dislocation looping in the H-charged peak-aged sample, in addition to the planar slip bands characteristic of the non-charged condition. This result suggests that hydrogen induces the particle shearing-to-looping transition at smaller particle sizes. Possible mechanistic explanations for this observed behavior are presented.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Using lifetime of point defects for dislocation bias in bcc Fe

The interaction between dislocations and point defects is key to deformation processes and microstructural evolution of structural materials. In this work, we compute the lifetime of point defects to describe their interaction with dislocations. This approach can accurately account for the effects of the dislocation core and anisotropic defect dynamics to accumulatively determine the capture efficiency, sink strength, and dislocation bias at different temperatures and dislocation densities. Particularly, the absorption of point defects by straight screw and edge dislocations in a model bcc iron system is studied. The maximum swelling rates based on the obtained bias factors are in close agreement with a variety of experimental measurements, including both neutron and ion-irradiation data, especially when considering the survival fraction for point defects from displacement cascades. Further, this approach applies to many other processes and sinks, such as dislocation loops and interfaces, providing a powerful means to develop fundamental insights critical for improving radiation resistance and mechanical properties of structural materials through controlling defect interaction and evolution.

36 MATERIALS SCIENCE↗
DOE OSTI2023

In-situ and ex-situ characterization of ion-irradiated AM materials

Additive manufacturing (AM) has attracted increasing attention in recent years as a new way of making high-quality components for nuclear reactors. While AM materials are compositionally similar to their conventionally produced counterparts, they do possess different microstructures, such as dislocation cells and chemical inhomogeneity, that can lead to different mechanical properties and performance behavior. In this study, the irradiation response of AM materials was investigated. In-situ and ex-situ ion irradiations were performed on AM316L and AM316H stainless steels (SS) at 300 and 600°C. The influence of the dislocation cell structure on the evolution of irradiation-induced dislocation loops was evident at 600ºC, but was much weaker at 300ºC. No voids were observed with the in-situ ion irradiation up to 10 dpa at both temperatures. Post-irradiation energy dispersive spectroscopy showed radiation-induced segregation (RIS) near grain boundaries and the formation of Cr-rich oxides throughout the matrix. The extent of segregation at dislocation cell walls varies with dose. Nanoindentation tests performed on the AM316L SS irradiated at 600ºC showed a complex dose dependence with softening at low doses and hardening at high doses.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Microchemical evolution of irradiated additive-manufactured HT9

The microstructural responses under 5 MeV Fe2 + single-ion-beam irradiation of three conditions of additive-manufactured (AM) HT9 steel using a powder-based directed energy deposition (DED) technique with and without postbuild heat treatments were investigated. Besides the observed dislocation loop formation and the absence of cavities at the irradiation condition of 50 dpa at 460 °C, Ni/Si/Mn-rich precipitates are found to form in all three conditions of AM-HT9, whereas Cu-rich clusters that arise from Cu uptake from the DED process are only observed in the heat-treated conditions, and not in the as-built (ASB) condition. Coprecipitation of the Cu- and Ni/Si/Mn-rich clusters occur near defect sinks such as line dislocations and grain boundaries in the heat-treated AM-HT9. The variation in microchemical evolution can be directly linked to the starting sink strength of the three AM-HT9 conditions, and the ASB condition with higher sink strength suppressed the responses observed in the postbuild heat-treated specimens.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Prediction of dislocation - grain boundary interactions in FCC aluminum bicrystals using a modified continuum criterion and machine learning methods

Mechanical properties of metals such as strength and toughness are strongly correlated to complex interactions between various defects in the crystalline structure. While elementary interactions between these defects have been investigated using recent micro- and nano-characterization techniques, understanding of the detailed interaction mechanisms has hardly been obtained. To understand defect-driven plasticity at various time and length scales, it is necessary to formulate a general guideline to predict both the interaction type (transmission or reflection) and the dislocation's subsequent slip system after the interaction. Many criteria based on the geometric alignment of the defects have been developed to predict this phenomenon, but these have yet to be found to be accurate when applied to general data sets of grain boundaries (GBs). With this motivation, we conduct a systematic study using molecular dynamics (MD) models of bicrystals to analyze defect interaction process between a prismatic dislocation loop and eleven different grain boundaries of the following character: three tilt, three twist, and five mixed. Based on the MD observations, two new prediction methods are developed: the first is a new data-driven parametric score function based on the classical geometric criteria, and the second is by applying Gaussian process machine learning methods to find the probability distribution of a hidden function. In conclusion, the proposed data-driven prediction methods could pave a new way to predict the unit interaction of dislocations with various GBs, which could show much higher accuracy compared to pre-existing geometric criteria.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Molecular dynamics studies of lattice defect effects on tritium diffusion in zirconium

