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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 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Plasticity in irradiated FeCrAl nanopillars investigated using discrete dislocation dynamics

Here, in this paper, we investigate plasticity in irradiated FeCrAl nanopillars using discrete dislocation dynamics simulations (DDD), with comparisons to transmission electron microscopic (TEM) in situ tensile tests of ion- and neutron-irradiated commercial C35M FeCrAl alloy. The effects of irradiation-induced defects, such as a/2<111> and a<100> type loops and composition fluctuations representative of phase separation in irradiated FeCrAl alloys, are investigated separately as well as superposed together in simulations. We explore the effects of defects on the stress-strain behavior, specifically yield strength and hardening response, of FeCrAl nanopillars. Our simulations confirm the widely accepted fact that irradiated alloys exhibit a stress-strain response with higher yield strength as compared to unirradiated alloys. However, our DDD calculations reveal an atypical superposition of the hardening contributions due to composition inhomogeneity and irradiation loops wherein composition inhomogeneity annihilates the hardening due to irradiation loops at small scales. As a result, we observe that the yield strength in irradiated alloys, after taking into consideration the effects of both composition inhomogeneity and irradiation loops, is smaller than the yield strength of the alloys with only irradiation loops and is approximately same for the alloy with composition inhomogeneity alone. This is referred to as “destructive interference” between the hardening contributions due to composition fluctuations and irradiation loops in the paper. We also identify this destructive interference in the superposition in our parallel TEM in situ tensile tests on unirradiated, ion-irradiated, and neutron-irradiated C35M FeCrAl alloy. This destructive interference in the hardening contributions contrasts with the dispersed barrier hardening (DBH) models widely utilized by the experimental community to model the hardening contributions due to different irradiation induced defects. The effects of the loading orientations on the yield strength and hardening are investigated and the mechanisms for the hardening in irradiated FeCrAl alloys are also reported.

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 OSTI2024

A Brief Review of the Impact of Neutron Irradiation Damage in Tungsten and Its Alloys

Neutron irradiation poses a substantial challenge in the development and application of tungsten (W) and its alloys, predominantly in the framework of nuclear fusion and fission environments. Although W is well-acknowledged for its unique properties like its high melting temperature and higher resistance to sputtering, transmutation products, such as Re and Os, form and impact the alloy properties as a result of neutron irradiation. This transmutation effect accompanied by significant microstructure damage due to neutron irradiation can lead to the significant degradation of mechanical properties. This review surveys the literature focusing on the microstructural modifications post-irradiation and its impacts on the irradiation hardening. This review provides insights into the elaborative understanding on the neutron radiation damage on W and W alloys by exploring the microstructural evolution and hardness changes post-irradiation. The gaps and future opportunities for understanding neutron radiation damage in W are briefly summarized.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Influence of grain size on α' Cr precipitation in an isothermally aged Fe-21Cr-5Al alloy

Cr-rich α' precipitation during aging typically leads to hardening and accordingly embrittlement of FeCrAl alloys, which needs to be suppressed. The influence of grain size on α' precipitation was studied by aging coarse-grained (CG), ultra-fine grained (UFG), and nanocrystalline (NC) ferritic Kanthal-D [KD; Fe-21Cr-5Al (wt.%) alloy] at 450, 500 and 550 °C for 500 h. After aging at 450 and 500 °C, less hardening was observed in the UFG KD than in CG KD. Atom probe tomography indicated a lower number density and larger sized intragranular α' in the UFG versus the CG alloy. The smaller grain size and higher defect (vacancy and dislocation) density in the UFG KD facilitated diffusion and accordingly enhanced precipitation kinetics, leading to coarsening of precipitates, as well as saturation of precipitation at lower temperatures, as compared to those in CG KD. No hardening occurred in UFG and CG KD after aging at 550 °C, indicating that the miscibility gap is between 500 and 550 °C. NC KD exhibited softening after aging owing to grain growth. α' precipitation occurred in NC KD aged at 450 °C but not at 500 °C, indicating that miscibility gap is between 450 and 500 °C. Thus, the significantly smaller grain size in NC KD decreased the miscibility gap, as compared to that in CG and UFG KD. Finally, this is attributed to the absorption of vacancies by migrating grain boundaries during aging, suppressing α' nucleation and enhancing Cr solubility.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Synergic grain boundary segregation and precipitation in W- and W-Mo-containing high-entropy borides

