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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 757 records · Page 42
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Characterization and Strain-Hardening Behavior of Friction Stir-Welded Ferritic Stainless Steel

In this study, friction stir-welded joint of 3-mm-thick plates of 409 ferritic stainless steel (FSS) was characterized in light of microstructure, x-ray diffraction analysis, hardness, tensile strength, ductility, corrosion and work hardening properties. The FSW joint made of ferritic stainless steel comprises of three distinct regions including the base metal. In stir zone highly refined ferrite grains with martensite and some carbide precipitates at the grain boundaries were observed. X-ray diffraction analysis also revealed precipitation of Cr{sub 23}C{sub 6} and martensite formation in heat-affected zone and stir zone. In tensile testing of the transverse weld samples, the failure eventuated within the gauge length of the specimen from the base metal region having tensile properties overmatched to the as-received base metal. The tensile strength and elongation of the longitudinal (all weld) sample were found to be 1014 MPa and 9.47%, respectively. However, in potentiodynamic polarization test, the corrosion current density of the stir zone was highest among all the three zones. The strain-hardening exponent for base metal, transverse and longitudinal (all weld) weld samples was calculated using various equations. Both the transverse and longitudinal weld samples exhibited higher strain-hardening exponents as compared to the as-received base metal. In Kocks–Mecking plots for the base metal and weld samples at least two stages of strain hardening were observed.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Gas nitriding behavior of refractory metals and implications for multi-principal element alloy design

Multi-principal element alloys (MPEAs) comprise a large, flexible compositional space that enables tuning of their chemistry, structure, and properties. Here, to facilitate the development of nitriding-based surface-enhancement strategies that harness a broad compositional space, this study examined the gas nitriding behavior of Hf, Mo, Nb, Ta, Ti, and Zr as a function of time, temperature (750 and 1000 °C), and nitriding potential (i.e. ammonia-to-hydrogen ratio). These metals were selected because they have a strong driving force to form nitrides, and appear in many promising refractory MPEA compositions. The nitriding temperatures were selected based on the phase transformation temperature of Ti and Zr, and the nitriding potentials were chosen such that all elements are expected to form nitrides. Mass gain measurements indicate that all six elements follow parabolic kinetics. The microstructure observations and quantitative microchemical analysis show formation of dense and well-adhered compound layers for Mo, Nb, and Ta. Thick diffusion zones appear in Hf, Ta, Ti, and Zr, and diffusion coefficients were fit to the composition profiles. Partial delamination of the compound layer occurred for Ti and Zr. Peak hardness values above 30 GPa are obtained in the dense compound layers, and the solute hardening of the underlying alloy is correlated with the nitrogen content. The results provide insight into the dynamics of nitride compound formation relative to interstitial dissolution of nitrogen, and are discussed in the context of MPEA composition and processing design.

36 MATERIALS SCIENCE↗
DOE OSTI2022

An investigation of the failure modes in U-10Mo monolithic fuel irradiated to high burnup

Here, this study investigated possible failure mode sequences in U-10Mo monolithic fuel irradiated to very-high-burnup. The U-Mo fuel plate used in this study was characterized using scanning electron microscopy (SEM), wavelength dispersive spectroscopy (WDS), and image analysis techniques to investigate how microstructural features may initiate crack formation and propagation through the microstructure, potentially leading to blistering. Distinctly large fission gas pores (FGPs) were observed to preferentially align 5-15µm away from the U-10Mo/Zr interface, parallel to the Zr diffusion barrier. Preferential growth, alignment, and interconnection of FGP could initiate blistering in the fuel plate by creating a fission gas channel, parallel to the U-Mo/Zr interface. FGP alignment in the fuel phase near the U-Mo/Zr interface is believed to be one of the precursors to crack and Type 2 blister formation in monolithic U-Mo fuels. The chemical maps revealed differences in fission product behavior near the U-Mo/Zr interface such that Nd remains in the fuel matrix at the FGP sites. On the other hand, other fission products like Xe and Cs can get trapped in FGPs by Nd in the fuel phase and/or diffuse toward the Zr diffusion barrier.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Electroless nickel – phosphorus coating on crab shell particles and its characterization

