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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 1,261 records · Page 70
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Fatigue of additive manufactured Ti-6Al-4V, Part I: The effects of powder feedstock, manufacturing, and post-process conditions on the resulting microstructure and defects

Additive manufacturing provides an appealing means to process titanium alloy parts with new levels of conformability, complexity, and weight reduction. However, due to the heating/cooling rates and heat transfer associated with directed energy source material processing, the as-built AM parts contain unique material and microstructural features. In order to confidently manufacture fatigue critical additive manufactured (AM) Ti-6Al-4V parts, a better understanding of the interrelationships between powder feedstock, AM processes, structure of the processed parts, their resulting mechanical properties, and their performance under realistic loadings is necessary. Part I of this two-part collection focuses on the powder-process-structure relationships and how powder feedstock, manufacturing, and post-processing conditions can affect the microstructure and defect features that ultimately contribute to the fatigue performance of Ti-6Al-4V parts. The material and physical phenomena inherent to the AM process of Ti-6Al-4V are discussed in detail and related to the phase composition/structure, grain morphology, surface characteristics, defect size/distribution, and post-process treatments available for AM parts. This investigation is the foundation for the structure-performance relationships that will be covered in detail in Part II.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Gradient microstructure induced shear band constraint, delocalization, and delayed failure in CuZr nanoglasses

The mechanical properties of metallic glasses (MGs) can be significantly tuned by introducing a nanoscale glass-glass interface microstructure. Here, we design three types of Cu 64 Zr 36 MG seamless gradient microstructures by combining glassy grain sizes from 3 to 15 nm. We perform tensile loading molecular dynamics simulations on the named gradient nanoglasses to characterize their mechanical behavior. In all cases, results indicate the generation of gradient plasticity following the gradient in grain size, i.e. local plasticity increases from large to small grain size regions. Loading parallel to the gradient direction results in initial broad necking followed by catastrophic failure. In contrast, loading perpendicular to the gradient direction results in diffused shear band propagation from large to small grain size regions, delaying the generation of a dominant shear band and further delocalizing plastic deformation. In conclusion, the results demonstrate the synergistic effects generated by heterogeneous gradient designs in the mechanical behavior of metallic glasses.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Elevated temperature microstructure evolution of a medium-entropy CrCoNi superalloy containing Al,Ti

A new medium-entropy superalloy was produced based on the compositions of equiatomic CrCoNi and Ni-base superalloy Inconel 740H. Initial alloy design was performed using Thermo-Calc. The aging response and microstructural stability were assessed following heat treatment at temperatures between 600 and 900 °C and durations up to 100 h. Aging from a fully recrystallized state resulted in negligible grain growth and produced γ’ and σ phases. The same phases were present after aging from a cold-rolled state, but partially recrystallized microstructures resulted in multi-modal size distributions and heterogeneous spatial arrangements. Finally, room temperature hardness measurements were used to correlate aging conditions with quantitative precipitate measurements and mechanical properties.

36 MATERIALS SCIENCE↗
DOE OSTI2022

In-situ microstructure observation of oxidized SiC layer in surrogate TRISO fuel particles under krypton ion irradiation

In accidental scenarios of high temperature gas-cooled reactors, both oxidation of and irradiation to the SiC layer in tri-structural-isotropic (TRISO) fuel particles can change the microstructure and integrity of the fuel elements. Here, in the present study, microstructure and defect evolution in the oxidized SiC layer of surrogate TRISO fuel particles under 1.2 MeV krypton ion irradiation was observed by in-situ transmission electron microscopy. The SiC layers oxidized in water vapor at 1200 °C were irradiated azt room temperature and 800 °C and at damage levels of 0.28–11.2 dpa, respectively. SiC and SiO 2 were found to still be in their crystal structures at the damage level of 11.2 dpa at 800 °C, while SiC was observed to have been amorphized at only 0.56 dpa irradiation at room temperature. The defect number density at 800 °C was an order of magnitude lower than that in the sample irradiated at room temperature. Also, crystalline SiO 2 had higher radiation resistance compared to SiC. A defect reaction rate theory was utilized to understand the fundamental defect evolution process and irradiation resistance difference.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Microstructure analysis and machinability of additively manufactured A205 aluminum with heat treatments