The structures of W- and W-Mo-containing high-entropy borides (HEBs) are systematically studied by combining atomic-resolution transmission electron microscopy imaging, electron diffraction, and chemical analysis. Here, we reveal that W or W-Mo addition in HEBs leads to segregation of these elements to the grain boundaries (GBs). In the meantime, W- or W-Mo-rich precipitates also form along the GBs. Crystallographic analysis and atomic-scale imaging show that the GB precipitates in both W- and W-Mo-containing HEBs have a cube-on-cube orientation relationship with the matrix. With further strain analysis, the coherency of the precipitate/matrix interface is validated. Nanoindentation tests show that the simultaneous GB segregation and coherent precipitation, as a supplement to the grain hardening, provide additional hardening of the HEBs. Our work provides an in-depth understanding of the GB segregation and precipitation behaviors of HEBs. It suggests that GB engineering could be potentially used for optimizing the performance of high-entropy ceramics.

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2022

Cluster dynamics modeling of Mn-Ni-Si precipitates coupled with radiation-induced segregation in low-Cu reactor pressure vessel steels

Formation of precipitates enhanced or induced by irradiation causes hardening and embrittlement of nuclear structural materials. Post-irradiation microstructure characterization of reactor pressure vessel (RPV) steels has shown that precipitation can be strongly associated with radiation-induced segregation (RIS) of solutes on dislocations. However, RIS and precipitation have not been coupled in previous precipitation modeling of RPV steels. Here in this study, a new hybrid and spatially-dependent precipitation model is developed that couples cluster dynamics with RIS, providing an unique way to account for concurrent evolution of heterogeneous cluster densities as well as solute and point defect concentration profiles. The model is applied to study the segregation of Mn, Ni, and Si on dislocations and heterogeneous nucleation of Mn-Ni-Si rich precipitates (MNSPs) in a low-Cu RPV steel. The result shows that the onset of MNSP nucleation on dislocations occurs at the fluence of 2 x 10 23 n.m -2 . The number density and mean radius can reach ~ 10 24 m -3 and 1-2 nm at high fluence, respectively. These observations suggest that the role of RIS and heterogeneous nucleation can be significant for RPV steels under a high fluence of irradiation. Results of simulations adopting various dose rates and dislocation densities show an increased fraction of MNSPs on dislocations at the condition of low dose rate, high dislocation density, and high fluence.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Microstructural evolution and strengthening mechanisms in a heat-treated additively manufactured Al–Cu–Mn–Zr alloy

In this report the microstructural and strength evolution of an additively manufactured Al-8.6Cu-0.5Mn-0.9Zr alloy upon aging at 300, 350, and 400 °C is investigated. The strengthening phases of the alloy evolve significantly upon aging, with breakdown and spheroidization of the interconnected θ-Al 2 Cu network, dissolution of metastable θ'-Al 2 Cu precipitates, and precipitation of nanometric L1 2 -Al 3 Zr from a matrix supersaturated in Zr. In the peak-aged states, the alloy displays a favorable combination of strength and ductility, with a room-temperature yield strength of 314–341 MPa and ductility of 11–13%. The measured yield strengths for microstructures with different aging treatments are compared to predictions of yield strengths from grain boundary, solid solution, and particle strengthening contributions. The observed strain hardening behavior is related to fundamental precipitate and dislocation interactions. Comparison between predicted and measured strength values indicates a continued need for strengthening models specifically developed for the heterogeneous microstructures of additively manufactured 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 OSTI2021

Optimization of the microstructure and mechanical properties of electron beam welded high-strength medium-entropy alloy (NiCoCr) 94 Al 3 Ti 3

For this work, the weldability and effects of subsequent heat treatment on the microstructure and mechanical properties of the medium-entropy alloy (MEA) (NiCoCr) 94 Al 3 Ti 3 have been explored using electron beam welding (EBW). It was found that EBW joints consist of a supersaturated f.c.c. matrix with coarse, asymmetric columnar grains (17.7–24 μm). The fraction of twins and Σ3 n CSL boundaries in weld joints was only 10% of that in the base metal, a value which increased to 30% after aging at 800 °C. The aging also led to improvements in the hardness, tensile strength, and work hardening rate, a result of the precipitation of a high volume fraction (0.29–0.33) of ultrafine (15–17.4 nm dia.) coherent spherical L1 2 -structured particles. The contributions of the L1 2 nanoparticles and Hall-Petch strengthening to the yield strength were estimated to be 537–573 MPa and 105–123 MPa, respectively. The MEA shows excellent weldability and, after aging, good strength (YS~1098 MPa, UTS~1368 MPa) and elongation to failure (~11%) by ductile fracture.