Being hydrophilic material, crab shell particles have only a limited number of applications. It is, therefore, necessary to modify the surface of the crab shell particles. To make them useful ever for the applications, the main theme we proposed in this article is to utilize crab shell particles (CSP) with the core coated with nickel phosphorus (NiP) as a shell using the electroless coating process. For dealing with serious environmental problems, utilization of waste bio-shells is always an important factor to be considered. Chelating ability of crab shell particles eliminates the surface activation in this work proceeding to the coating process. The functional group, phase structure, microstructure, chemical composition and thermal analysis of CSP and NiP/CSP were characterized using Fourier transform infra-red spectroscopy (FTIR), x-ray diffraction analyzer (XRD), scanning electron microscope (SEM), energy-dispersive x-ray spectroscopy (EDS), and thermogravimetric analysis (TGA). The combination of an amorphous and crystalline structure was exhibited by CSP and NiP/CSP. NiP/CSP has shown a better thermal stability when compared to uncoated CSP. Stability test, adsorption test, and conductivity test were conducted for the study of adsorption behavior and conductivity of the particles. CSP presented a hydrophilic property in contrast to hydrophobic NiP/CSP. NiP/CSP presented a conductivity of about 44% greater compared to the CSP without any fluctuations. - Highlights: • Utilization of crab shell waste is focused on. • NiP coating on crab shell particle is fabricated using electroless process. • Thermal analysis, stability test, adsorption test and conductivity test were done. • Organic matrix of crab shell particle favors the coating process. • Results demonstrate the characterization of CSP core – NiP shell structure.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

Tracking Dendritic Growth in Hydrogen-Based Hematite Reduction via Computer Vision

The reduction of hematite to metallic iron using hydrogen (H2) as a reducing agent presents a promising pathway for decarbonizing steel production. In this study, we employ a combination of in situ confocal scanning laser microscopy (CSLM) and advanced computer vision techniques to quantitatively analyze dendritic growth of ferrite during H2-based reduction of iron oxide at high temperatures. A workflow integrating Watershed Image Segmentation (WIS) and Lucas-Kanade Optical Flow (LKOF) is developed to extract both global and local kinetic information from time-resolved micrograph sequences. H2 reduction experiments conducted at 1400 degrees C and 1500 degrees C demonstrate a clear correlation between temperature and reduction rate, as evidenced by accuracy of fitted Johnson-Mehl-Avrami-Kolmogorov (JMAK) parameters. Optical flow analysis further elucidates the anisotropic and branched nature of dendritic growth, providing spatially resolved velocity fields that correlate well with global transformation kinetics. The proposed methodology demonstrates strong agreement with experimental measurements and literature values, offering a robust framework for automated image-based analysis to study kinetics through microstructural evolution in the reduction of iron ore, and likely other reaction-diffusion phenomena.

08 HYDROGEN↗
DOE OSTI2014

Study of microstress state of P91 steel using complementary mechanical Barkhausen, magnetoacoustic emission, and X-ray diffraction techniques

The paper deals with assessment of microstress state of martensite P91 steel using three complementary techniques: mechanical Barkhausen emission, magnetoacoustic emission (MAE), and X-ray diffraction (XRD) profile analysis. Magnetic coercivity Hc and microstructure were investigated with inductive magnetometry and magnetic force microscopy (MFM), respectively. Internal stress level of P91 steel was modified by heat treatment. Steel samples were austenitized, quenched, and then tempered at three temperatures (720 °C, 750 °C, and 780 °C) during increasing time (from 15 min up to 240 min). The microstrain level ε{sub i} was evaluated using Williamson–Hall method. It was revealed that during tempering microstrain systematically decreases from ε{sub i} = 2.5 × 10{sup −3} for as quenched state down to ε{sub i} = 0.3 × 10{sup −3} for well tempered samples. Both mechanical hardness (Vicker's HV) and magnetic hardness (coercivity) decrease almost linearly with decreasing microstrain while the MAE and MBE intensities strongly increase. Tempering leads to evident shift of the MeBN intensity maximum recorded for the first load towards lower applied strain values and to increase of MAE intensity. This indicates that the microstress state deduced by magnetic techniques is correlated with microstrains evaluated with XRD technique.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Probing Plasticity and Strain-Rate Effects of Indium Submicron Pillars Using Synchrotron Laue X-ray Microdiffraction

