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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 1,135 records · Page 63
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Experimental characterization and atomistic simulation of grain boundary segregation in Mg-Y alloys

As a rare earth solute element in Mg alloys, Y has the beneficial effects of increasing both the strength and the ductility as well as weakening the crystallographic texture. To achieve a more fundamental understanding on how Y addition affects the microstructural evolution and mechanical properties, the Y segregation behavior at grain boundaries was investigated in Mg-1wt.%Y and Mg-7wt.%Y alloys at different conditions. The segregation intensity and its dependence on the grain boundary misorientation angle were experimentally characterized and computationally predicted. Strong segregation at grain boundaries was observed in both low and high Y-containing alloys. Y segregation was found to remain in alloy Mg-7Y after high-temperature annealing heat treatment at 540 °C. No direct correlation between the Y segregation intensity and the grain boundary misorientation angle could be established based on either the experimental characterization or the atomistic simulation with a spectral model. We thus conclude that grain boundary segregation of Y is independent of grain boundary misorientation angle.

Grain boundary↗
DOE OSTI2021

Friction Stir Dovetailing of AA7099 to Steel with AA6061 Interlayer for Reduced Zn Embrittlement at Dissimilar Interface

Joining of high strength Al alloy (AA7099) to steel (rolled homogeneous armor) with and without AA6061 interlayer was achieved using a friction stir dovetailing (FSD) method. Detailed investigations were conducted to establish correlations between mechanical performance and microstructural evolution of FSD joints. Lap shear tensile testing revealed that incorporation of the AA6061 interlayer combined with tailored processing methodology significantly improved both mechanical interlocking and load transmission across the dovetail interface. Following incorporation of the interlayer, the maximum load carrying capacity and peak extension (of the AA7099/AA6061/steel joint) are 60% and 210% higher, respectively, than results obtained from an AA7099/steel joint without an interlayer. Without an AA6061 interlayer, formation of zinc rich intermetallic compounds led to brittle interfacial failure of the AA7099-steel FSD joint. Incorporation of novel tool and joint configuration eliminated asymmetric joint performance as a function of loading direction, which is generally observed in friction stir lap welding. The efficacy of the use of an AA6061 interlayer in joining AA7099 to steel was demonstrated through mechanical testing as well as high resolution analytical electron microscopy characterization.

Dissimilar metal joining, Friction stir dovetailin↗
DOE OSTI2022

Manufacture aluminum alloy tube from powder with a single-step extrusion via ShAPE

Mechanical performance of aluminum in terms of strength, wear and corrosion resistance, especially high-temperature strength has been shown to improve with the addition of transition metal (TM) elements of Fe, Cr, and Ti. However, these feedstock materials occur as powders. As such, the traditional fabrication process is complex and expensive because it requires multiple procedures and consumes considerable energy. This study developed Shear Assisted Processing and Extrusion (ShAPE) as a single-step process that manufactures tubes directly from Al-TM powders obtained via gas atomization. Meter-long Al-TM alloy tubes are extruded from powders with different processing conditions. The ShAPE tubes have very low porosity and the average density is 2.94 kg/cm 3 , which is equal to or higher than parts fabricated by hot extrusion and sintering. The powder-to-tube fabrication process was revealed and discussed by examining the microstructural evolution. The Vickers hardness of ShAPE tubes ranges from 110 to 140 HV through wall thickness and at different extrusion speeds. The variation in hardness was correlated with the extent of refinement of intermetallics and attributed to the shear deformation per unit extrusion length. Energy cost analysis shows that ShAPE save ~60% energy compared to the conventional sintering and extrusion processes. Results indicate ShAPE is a low-cost, high-efficiency manufacturing process for producing tubes from metallic powders.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Relationship between nanostructure-magnetic property induced by temperature for iron oxide nanoparticles in vacuum, Ar and O 2 environments

It is known that superparamagnetic magnetite (Fe 3 O 4 ) nanoparticles (NPs) become ferromagnetic under Si 2+ ion irradiation due to particle size growth and microstructure evolution; it has been proposed that this feature could be used for in situ high temperature (up to 500 °C) radiation monitoring in the core of nuclear reactors. Herein, magnetite NPs synthesized by a nanocluster deposition system are heated to 800 °C in three different environments (argon, oxygen and vacuum), and nanostructure-magnetic property correlations are investigated by vibrating sample magnetometry, scanning electron microscopy, and X-ray diffraction. Magnetization of the NPs is increased due to the sintering and overall size growth by the agglomeration of the particles, while the morphology remains nearly unchanged up to 800 °C, with the one anomaly that zerovalent Fe appeared due to the reduction of the Fe 3 O 4 at 800 °C in vacuum. In argon and oxygen at high-temperature, antiferromagnetic hematite is created, which causes a reduction of the magnetization, and abrupt growth of particle size above 500 °C.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Link statistics of dislocation network during strain hardening

