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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 973 records · Page 54
All sourcesNASA NTRSDOE OSTI
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 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↗
DOE OSTI2021

Effects of sample bias on adhesion of magnetron sputtered Cr coatings on SiC

Swelling of SiC at 300 C due to in-service neutron irradiation causes tensile residual stresses in coatings which are expected to adversely affect the performance of coated SiC composite fuel cladding for light water reactors. Matching the coating swelling with the substrate, a solution common for thermal expansion, is not practical in the case of neutron irradiation. Biasing samples during magnetron sputtering deposition induces compressive residual stress which may counteract this. In this study, chromium coatings were deposited on SiC by DC magnetron sputtering with no external heating at bias voltages of –50V, –75V, and –100V. The effects of the bias voltage on morphology, residual stress, microstrain, texture, and adhesion are shown. The low deposition temperature resulted in the coating microstructure evolution following an energetic particle bombardment dominated trend. Additionally, at the two lower bias voltages knock-on implantation dominated increasing the residual stress and microstrain while at the highest bias voltage, thermal spike migration allowed for defect relaxation. When the knock-on induced compressive residual stress exceeded 0.8 GPa microcrack formation in the SiC substrate decreased coating adhesion. While no microcracks formed at the lowest bias voltage, insufficient atomic mobility during coating growth lead to voids forming in the coating. A balance is needed to form void-free coatings that have high compressive residual stress.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Void swelling of conventional and composition engineered HT9 alloys after high-dose self-ion irradiation

Ferritic/martensitic (F/M) steels are being considered as potential structural materials for next generation nuclear reactors, and variants of the alloy HT9 are some of the most promising candidates. In this study, two conventional and two composition engineered HT9 alloys were irradiated using 3.5 MeV Fe 2+ up to 600 peak displacement-per-atom (dpa) at 450 °C. Void swelling and microstructure evolution were characterized for each alloy and compared. The two conventional HT9 alloys (INL and ACO3) showed similar void swelling behavior due to their similar elemental composition and processing conditions. The INL HT9 exhibited a maximum of 2.4% swelling and the ACO3 HT9 showed a maximum of 2.8% swelling at 342 and 393 average local dpa, respectively. On the other hand, the two-composition engineered HT9 alloys with varying N contents (10 ppm for low N and 440 ppm for high N) showed disparate swelling behavior. The low N HT9 exhibited a maximum of 4.6% swelling, while the high N HT9 showed a maximum of 0.7% swelling at 342 average local dpa. Changes in the N content also affected Ni/Si rich G-phase formation. The low N HT9 showed a larger size and lower density of G-phase precipitates compared with the high N HT9 after 600 peak dpa irradiation. This study compares the void swelling behavior of the ion irradiated four current HT9 alloys to extremely high doses, with the void swelling data from neutron irradiated HT9 alloys. Furthermore, the comparison lends critical insights into how well these current alloys can withstand high neutron fluxes in future reactors, especially since the low N and high N HT9 alloys have never been exposed to such high doses before.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Irradiation-induced chemical disordering in ceramics: The case of SiC

Chemical disordering is one feature of damage that occurs in irradiation environments and is common in ceramic compounds. This paper reviews irradiation-induced chemical disordering in SiC-based materials at elevated temperatures. Here the results were obtained using advanced analytical tools to characterize atomistic defects and to evaluate the chemistry of defect clusters and homonuclear bond structures. Finally, this paper demonstrates that chemical disordering is crucial to understanding microstructural stability and establishing a mechanistic modeling of microstructural evolution of irradiated SiC.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Irradiation response of innovatively engineered metastable TRIP high entropy alloy

For this work, properties and radiation responses of a metastable high alloy (HEA) exhibiting the transformation induced plasticity (TRIP) effect were studied. Innovative engineering used to manufacture this HEA has shown superior mechanical and corrosion properties in 3.5% NaCl than most advanced stainless steels. The microstructural evolution and corresponding mechanical response after irradiation have been evaluated using detailed transmission electron microscopy, and nanoindentation. The study shows a change in metastability of the alloy with irradiation via a recovery mechanism, where the irradiation-induced transformation is reversed by the temperature-induced transformation, thereby introducing the concept of self-healing, made possible due to the TRIP behavior of HEA.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Atomistic simulations of He bubbles in Beryllium

Formation of He bubbles can have a significant effect on the microstructural evolution and properties of irradiated materials. In this report, we use atomistic simulations based on machine learning potentials to investigate the fundamental behavior of He bubbles in Be, with a specific focus on the shape, stability, and diffusivity of bubbles. Stability of He bubbles is quantified in terms of formation energies, which are determined as a function of the ratio of He/V. We find that He bubbles become unstable with respect to plastic deformation through punch-out dislocations around the bubble when the He/V ratio is larger than ~1.25, and the punch-out process induces the change of the regular bubble shape. In general, the bubble shape of He in Be is found to be ellipsoid-like. It is also found that for a fixed He/V ratio, the bubble attracts vacancies to become larger in size. If the bubble size is constant, the bubble attracts additional He atoms until the punch-out reaction occurs. The dominant diffusion mechanism of He bubbles changes from surface diffusion to volume diffusion as the temperature is increased, with a crossover occurring at about 900 K.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Manufacturing porous U-10Zr metallic fuels with controllable microstructure by volume control spark plasma sintering

In this paper, the volume control spark plasma sintering tool has been designed and applied to sinter porous U-10Zr metallic fuels, by which the sintered sample volume can be precisely controlled. Ethanol and NH 4 HCO 3 are used to control the powder compact or as pore formers to control the pore size and pore structure. Without pore formers, the fuel pellet displays an inhomogeneous microstructure consisting of highly porous and highly densified areas. Uneven powder stacking in the green body results in a non-uniform microstructure, and in the closed-packed area, Joule heating accelerates the neck formation and densification. The addition of ethanol reduces the friction between the powders, resulting in isolated pores formed by the stacking of powders during the sintering. By adding NH 4 HCO 3 , the pore size, and structure can be well controlled, and an interconnected pore structure can be obtained upon the decomposition of the NH 4 HCO 3 . Further, a uniform microstructure and pore distributions can be achieved through the U-10Zr fuel pellets by controlling current flow during the volume control SPS sintering. The microstructure and phase characterization of the sintered porous U-10Zr pellets show major phases of α-U and α-Zr for the sample with short dwelling. For the sample with long dwelling (30 min), the ω UZr 2 in the Zr-enriched area has been observed. The strategy of volume control SPS sintering with the assistance of pore formers could be used to fabricate porous U-10Zr metallic fuels to mimic the microstructure evolution of irradiated metallic fuels (including porosity) and could enable a possible solution for the design of new sodium-free metallic fuels for high burnup.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