We present that mechanical behaviors and especially the strain-rate responses at the nanoscales of low melting temperature metals, such as indium have not been much studied. Indium is one of the key materials or alloy-components in advanced microelectronics and nanotechnology industry, and understanding their mechanical behaviors at nanoscales becomes increasingly important to ensure lifetime reliability of their applications in novel nanoscale devices or advanced systems (for packaging at the nanoscales, for instance). Synchrotron X-ray microdiffraction has been utilized to examine defect structures of nanoscale materials as well as their strain-rate responses. Nanoscale or advanced microelectronics packaging, for instance, require acceptable levels of drop test results. For these low melting temperature materials especially, this technique offers a unique advantage as conventional methods such as TEM and EBSD will expose the structure to high energy electron beams that may significantly alter the microstructure and defect structure during analysis. Using this approach, we found interesting differences in term of X-ray peak broadening after deformation with different strain rates, which could indicate differences in plasticity mechanisms in the submicron pillars of indium, which could be important for its applications in nanodevices. Finally, understanding these differences could lead to better control of mechanical properties of low melting temperature metals at the nanoscales and thus have important implications for nanodevice reliability.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Evaluation of a single cell and candidate materials with high water content hydrogen in a generic solid oxide fuel cell stack test fixture, Part II: materials and interface characterization

A generic solid oxide fuel cell (SOFC) test fixture was developed to evaluate candidate materials under realistic conditions. A commerical 50 mm x 50 mm NiO-YSZ anode supported thin YSZ electrolyte cell with lanthanum strontium manganite (LSM) cathode was tested to evaluate the stability of candidate materials. The cell was tested in two stages at 800oC: stage I of low (~3% H2O) humidity and stage II of high (~30% H2O) humidity hydrogen fuel at constant voltage or constant current mode. Part I of the work was published earlier with information of the generic test fixture design, materials, cell performance, and optical post-mortem analysis. In part II, detailed microstructure and interfacial characterizations are reported regarding the SOFC candidate materials: (Mn,Co)-spinel conductive coating, alumina coating for sealing area, ferritic stainless steel interconnect, refractory sealing glass, and their interactions with each other. Overall, the (Mn,Co)-spinel coating was very effective in minimizing Cr migration. No Cr was identified in the cathode after 1720h at 800oC. Aluminization of metallic interconnect also proved to be chemically compatible with alkaline-earth silicate sealing glass. The details of interfacial reaction and microstructure development are discussed.

(Mn,Co)-spinel↗
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 OSTI2011

DFT Study Revises Interstitial Configurations in HCP Zr

Analysis of experimental result on microstructure evolution in irradiated Zr and alloys has demonstrated that available knowledge on self-interstitial defects in Zr is in contradiction. We therefore have initiated an extensive theoretical and modeling program to clarify this issue. In this report we present first ab initio calculations results of single SIA configurations in Zr. We demonstrate importance of simulations cell size, applied exchange-correlation functional and simulated c/a ratio. The results obtained demonstrate clearly that the most stable configurations are in basal plane and provide some evidences for enhanced interstitial transport along basal planes. The results obtained will be used in generation a new interatomic potential for Zr to be used in large-scale atomistic modeling of mechanisms relevant for radiation-induced microstructure evolution.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Investigation of laser direct energy deposition for production of ODS alloys