Dislocations are line defects in crystals that multiply and self-organize into a complex network during strain hardening. The length of dislocation links, connecting neighboring nodes within this network, contains crucial information about the evolving dislocation microstructure. By analyzing data from Discrete Dislocation Dynamics (DDD) simulations in face-centered cubic (fcc) Cu, we characterize the statistical distribution of link lengths of dislocation networks during strain hardening on individual slip systems. Here, our analysis reveals that link lengths on active slip systems follow a double-exponential distribution, while those on inactive slip systems conform to a single-exponential distribution. The distinctive long tail observed in the double-exponential distribution is attributed to the stress-induced bowing out of long links on active slip systems, a feature that disappears upon removal of the applied stress. We further demonstrate that both observed link length distributions can be explained by extending a one-dimensional Poisson process to include different growth functions. Specifically, the double-exponential distribution emerges when the growth rate for links exceeding a critical length becomes super-linear, which aligns with the physical phenomenon of long links bowing out under stress. This work advances our understanding of dislocation microstructure evolution during strain hardening and elucidates the underlying physical mechanisms governing its formation.

Crystal plasticity↗
DOE OSTI2026

Extended continuous cooling transformation diagram for 410 stainless steel encompassing δ → γ and γ → αʹ transformations determined by dilatometry

In this study, we use dilatometry to construct an extended Continuous Cooling Transformation (CCT) diagram for 410 stainless steel, capturing both δ → γ and γ → αʹ transformations specifically for directed energy deposition (DED) additive manufacturing (AM) applications. To ensure that the findings are representative of AM specific microstructures and compositions, dilatometry coupons were extracted from a wire arc DED wall, thereby preserving the inherent microstructural and compositional variability within the material. The on-heating phase transformations, including A c1 , A c3 , and γ → δ, were found to be essentially location independent, despite the typical heterogeneity associated with AM. Conversely, increasing cooling rates were observed to shift the start and finish temperatures of both δ → γ and γ → αʹ transformations to lower values. The resulting extended CCT diagram offers a more complete depiction of phase transformation behavior in the system, enabling accurate microstructure evolution modeling and residual stress predictions specifically for directed energy deposition additive manufacturing applications.

Additive manufacturing↗
DOE OSTI2021

Capture efficiency and bias from the defect dynamics near grain boundaries in BCC Fe using mesoscale simulations

The capture efficiency describes the capability of a sink, such as a grain boundary (GB), dislocation, and void, to absorb point defects (PDs). The bias defines the difference in capture efficiency between the absorption of a vacancy and dumbbell at a sink. Complete kinetic information on PDs, including diffusion barriers and diffusion orientations, as well as accurate saddle points, are needed to determine the capture efficiency and bias at a sink accurately, which is computationally demanding. Here, in the present study, the Self-Evolving Atomistic Kinetic Monte Carlo (SEAKMC) method was used to investigate the defect dynamics of PDs near different types of grain boundaries (GBs) (with both $\langle$100$\rangle$ and $\langle$110$\rangle$ families) accurately in body-centered cubic (BCC) iron (Fe). The capture efficiency, sink strength, and bias factor of different types of GBs were determined in Fe, which, different from traditional rate theory estimation, showed a distinct capture efficiency, sink strength, and bias in different GBs. The results demonstrate a strong positive correlation between the capture efficiency and the GB strain width, instead of the GB misorientation, GB energy, or GB-PD binding energy, which have been investigated previously. This work provides valuable insight into the radiation-induced microstructural evolution of GBs.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Dimensional isotropy of 6H and 3C SiC under neutron irradiation

This investigation experimentally determines the as-irradiated crystal axes dimensional change of the common polytypes of SiC considered for nuclear application. Single crystal α-SiC (6H), β-SiC (3C), CVD β-SiC, and single crystal Si have been neutron irradiated near 60 °C from 2 × 10 23 to 2 × 10 26 n/m 2 (E > 0.1 MeV), or about 0.02–20 dpa, in order to study the effect of irradiation on bulk swelling and strain along independent crystalline axes. Single crystal, powder diffractometry and density measurement have been carried out. For all neutron doses where the samples remained crystalline all SiC materials demonstrated equivalent swelling behavior. Moreover the 6H–SiC expanded isotropically. The magnitude of the swelling followed a ~0.77 power law against dose consistent with a microstructure evolution driven by single interstitial (carbon) mobility. Extraordinarily large ~7.8% volume expansion in SiC was observed prior to amorphization. Above ~0.9 × 10 25 n/m 2 (E > 0.1 MeV) all SiC materials became amorphous with an identical swelling: a 11.7% volume expansion, lowering the density to 2.84 g/cm 3 . As a result, the as-amorphized density was the same at the 2 × 10 25 and 2 × 10 26 n/m 2 (E > 0.1 MeV) dose levels.