Tritium diffusion in α-Zr containing point defects such as vacancies or self-interstitial atoms (SIAs) is simulated using molecular dynamics. Point defects rapidly aggregate to form extended defects, such as 3D nanoclusters and Frank loops. The geometry of extended defects is affected by the presence of tritium. At low temperature and in the absence of tritium, vacancies aggregate to form stacking fault pyramids. Addition of tritium at these temperatures promotes aggregation of vacancies to form 3D nanoclusters, within which the tritium concentration can be sufficiently high to suggest that these defects may serve as nucleation sites for hydride precipitation. Trapping of tritium in vacancy nanocluster reduces the calculated bulk diffusivity by an amount proportional to the vacancy concentration. At high temperature, vacancy clusters change shape to form planar basal dislocation loops, which bind tritium less strongly, leading to a sharp reduction in the fraction of trapped tritium and a corresponding increase in tritium diffusivity at high temperature. In contrast, SIAs increase tritium diffusion through α-Zr. Analysis of atomic trajectories shows that tritium does not interact directly with SIAs. In conclusion, diffusion enhancement is instead related to expansion of the lattice.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Post Irradiation Examination of Pressurized Water Reactor Stainless Steel Internal Components

Internal structural components of pressurized water reactors (PWR), such as baffle former bolts, are subjected to significant neutron irradiation and mechanical stresses at elevated temperatures during plant operation. Over the long operation of the power plant, these conditions lead to potential degradation and reduced load-carrying capacity of these bolts. To understand property degradation more fully, to confirm the results of experimental irradiation programs, and to predict operational lifetime performance of structural materials in internal components, post irradiation examination of harvested materials from operating nuclear reactors is required. In this work, two high fluence 316 stainless steel baffle former bolts were retrieved from a commercial Westinghouse two-loop downflow type PWR and then sectioned for analysis via mechanical testing and microstructural characterization. The irradiation damage fluctuated along the bolt length with damage levels from 15 to 41 displacements per atom with the bolt head receiving approximately twice the neutron damage levels as that of the bolt thread section. Mechanical testing evaluations showed extensive irradiation hardening and a sharp decrease in fracture toughness in all parts of the bolt, though with limited variation along the bolt length. Microstructural characterization using analytical scanning transmission electron microscopy and atom probe tomography showed significant radiation induced precipitation, segregation, and dislocation loop and cavity formation. However, unlike the mechanical behavior, there was considerable variation along the bolt length but opposite of what may be expected, with more precipitation and cavity formation in the bolt thread where the neutron radiation dose is less than in the bolt head where the neutron radiation dose is higher. The cause of this variation is likely due to gradients in temperature, neutron energy spectra, and gamma irradiation.

Lach, Tim↗
DOE OSTI2021

Molecular dynamics simulation of vacancy and void effects on strain-induced martensitic transformations in Fe-50 at.% Ni model concentrated solid solution alloy

Strain-induced martensitic transformations can improve the strength and ductility of some face centered cubic (FCC) metals and alloys. Irradiation-induced defects such as vacancies, dislocation loops, and voids activate martensitic transformations over a wider range of mechanical loading conditions than in pristine material. Furthermore, the mechanisms underlying irradiation-enabled martensite transformations remain unclear. In this work, we use molecular dynamics simulations to study the effect of vacancies and voids on strain-induced martensitic transformations in a model concentrated solid solution alloy Fe-50 at.% Ni. It has been found that single vacancies have no resolvable effect on the transformation because they reduce the stacking fault energy by a relatively insignificant margin and do so only if the vacancy is located on the stacking fault plane. Voids, however, activate the martensite transformation through shear strain accumulation around the void due to dislocation pileup. The larger the void, the more pronounced this effect.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Deconvoluting the Effect of Chromium and Aluminum on the Radiation Response of Wrought FeCrAl Alloys After Low-Dose Neutron Irradiation