A205 aluminum is one of the few high-strength aluminum alloys discovered for additive manufacturing (AM). The microstructure and heat treatment effects of AM A205 vary from those of its cast form, indicating possible differences in machinability during post-processing. Here, the objective of this research is to determine these variations of AM A205 and their effects on machinability using the cast condition as the baseline. As built and two heat-treated conditions (solution treatment and age hardening) were applied and compared in terms of their microstructures, micro-hardness, and machinability including vibration, specific cutting energy, chip morphology, and surface finish. The effects of print orientations on machinability were tested as well and found to be insignificant. AM as built had a higher hardness and slightly better machinability than its cast as built cast counterpart due to a smaller grain size (1 µm vs. 15 µm), eutectic θ-Al 2 Cu lattice and a fine, homogeneous TiB 2 distribution. The solution treatment dissolved this θ-Al 2 Cu lattice and increased the grain size (3 µm) which decreased the hardness, increased the ductility seen in the chips, and reduced the overall machinability. The presence of θ’-Al 2 Cu precipitates improved the machinability of aged AM A205. In addition, it was found that A205 exhibited both ductile and brittle behavior depending on the cutting speed, thus producing different machinability. The speed dependency is related to the material condition and heat treatment.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Densification, microstructure, and mechanical properties of Mo–30W alloys fabricated from conditioned powders

Refractory alloys, such as molybdenum-based systems, are attracting growing interest for applications in extreme environments, such as in the nuclear and aerospace industries. Recent advances in sintering technologies, coupled with mechanical alloying, have enabled the tailored design of these alloys by leveraging powder characteristics to control final microstructures and mechanical properties. In this study, Mo-30W alloys were fabricated using electric field-assisted sintering (EFAS) from ball-milled powders with and without hydrogen treatment to investigate the influence of surface oxides on material properties and sintering behavior. The results revealed that samples processed from as-ball-milled powder contained a high density of oxides within the microstructure, whereas oxide presence was significantly reduced in samples fabricated from hydrogen-treated powders. Interestingly, the two powder types led to opposite trends in grain size distribution: samples from untreated powders exhibited grain refinement from sample periphery to the center, while samples from hydrogen-treated powders showed grain coarsening toward the center. This behavior is attributed to temperature gradients present during sintering due to electrical percolation pathway differences during Joule heating. The powder surface oxides may have influenced the temperature distribution and grain evolution. Microhardness profiles measured along both axial and thickness directions were consistent with the grain size distribution. Furthermore, oxide films on powder surfaces have delayed densification by hindering particle necking and atomic diffusion during sintering.

36 - MATERIALS SCIENCE↗
DOE OSTI2014

Microstructural Modeling of Thermal Conductivity of High Burn-up Mixed Oxide Fuel

Predicting the thermal conductivity of oxide fuels as a function of burn-up and temperature is fundamental to the efficient and safe operation of nuclear reactors. However, modeling the thermal conductivity of fuel is greatly complicated by the radially inhomogeneous nature of irradiated fuel in both composition and microstructure. In this work, radially and temperature-dependent models for effective thermal conductivity were developed utilizing optical micrographs of high burn-up mixed oxide fuel. The micrographs were employed to create finite element meshes with the OOF2 software. The meshes were then used to calculate the effective thermal conductivity of the microstructures using the BISON fuel performance code. The new thermal conductivity models were used to calculate thermal profiles at end of life for the fuel pellets. These results were compared to thermal conductivity models from the literature, and comparison between the new finite element-based thermal conductivity model and the Duriez–Lucuta model was favorable.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Neutron energy spectrum influence on irradiation hardening and microstructural development of tungsten

We performed a neutron irradiation to single crystal pure tungsten in the mixed spectrum High Flux Isotope Reactor (HFIR). In order to investigate the influences of neutron energy spectrum, the microstructure and irradiation hardening were compared with previous data obtained from the irradiation campaigns in the mixed spectrum Japan Material Testing Reactor (JMTR) and the sodium-cooled fast reactor Joyo. The irradiation temperatures were in the range of ~90–~800 °C and fast neutron fluences were 0.02–9.00 × 10 25 n/m 2 (E > 0.1 MeV). Post irradiation evaluation included Vickers hardness measurements and transmission electron microscopy. Moreover, the hardness and microstructure changes exhibited a clear dependence on the neutron energy spectrum. The hardness appeared to increase with increasing thermal neutron flux when fast fluence exceeds 1 × 10 25 n/m 2 (E > 0.1 MeV). Finally, irradiation induced precipitates considered to be χ- and σ-phases were observed in samples irradiated to >1 × 10 25 n/m 2 (E > 0.1 MeV), which were pronounced at high dose and due to the very high thermal neutron flux of HFIR. Although the irradiation hardening mainly caused by defects clusters in a low dose regime, the transmutation-induced precipitation appeared to impose additional significant hardening of the tungsten.