Laser blown powder additive manufacturing (AM) was performed to evaluate the feasibility of using the technique to fabricate oxide dispersion strengthened (ODS) alloys. The idea was to utilize an in-situ oxidation reaction to oxidize reactive elements to generate a refined dispersion of the dispersoids with a uniform size distribution. Builds were fabricated using Y added 316L in a reactive atmosphere and an inert atmosphere for comparisons. Detailed characterization was performed to understand the as-fabricated microstructure. Detailed characterization and chemical analysis show that there is significant Y loss during deposition. The loss in Y during processing leads to a reduced volume fraction of precipitates (~2 x 10 18 /m 3 ) with an average size of ~97 nm. Mechanical testing of the samples fabricated in both air and in inert atmosphere were performed at RT, 200 °C, 400 °C and 600 °C. The mechanical properties were then compared with austenitic ODS alloys that were reported in the literature. The results show that while the ultimate tensile strengths were comparable with wrought ODS alloys, the deterioration in yield strength with temperature was more severe in the case of the AM fabricated ODS material. However, the ductility was observed to be higher than all wrought 316L ODS alloys. Future efforts will focus on increasing the percentage of Y in the pre-alloyed powder to refine the dispersoids. In addition, a close control of process parameters and post processing treatments need to be explored to optimize the dispersoids volume fraction.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

ZrN Phase Formation, Hardening and Nitrogen Diffusion Kinetics in Plasma Nitrided Zircaloy-4

Plasma nitridation was conducted to modify the surfaces of Zircaloy-4. Scanning electron microscopy (SEM), transmission electron microscopy (TEM), and Raman analysis were used to characterize microstructures and phases. Surface indentation and cross-sectional indentation were performed to evaluate mechanical property changes. Nitridation forms a thin layer of ZrN phase, followed by a much deeper layer affected by nitrogen diffusion. The ZrN phase is confirmed by both TEM and Raman characterization. The Raman peaks of ZrN phase show a temperature dependence. The intensity increases with increasing nitridation temperatures, reaches a maximum at 700 °C, and then decreases at higher temperatures. The ZrN layer appears as continuous small columnar grains. The surface polycrystalline ZrN phase is harder than the bulk by a factor of ~8, and the nitrogen diffusion layer is harder by a factor of ~2–5. The activation energy of nitrogen diffusion was measured to be 2.88 eV. The thickness of the nitrogen-hardened layer is controllable by changing the nitridation temperature and duration.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Post Irradiation Examination Results of the NT-02 Graphite Fins NUMI Target

The NT-02 neutrino target in the NuMI beamline at Fermilab is a 95 cm long target made up of segmented graphite fins. It is the longest running NuMI target, which operated with a 120 GeV proton beam with maximum power of 340 kW, and saw an integrated total proton on target of 6.1 1020. Over the last half of its life, gradual degradation of neutrino yield was observed until the target was replaced. The probable causes for the target performance degradation are attributed to radiation damage, possibly including cracking caused by reduction in thermal shock resistance, as well as potential localized oxidation in the heated region of the target. Understanding the long-termstructural response of target materials exposed to proton irradiation is critical as future proton accelerator sources are becoming increasingly more powerful. As a result, an autopsy of the target was carried out to facilitate post-irradiation examination of selected graphite fins. Advanced microstructural imaging and surface elemental analysis techniques were used to characterize the condition of the fins in an effort to identify degradation mechanisms, and the relevant findings are presented in this paper.

43 PARTICLE ACCELERATORS↗
DOE OSTI2017

Post Irradiation Examination Results of the NT-02 Graphite Fins NuMI Target

The NT-02 neutrino target in the NuMI beamline at Fermilab is a 95 cm long target made up of segmented graphite fins. It is the longest running NuMI target, which operated with a 120 GeV proton beam with maximum power of 340 kW and saw an integrated total proton on target of 6.1 x 10^20. Over the last half of its life, gradual degradation of neutrino yield was observed until the target was replaced. The probable causes for the target performance degradation are attributed to radiation damage, possibly including cracking caused by reduction in thermal shock resistance, as well as potential localized oxidation in the heated region of the target. Understanding the long-term structural response of target materials exposed to proton irradiation is critical as future proton accelerator sources are becoming increasingly more powerful. As a result, an autopsy of the target was carried out to facilitate post-irradiation examination of selected graphite fins. Advanced microstructural imaging and surface elemental analysis techniques were used to characterize the condition of the fins in an effort to identify degradation mechanisms, and the relevant findings are presented in this paper.