36 MATERIALS SCIENCE↗
DOE OSTI2020

On the use of charged particles to characterize precipitation in irradiated reactor pressure vessel steels with a wide range of compositions

Nuclear reactor lifetimes may be limited by nano-scale Cu-Mn-Ni-Si precipitates (CRPs and MNSPs) that form under neutron irradiation (NI) of pressure vessel (RPV) steels, resulting in hardening and ductile to brittle transition temperature increases (embrittlement). Physical models of embrittlement must be based on characterization of precipitation as a function of the combination of metallurgical and irradiation variables. Here we focus on rapid and convenient charged particle irradiations (CPI) to both: a) compare to precipitates formed in NI; and, b) use CPI to efficiently explore precipitation in steels with a very wide range of compositions. Atom probe tomography (APT) comparisons show NI and CPI for similar bulk steel solute contents yield nearly the same precipitate compositions, albeit with some differences in their number density, size and volume fraction (f) dose (dpa) dependence. Yet, the overall precipitate evolutions are very similar. Advanced high Ni (>3 wt%) RPV steels, with superior unirradiated properties, were also investigated at high CPI dpa. For typical Mn contents, MNSPs have Ni 16 Mn 6 Si 7 or Ni 3 Mn 2 Si phase type compositions, with f values that are close to the equilibrium phase separated values. In contrast, in steels with very low Mn and high Ni, Ni 2-3 Si silicide phase type precipitate compositions are observed; and when Ni is low, the precipitate compositions are close to the MnSi phase field. Low Mn significantly reduces, but does not eliminate, precipitation in high Ni steels. A comparison of dispersed barrier model predictions with measured hardening data suggests that the Ni-Si dominated precipitates are weaker dislocation obstacles than the G phase type MNSPs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Correlating microstructure and mechanical properties of harvested high dose Zorita light water reactor internals

Here, in this study, microstructural studies and micro-mechanical testing of an ex-plant material harvested from the decommissioned pressurized water reactor (PWR) is carried out. Irradiated 304 stainless steel (SS) components were harvested by the Electric Power Research Institute, U.S. Nuclear Regulatory Commission, and members of the José Cabrera Nuclear Power Station (Zorita) in Spain. The bi-crystalline micro-tensile specimens, atom probe tomography (APT) tips, and transmission electron microscopy (TEM) lamellae were fabricated from the baffle plate materials irradiated to 0.05, 15 and 50 dpa to elucidate microstructural, microchemical, and local mechanical properties changes as a function of dose. TEM and APT studies reveal radiation induced Ni, Cr, and Si rich precipitates in the 15 dpa sample that become smaller and partially redissolve into the matrix of the 50 dpa sample (decreasing size and number density). Dislocation loop and cavity number density and size were quantified as a function of dose as well as swelling. Intergranular Ni and Si enrichments with concomitant Fe and Cr depletion were observed in the 15 dpa sample and were more pronounced in the 50 dpa sample. Micro-tensile testing performed in the scanning electron microscope (SEM) at room temperature and 300 °C shows that the material's local yield strength and ultimate tensile strength decreases with increased dose and elevated test temperature and provides further insights via localized strain mapping. Current and previous mechanical data was compared with calculated values using the dispersed barrier hardening model with inputs of dislocation loops, precipitates, and cavities from microstructural characterization.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Post-irradiation examination of commercial tantalum alloys following neutron irradiation

Post irradiation examination of two commercial tantalum alloys (based on Ta-8 %W) revealed significant irradiation hardening following neutron irradiation in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Tensile samples of commercial Ta alloys T-111 and ASTAR-811C were irradiated to a total neutron fluence (E > 0.1 MeV) in the range (2–4)x10 21 n/cm 2 (0.39–0.75 dpa) at an irradiation temperature near 900 °C. Tensile testing conducted at room temperature and at 800 °C revealed significant irradiation hardening under all test conditions. Samples that were tensile tested at room temperature exhibited brittle failure, whereas ductility was maintained for elevated temperature testing. The irradiation hardening and embrittlement noted for both alloys was attributed to irradiation-induced dislocation loop formation as confirmed through scanning transmission electron microscopy analysis. Corresponding energy dispersive x-ray spectroscopy revealed radiation-induced segregation of Hf in these loop populations. This Hf enrichment is shown to result in eventual co-precipitation of elongated (Hf,O)-rich precipitates along the dislocation loops. Further, these results show that irradiation hardening is enhanced via irradiation-enhanced precipitation at elevated irradiation temperatures. Consequently, the minimum recommended operating temperature window for Ta alloys should be increased to account for observed hardening/embrittlement at irradiation temperatures as high as 900 °C (1,173 K).