36 MATERIALS SCIENCE↗
DOE OSTI2019

The effect of thermal-aging on the microstructure and mechanical properties of 9Cr ferritic/martensitic ODS alloy

In this paper, the thermal-stability of 9Cr F/M ODS alloy was evaluated by the aging treatment in the air at 700 degrees C for 10 h, 100 h, 1000 h, and 2005 h. The variations of mechanical properties with different aging periods were tested in terms of tensile testing and vickers microhardness. And the microstructural evolution of strengthening particles and grain morphologies in the 9Cr F/M ODS alloys were characterized using field emission scanning electron microscopy, transmission electron microscopy and high-energy X-ray diffraction. It was found that the as-rolled 9Cr F/M ODS alloy slightly recovered without recrystallization during the initial stage of less than 10 h aging time, resulting in the slight loss of tensile strength and microhardness and the increase of elongation. With the aging time prolonged to 2005 h, several of nano-sized Y-Ti-O particles coarsened and the composition changed to Y-Ti-Al-O, and the number density of large size nitride precipitations increased and the composition also changed from TiN to (Ti,AI)N. Meanwhile, some of grains tend to grow up in some preferential directions due to the zener pinning effect by the localized uneven distributed particles inside the grains, causing more particles, especially the particles with a size of larger than 40 nm, located at grain boundaries. These microstructural evolutions lead to the increasing of microhardness and a little bit decreasing of eleongation during 10 h-2005 h, and the increasing of tensile strength during 10 h-1000 h with the subsequent slight decreasing during 1000 h-2005 h. In conclusion, 9Cr F/M alloy shows a relatively good thermal-aging stability at 700 degrees C for 2005 h.

36 MATERIALS SCIENCE↗
DOE OSTI2020

On residual stress analysis and microstructural evolution for stainless steel type 304 spent nuclear fuel canisters weld joint: Numerical and experimental studies

Chloride-induced stress corrosion cracking (CISCC) in the weldments of austenitic stainless steel canisters is one of the primary safety concerns during the dry storage of used nuclear fuel at Independent Spent Fuel Storage Installations in coastal areas. In order to evaluate the CISCC behavior in the canister, a 3D sequentially coupled thermo-mechanical finite element model was built to simulate the residual stresses induced by two intersecting longitudinal and circumferential multi-pass welds in austenitic stainless steel mockup canister. Weld-induced residual stresses from simulation were in good agreement with the experimental measurement results by deep-hole drilling and contour methods. Through-wall high tensile axial and hoop stresses were observed in both the fusion zone and heat affected zone (HAZ) in the longitudinal and circumferential weld, respectively. On the other hand, the microstructure within the weldments of 304 stainless steel mockup was characterized using optical microscopy, scanning electron microscopy, electron back-scattered diffraction and energy dispersive X-ray spectroscopy methods. A few amount of MnS inclusions in both the base metal and HAZ, and a lot of (Mn,Si) x O y inclusions in the fusion zone were found. Meanwhile, there is some α'-martensite in the base metal, while the martensite was reversed in the HAZ. In the fusion zone, there is a lot of δ-ferrite. The martensite reversion phenomenon was also verified by the combination of Feritscope test and thermal simulation, which could predict a martensite reversion temperature as 400C°. The welding residual stresses and microstructure analysis results could offer the instructive information for the prediction of CISCC behavior of the spent nuclear fuel canisters.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Effect of stacking fault energy on damage microstructure in ion-irradiated CoCrFeNiMn concentrated solid solution alloys

In order to understand the effect of stacking fault energy on microstructural evolution in concentrated solid solution alloys, Kr 2+ ion irradiation was performed for FCC-type CoCrFeNiMnx alloys. During Kr 2+ ion irradiation at 500 degrees C the formation of black dots, self-interstitial atom faulted loops and perfect loops were observed, but no observable voids were found in all the alloys. The comparison of microstructural evolution among CoCrFeNiMn x (x = 0.7, 1, 1.3) alloys revealed less faulted loop number density and size in CoCrFeNiMn 1.3 . The stacking fault energy of each alloy was investigated using the deformed CoCrFeNiMn x alloys, and it was revealed that CoCrFeNiMn 1.3 had the highest stacking fault energy, which is higher value than that of 316SS. From these results, it is suggested that FeCrNiMn-based concentrated solid solution alloys could be well-designed by controlling the stacking fault energy with optimized Mn concentration as one of high radiation resistance structure materials. (C) 2020 Elsevier B.V. All rights reserved.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Friction stir welding and self-ion irradiation effects on microstructure and mechanical properties changes within oxide dispersion strengthened steel $\mathrm{MA956}$