Accelerator, Target, Graphite, NuMI, FermiLab, Neu↗
DOE OSTI2025

Helium Release During Fracture and Granular Fragmentation of Rocks

Geogenic Helium-4 ( 4 He) in-situ increases locally in regions of large deformation generated naturally or anthropogenically. This gas release by deformation is a potential geochemical precursor signal for subsurface deformation. To evaluate the applicability of 4 He degassing for correlating deformation in different lithologies, we conducted high force crush tests, up to 97,800 N axial load, to assess the total 4 He released during fragmentation of the rocks. We observed that the highest 4 He released occurred in the sedimentary rocks and that release correlated strongly with lithologic age and U/Th content. Microstructural changes of the pre- and post-test rocks indicate that the degree of grain size reduction relates directly to the total 4 He released during crushing. The range of in-place 4 He was calculated based XRF measurements of uranium and thorium in each lithology, with the results indicating that the majority of the trapped 4 He was not released. However, the 4 He released by deformation depended upon how the each rock deformed during deformation and the degree of grain size reduction. We postulate that 4 He precursor signals can be used to understand subsurface deformation only if geomechanical and geochemical conditions for 4 He enrichment in a lithology are met.

Deformation signals↗
DOE OSTI2016

Microstructural Characterization of Irradiated Fuels and Materials for the United States High Performance Research Reactor Fuel Development Program and the Fuel Cycle Research and Development Program

In an effort to contribute to the development of advanced nuclear fuel for research reactors, monolithic and dispersion plate fuels will be examined to determine how irradiation parameters such as fission density, fission rate, and irradiation temperature, as well as mechanical parameters such as composition and fabrication, affect the final microstructure of the fuel, post irradiation. Electron Backscatter Detection (EBSD) analysis capabilities of irradiated materials will be developed as well as improvements on irradiated sample preparation in order to provide higher quality samples for SEM, FIB, and TEM. In addition, image analysis techniques developed at Idaho National Laboratory (INL) will be employed to examine microstructural features such as fission product precipitates and fission gas bubbles. The main goals are to further develop EBSD capabilities on irradiated material; characterize microstructure of irradiated fuel to support qualification; develop a consistent, effective process for image analysis using INL developed software; and provide repeatable, statistically-confident, critically assessed data to further modeling efforts. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Investigations of cations distributions and morphology of cobalt ferrite magnetic nanoparticles

Cobalt ferrite nanoparticles were synthesized by co-precipitation method and structural properties was investigated by X-ray diffraction (XRD) at room temperature. X-ray diffraction data was used to determine lattice parameter, X-ray density, distributions of cations among tetrahedral and octahedral sites, site radii, ionic radii and bond length of inverse spinel cobalt ferrite. XRD analysis revealed crystallinity and high intense peak correspond to cubic inverse spinel structure with average crystalline size measured by X-ray line profile fitting was found to be 13nm for most intense peak (311). The surface morphology and microstructural feature was investigated by TEM analysis which revealed that particle size varying from 12-22 nm with selected electron diffraction pattern (SAED).

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2021

Carbonation of model cement pastes: The mineralogical origin of microstructural changes and shrinkage

This study explored the reactive processes of atmospheric carbonation and the consequences with respect to cementitious materials. Two model pastes were used: hydrated C{sub 3}S (including C-S-H and portlandite) and a paste prepared by hydrating a blend of C{sub 3}S and nanosilica (including C-S-H only). The two pastes were carbonated under accelerated conditions in the laboratory. The resulting mineralogical assemblage was examined using X-ray diffraction, thermogravimetric analysis and nuclear magnetic resonance. The microstructural changes were studied by X-ray tomography and porosimetry, and their macroscopic impacts were evaluated through gas diffusion and shrinkage measurements. The use of model pastes allowed for the evaluation of the change in solid volume induced by the carbonation of C-S-H. C-S-H decalcification and subsequent silica chain polymerisation were found responsible for carbonation shrinkage (and potentially cracking). Finally, the results highlight the protective role of portlandite: portlandite helped in limiting C-S-H decalcification and then reducing carbonation shrinkage and cracking.

36 MATERIALS SCIENCE↗