36 MATERIALS SCIENCE↗
DOE OSTI2016

High energy X-ray diffraction study of the relationship between the macroscopic mechanical properties and microstructure of irradiated HT-9 steel

Samples harvested from an HT-9 fuel test assembly (ACO-3) irradiated for six years in the Fast Flux Test Facility (FFTF) reaching 2–147 dpa at 382–504°C were deformed in-situ while collecting high-energy X-ray diffraction data to monitor microstructure evolution. With the initiation of plastic deformation, all samples exhibited a clear load transfer from the ferrite matrix to carbide particulate. This behavior was confirmed by modeling of the control material. The evolution of dislocation density in the material as a result of deformation was characterized through full pattern line profile analysis. The dislocation densities increased substantially after deformation, the level of dislocation evolution observed was highly dependent upon the irradiation temperature of the sample. The differences in both the yield and hardening behavior between samples irradiated at higher and lower temperatures suggest the existence of a transition in tensile behavior at an irradiation temperature near 420°C dividing regions of distinct damage effects.

36 MATERIALS SCIENCE↗
DOE OSTI2016

The effect of low-temperature aging on the microstructure and deformation of uranium- 6 wt% niobium: An in-situ neutron diffraction study

The mechanical properties of uranium-niobium alloys evolve with aging at relatively low temperatures due to subtle microstructural changes. In-situ neutron diffraction measurements during aging of a monoclinic U-6Nb alloy at temperatures to 573 K were performed to monitor these changes. Further, in-situ neutron diffraction studies during deformation of U-6Nb in the as-quenched state and after aging for two and eight hours at 473 K were completed to assess the influence of microstructural evolution on mechanical properties. With heating, large anisotropic changes in lattice parameter were observed followed by relaxation with time at the aging temperature. The lattice parameters return to nearly their initial values with cooling. Finally, the active plastic deformation mechanisms including, in order of occurrence, shape-memory de-twinning, mechanical twinning, and slip-mediated deformation do not change with prior aging. However, the resistance to motion of the as-quenched martensitic twin boundaries increases following aging, resulting in the observed increase in initial yield strength.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Radiation stable, hybrid, chemical vapor infiltration/preceramic polymer joining of silicon carbide components

Here, this paper reports on a nuclear-grade joining material for bonding of silicon carbide-based components. The joint material is fabricated via a hybrid preceramic polymer, chemical vapor infiltration process. The joint is comprised entirely of β-SiC and results in excellent mechanical and permeability performance. The joint strength, composition, and microstructure have been characterized before and after irradiation to 4.5 dpa at 730 °C in the High Flux Isotope Reactor. The hybrid preceramic polymer-chemical vapor infiltrated joint exhibited complete retention of shear strength and no evidence of microstructural evolution or damage was detected following irradiation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Ion-irradiation-induced microstructural modifications in ferritic/martensitic steel T91

In this paper, in situ transmission electron microscopy investigations were carried out to study the microstructural evolution of ferritic/martensitic steel T91 under 1 MeV Krypton ion irradiation up to 4.2 x 10(15) ions/cm(2) at 573 K, 673 K, and 773 K. At 573 K, grown-in defects are strongly modified by black dot loops, and dislocation networks together with black-dot loops were observed after irradiation. At 673 K and 773 K, grown-in defects are only partially modified by dislocation loops; isolated loops and dislocation segments were commonly found after irradiation. Post irradiation examination indicates that at 4.2 x 1015 ions/cm(2), about 51% of the loops were a(0)/2 < 111 > type for the 673 K irradiation, and the dominant loop type was a(0)< 100 > for the 773 K irradiation. Finally, a dispersed barrier hardening model was employed to estimate the change in yield strength, and the calculated ion data were found to follow the similar trend as the existing neutron data with an offset of 100-150 MPa. (C) 2017 Elsevier B.V. All rights reserved.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Characterization of faulted dislocation loops and cavities in ion irradiated alloy 800H