FeCrAl alloys have been extensively investigated over the past decade as a candidate material for accident-tolerant fuel cladding in light water reactors. Here, we have completed a first of its kind study where Al and Cr concentrations are varied systematically under neutron irradiation to elucidate the post irradiation microstructural and mechanical response as a direct result of alloying content. Neutron irradiations were performed on alloys with composition Fe-(10-13)Cr-(5-7)Al in wt.% at temperatures of 214, 357, and 557°C to a dose of ~1.8 dpa (displacements per atom). Dislocation loop sizes and dispersion characteristics did not show strong compositional dependence. However, there were noticeable effects of composition on the dispersion of Cr-rich α'-precipitates, particularly a decrease in number density with increasing Al content or a decrease in Cr content. The present results confirm previous studies indicating the Cr concentration in these precipitates is lower than that expected in binary FeCr alloys and that Al can act as a destabilizing alloying element for the deleterious α'-phase.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Comparison of hardening and microstructures of ferritic/martensitic steels irradiated with fast neutrons and dual ions

Ferritic/martensitic steels T91 and HT9 were irradiated with neutrons (BOR-60 reactor) and dual ions (9 MeV Fe 3+ and 3.42 MeV energy degraded He 2+ ) from 369 to 520 °C and damage levels of 16.6 to 72 dpa to quantify the possibility of using ion irradiation to simulate neutron irradiation in terms of microstructures and mechanical properties. Nanoindentation testing was performed to obtain the bulk equivalent hardness of the dual-ion irradiated samples. For the neutron irradiated samples, both nanoindentation and Vickers hardness testing were conducted. Transmission Electron Microscopy (TEM) characterizations of the cavities, dislocation loops and precipitates were conducted to account for the strengthening contribution of each microstructure element. The good agreement between the microstructure-predicted (dispersed barrier hardening) and measured strength of the irradiated specimens demonstrated the accuracy of the strengthening model and the nanoindentation tests. Furthermore, the comparison of mechanical property and microstructure changes in ion and neutron irradiated structural materials indicated that ion irradiation replicated many neutron irradiation features. However, a single 70 °C temperature shift is insufficient to match all complex microstructures of neutron vs. ion irradiation over the irradiation temperature range of 369–520 °C.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Defect-specific strength factors and superposition model for predicting strengthening of ion irradiated Fe18Cr alloy

Here, a high-purity binary alloy Fe18Cr was subjected to heavy-ion irradiation in order to provide improved understanding of the irradiation effect on radiation hardening associated with dislocation loop and network formation, α’ precipitation. The specimens were irradiated with 8 MeV Fe ions (~2 μm ion range) to midrange doses of 0.37 and 3.7 dpa at 300, 350 and 450 °C using dose rates of ~10 –5 –10 –3 dpa/s. Nanoindentation testing was performed to extract the bulk equivalent hardness at low depth from these ion irradiated specimens. High-quality transmission electron microscopy (TEM) images were acquired by utilizing a flash electropolishing method on the Focused Ion Beam (FIB) prepared liftouts. An accurate calculation of the strength factor was provided based on detailed TEM characterization of the irradiated microstructures. A newly refined hardening superposition method was applied to combine the strengthening components from each microstructure. The good agreement between microstructure predicted strength and measured strength further demonstrated the fidelity to use hardening model to quantify the mechanical properties of ion irradiated materials.

36 MATERIALS SCIENCE↗
DOE OSTI2020

The correlation between microstructure and nanoindentation property of neutron-irradiated austenitic alloy D9

The microstructure and nanomechanical properties of three samples of the modified stainless steel (referred as the alloy D9) were systematically characterized after neutron irradiation in the Advanced Test Reactor. The samples were irradiated to 5.0, 8.2, and 9.2 displacements per atom at 448, 430, and 683°C, respectively. The evolutions of dislocation loops, cavities, and radiation-induced precipitates were quantitatively studied to reveal their dose and temperature dependencies. Nanohardness and nanoindentation creep tests were conducted at room temperature on the irradiated samples. Unexpected radiation hardening was observed in the highest-temperature-irradiated sample due to the formation of an unknown type of Ni- and Si-rich precipitates whose contributions to the radiation and mechanical performances of the alloy were discussed. We provide the radiation-microstructure-property correlations of alloy D9 with new insights, which can benefit the development and optimization of advanced austenitic alloys for future nuclear applications.

36 MATERIALS SCIENCE↗