36 MATERIALS SCIENCE↗
DOE OSTI2026

Defect formation and transmutation behaviors in irradiated tungsten under thermal-neutron shielded and mixed spectrum conditions

Here, this study investigates the microstructural evolution of pure tungsten irradiated under thermal-neutron shielded and mixed spectrum conditions in the High Flux Isotope Reactor (HFIR). Four samples were irradiated at temperatures from 570 °C to 1130 °C up to 0.73 dpa. Neutron spectrum significantly influenced the accumulation of transmutation products, with Re+Os content estimated at ∼0.3–0.6% under thermal-neutron shielded conditions and ∼5.2% under the mixed spectrum condition. Irradiation temperature strongly influences tungsten’s microstructure, with dislocation loops and fine voids forming at lower temperatures and only larger voids and Re/Os segregation observed at higher temperature. Under thermal-neutron shielded conditions, dislocation loops and voids were observed at 570 °C and 790 °C. At the highest irradiation temperature (1130 °C), dislocation loops were no longer observed, while larger but less dense voids remained. Re and Os segregation to void surfaces was evident at 790 °C and 1130 °C, though no precipitation was observed. In contrast, under the mixed-spectrum condition, both spherical and needle-like Re/Os-rich precipitates were observed, frequently accompanied by large voids. Dislocation loops were not observed, but loop-like contrast within the precipitates suggests they may have nucleated on pre-existing loops. Irradiation-induced hardening was assessed for the shielded samples at 570 °C and 790 °C. Dispersed barrier hardening (DBH) analysis, based on TEM-resolved defects, revealed that voids were the dominant contributors to hardening, consistent with literature results. A schematic model is proposed to describe defect and precipitate evolution in tungsten under fusion-relevant transmutation-to-dpa conditions.

Dislocation loops↗
DOE OSTI2022

The effect of Ti content on age hardening and mechanical properties of uranium-titanium alloys

Here, the effect of Ti content on age hardening and the resulting mechanical properties are described for -quenched U-Ti alloys containing 0.3 wt.% to 2.0 wt.%Ti. Age hardening occurs between ~250°C and ~450°C. Overaging occurs at higher temperatures by cellular decomposition. Age hardening kinetics suggest that different mechanisms occur depending on Ti content and initial microstructure. Strengthening in 'a acicular martensites begins by the formation of Ti clusters which evolve into thin U 2 Ti disc shaped precipitates and later mature into continuous U 2 Ti rods beginning at peak hardness. The mechanism of hardening in 'b banded martensite is more elusive, as significant hardening occurs where atomic mobility is lower than that required for precipitate formation, similar to that reported for age hardening in ''b banded martensite in U-6%Nb. The activation energy for aging varies with Ti content and microstructure. In fully martensitic alloys containing 0.75% to 2.0%Ti it is in the vicinity of ~44 kcal/mole (184 kJ/mole). But it is lower in alloys containing less than 0.6%Ti where quenched microstructures are less than fully martensitic. Tensile ductility is high prior to aging, decreases with age hardening, is effectively zero at peak hardness, and remains low in overaged conditions. Attractive combinations of strength and ductility are best obtained in alloys containing 0.6% to 1.0%Ti which have been partially aged to fractional hardening levels no greater than ~0.6. This corresponds to the very early stages of aging, associated with clustering and the earliest stages of U 2 Ti disc formation. Alloys containing 0.45%Ti or less are not as responsive to age hardening. Alloys containing 1.5% and 2.0%Ti can be aged to higher strengths, but extreme quench rate sensitivity prevents them from being effectively heat treated in realistic section thicknesses.