Alloy 800H is a high nickel austenitic stainless steel with good high temperature mechanical properties which is considered for use in current and advanced nuclear reactor designs. The irradiation response of 800H was examined by characterizing samples that had been bulk ion irradiated at the Michigan Ion Beam Laboratory with 5 MeV Fe 2+ ions to 1, 10, and 20 dpa at 440 °C. Transmission electron microscopy was used to measure the size and density of both {111} faulted dislocation loops and cavities as functions of depth from the irradiated surface. The faulted loop density increased with dose from 1 dpa up to 10 dpa where it saturated and remained approximately the same until 20 dpa. The faulted loop average diameter decreased between 1 dpa and 10 dpa and again remained approximately constant from 10 dpa to 20 dpa. Cavities were observed after irradiation doses of 10 and 20 dpa, but not after 1 dpa. The average diameter of cavities increased with dose from 10 to 20 dpa, with a corresponding small decrease in density. Cavity denuded zones were observed near the irradiated surface and near the ion implantation peak. To further understand the microstructural evolution of this alloy, FIB lift-out samples from material irradiated in bulk to 1 and 10 dpa were re-irradiated in-situ in their thin-foil geometry with 1 MeV Kr 2+ ions at 440 °C at the Intermediate Voltage Electron Microscope. We observed that the cavities formed during bulk irradiation shrank under thin-foil irradiation in-situ while dislocation loops were observed to grow and incorporate into the dislocation network. The thin-foil geometry used for in-situ irradiation is believed to cause the cavities to shrink.

36 MATERIALS SCIENCE↗
DOE OSTI2018

In-situ TEM study of the ion irradiation behavior of U 3 Si 2 and U 3 Si 5

We report that U 3 Si 2 and U 3 Si 5 are two important uranium silicide phases currently under extensive investigation as potential fuel forms or components for light water reactors (LWRs) to enhance accident tolerance. In this paper, their irradiation behaviors are studied by ion beam irradiations with various ion mass and energies, and their microstructure evolution is investigated by in-situ transmission electron microscopy (TEM). U 3 Si 2 can easily be amorphized by ion beam irradiations (by 1MeV Ar 2+ or Kr 2+ ) at room temperature with the critical amorphization dose less than 1 dpa. The critical amorphization temperatures of U 3 Si 2 irradiated by 1MeV Kr 2+ and 1MeV Ar 2+ ion are determined as 580±10K and 540±5K, respectively. In contrast, U 3 Si 5 remains crystalline up to 8 dpa at room temperature and is stable against ion irradiation-induced amorphization up to ~50 dpa by either 1MeV Kr 2+ or 150KeV Kr + at 623K. In conclusion, these results provide valuable experimental data to guide future irradiation experiments, support the relevant post irradiation examination, and serve as the experimental basis for the validation of advanced fuel performance models

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Crystallographic and Compositional Analysis of impurity phase: U2MoSi2C in UMo Alloys

Minor impurity elements in metallic nuclear materials can critically influence the microstructural evolution and mechanical properties, making it crucial to understand the structure, composition and distribution of the impurity phases. Using Transmission Electron Microscopy, Atom Probe Tomography and ab-initio calculations we provide the atomic scale crystallographic structural and compositional analysis of a new impurity phase: U2MoSi2C in an important nuclear fuel U-10Mo alloy. We identify the new phase to have tetragonal symmetry with lattice parameters of a=6.67Å, c=4.33Å, and space group P4/mbm (No.127). Ab-initio calculations were performed to verify structural stability and to perform atom positions refinement.

Kovarik, Libor↗
DOE OSTI2021

Atom Probe Tomography Characterization of High-Dose Ion Irradiated MA957

This paper presents a comprehensive characterization study of the microstructural responses to ion irradiation of the oxide dispersion strengthened (ODS) ferritic alloy MA957 using atom probe tomography (APT). The MA957 specimens were irradiated by 1.8 MeV Cr3+ ions to doses and temperatures relevant to advanced fast reactors (100 and 500 dpa at 400-500 °C). The average sizes of the YTiO particles decreased slightly after irradiation at all temperatures, with the most pronounced change at 400 °C. The particle number densities for 400 and 500 °C were slightly below number densities for the unirradiated condition, while the values were slightly higher at 420 and 450 °C. This behavior is consistent with the Nelson-Hudson-Mazey rate theory model that considers the competing effects between recombination and dissolution. An increase in the Y/Ti ratio of the YTiO particles was observed at the two lower temperatures. The formation of Cr-rich precipitates (a') was observed in specimens irradiated at temperatures up to 450 °C. Irradiation-induced segregation on grain boundaries was also investigated. Grain boundary element concentration profiles and ion maps demonstrate differences among the irradiation temperatures, with most datasets differing from the unirradiated condition. A key focus of this ion irradiation study is the high-dose response, and this quantitative assessment revealed relatively small but distinct effects of irradiation on microstructure evolution. Overall, these results showing minor second phase evolution suggest that MA957 should retain mechanical performance generally well up to 500 dpa for the irradiation temperatures studied.

Wang, Jing↗