36 MATERIALS SCIENCE↗
DOE OSTI2025

A novel microstructural approach for TRIP/TWIP β-Ti alloys using intermetallic precipitation

Interest in metastable β-Ti has grown, due to the exceptional work hardening caused by TRansformation and TWinning Induced Plasticity (TRIP/TWIP). A current challenge is increasing the yield strength, while maintaining the benefits of TRIP/TWIP. This study presents the first report of TRIP/TWIP in a metastable β-Ti alloy containing intermetallic compounds, specifically Ti silicide precipitates. Silicon (Si) additions enable the formation of silicide particles, which can be precipitated, dissolved, and controlled through tailored heat treatments. In-situ synchrotron x-ray diffraction during tensile testing reveals primary {332}<113> twinning activates first, followed by TRIP, leading to extensive work hardening and uniform elongation in the solutionized state. In conclusion, these findings open a novel microstructural approach to the design of multiphase TRIP/TWIP titanium alloys, relying on intermetallic precipitation rather than allotropes such as α or ω phase, and lay the groundwork for future studies on intermetallic precipitation strengthening.

Materials science↗
DOE OSTI2020

Hierarchical microstructures and deformation behavior of laser direct-metal-deposited Cu–Fe alloys

Cu25Fe75 and Cu50Fe50 (nominal composition, at. %) alloys were fabricated using laser direct metal deposition (DMD) based additive manufacturing technique. These alloys exhibit hierarchical microstructures with bi-phasic Cu and Fe dendrites that contain nanoscale precipitates of varying sizes and structures. In the Cu25Fe75 alloy, Fe dendrites contained nanoscale, coherent, metastable BCC Cu and semi-coherent FCC Cu precipitates while the Cu matrix had nanoscale, coherent, metastable FCC Fe precipitates. In the Cu50Fe50 alloy, Fe dendrites only contained nanoscale semi-coherent FCC Cu precipitates while the Cu matrix had nanoscale coherent metastable FCC Fe precipitates. Both alloys exhibited enhanced flow strengths in the range of 750 – 980 MPa and significant plasticity, in compression. Here, the Cu25Fe75 alloy had lower yield strength than Cu50Fe50 alloy but higher maximum compressive strength due to higher strain hardening resulting from slightly coarser dendrites with hierarchy of nanoscale precipitation.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Effect of Sn microalloying on the nucleation of L1 2 Al 3 Zr precipitates in a dilute aluminum-zirconium alloy

While L1 2 -Al 3 Zr nanoprecipitates provide a balance between strengthening and good electrical conductivity, the precipitation of L1 2 -Al 3 Zr in aluminum requires aggressive heat treatments. An improved age-hardening response was observed during isochronal aging of an Al-0.24Zr (wt%) alloy when microalloyed with Sn. A new mechanism termed Low melting point Element-Assisted Nucleation (LEAN) is proposed to explain the lower temperature nucleation of L1 2 -Al 3 Zr precipitates observed in this alloy based on the addition of a low melting point element, such as Sn. Characterization verified the first-principles density functional theory prediction that Zr and Sn atoms cluster during homogenization owing to the favorable binding energy of Zr-Sn-vacancy triplets. Direct microstructural observations revealed these clusters form Sn nanoprecipitates that assist the nucleation of L1 2 -Al 3 Zr at 200°C, where L1 2 -Al 3 Zr precipitation is not expected due to the low diffusivity of Zr atoms in Al. At higher temperatures (≳350°C), the acceleration of L1 2 -Al 3 Zr precipitation is driven by faster Zr diffusion in Al with Sn microalloying and the nuclei formed via the LEAN mechanism. In conclusion, this combination of mechanisms explains the improvement in age hardening through L1 2 -Al 3 Zr precipitation with Sn microalloying.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Multiscale hierarchical and heterogeneous mechanical response of additively manufactured novel Al alloy investigated by high-resolution nanoindentation mapping

Smart alloying and microstructural engineering mitigate challenges associated with laser-powder bed fusion additive manufacturing (L-PBFAM). A novel Al–Ni–Ti–Zr alloy utilized grain refinement by heterogeneous nucleation and eutectic solidification to achieve superior performance-printability synergy. Conventional mechanical testing cannot delineate complex micromechanics of such alloys. This study combined multiscale nanomechanical and microstructural mapping to illustrate mechanical signatures associated with hierarchical heat distribution and rapid solidification of L-PBFAM. The disproportionate hardening effect imparted by Al 3 (Ti,Zr) precipitates in the pool boundaries and the semi-solid zone was successfully demonstrated. Nanomechanical response associated with heterogeneity in particle volume fraction and coherency across melt pool was interpreted from nanoindentation force–displacement curves. The hardness map effectively delineated the weakest and strongest sections in the pool with microscopic accuracy. The presented approach serves as a high throughput methodology to establish the chemistry-processing-microstructure-properties correlation of newly designed alloys for L-PBFAM.

36 MATERIALS SCIENCE↗