We report the joining process for oxide dispersion strengthened (ODS) alloys remains a key challenge facing the nuclear community. The microstructure and mechanical properties were characterized in the base material and friction stir welded ODS MA956 irradiated with 5 MeV Fe 2+ ions from 400 to 500°C up to 25 dpa. Nanoindentation was performed to assess changes in hardness and yield stress, and the dispersed barrier hardening (DBH) model was applied to described results. A combination of scanning transmission electron microscopy (STEM) and atom probe tomography (APT) were used to assess evolution of the microstructure including dispersoids, network dislocations and dislocation loops, nanoclusters, and solid solution concentrations. Overall, softening was observed as a result of increased dose, which was exacerbated at 500°C. The formation and coarsening of new dispersoids was noted while nanoclusters tended to dissolve in the base material, and were not observed in the stir zone. Solute nanocluster evolution was identified as a primary driver of the changes in mechanical properties.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Effect of heat treatment on the microstructure of medium burn-up U-Mo monolithic fuel foils

Using scanning electron microscopy (SEM), this study evaluates the microstructure evolution of U-Mo fuel foils made with and without heat treatment at medium burn-up (of approximately 5 x 10 21 f/cm 3 ). The impact of annealing treatments on critical microstructural properties of the U-Mo fuel foils, including porosity, grain structure, Mo homogeneity, and fuel interaction with the Zr interlayer, was examined using large area lift outs (LALOs). The heat-treated specimens presented less grain refining at these burnups when compared to the un-heated specimens. Grain refinement was associated with porosities and fission products precipitation. Heat treatment can reduce fuel swelling during irradiation. Chemical inhomogeneity (Mo banding) was found to persist in the un-heated samples but was not present in the heat-treated samples. Thus, heat-treated U-Mo foils allows for more predictable fuel behavior under irradiation with respect to un-heated foils. The U-Mo and Zr interaction layer appears to be thicker and more continuous in the heat-treated sample which has been associated with stronger interface integrity during irradiation, as also observed in previous studies. These observations may indicate an overall improved performance of heat-treated fuel foil in a reactor. Further, the effect of local burn-up on grain size/refinement and porosities in each LALO specimen, sampled from different positions in the fuel foil, was difficult to analyze due to the large standard deviation of these parameters. Finally, evidence of grain refinement by polygonization may be present in these specimens.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Insight into the interfacial microstructure and chemistry of hot isostatically pressed AA6061-AA6061 bonds for U-10Mo fuel cladding application

Here, in the present study, hot isostatic pressing (HIP) bonding was carried out using AA6061-AA6061 plates for U-10Mo monolithic fuel cladding application. HIP was performed at 560°C with an applied isostatic pressure of 103 MPa for 90 min. Interfacial microstructure characterizations were performed using scanning electron microscopy (SEM) combined with electron backscatter diffraction (EBSD) to quantify precipitate fractions and study recrystallization, grain growth, and orientation relationships across the interface. EBSD results confirmed that no interfacial grain growth occurred during the HIP process. Transmission electron microscopy (TEM) was also employed to quantify composition and identify phases of various precipitates and an oxide layer that formed at the interface during HIP bonding. Microstructural analyses revealed the formation of a large fraction of Mg 2 Si precipitate and a ~50 nm thick Mg-oxide layer at the interface. Further TEM analysis confirmed that the oxide layer was not continuous, and Mg 2 Al 2 O 5 oxide particles with an average size of 18 nm were present in the oxide layer. Finite element modeling analysis was performed to study the stress distribution at the interface during HIP processing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

A critical analysis of U-Pu-Zr phase transitions using calorimetric, microstructural, and phase equilibria data

Metallic fuels consisting primarily of uranium, plutonium, and zirconium (U-Pu-Zr) are a leading material candidate for fast-spectrum nuclear reactors. Early demonstration programs proved the principle of safe and efficient fast reactor operation, however there is still considerable uncertainty regarding the phase equilibria and microstructural evolution across the ternary composition space. Quantitative phase formation and identification measurements are scarce and often incomplete, with studies reporting either phase transition temperatures or phase identification data, but not both from the same specimens. In this study, we critically compared experimental and calculated phase transition data and correlated with the microstructure and phase characterization data of as-cast and annealed U-Pu-Zr alloys. Differential scanning calorimetry (DSC) was used to measure phase transitions in the subsolidus regions (723−948 K) of three ternary U-Pu-Zr alloys with similar plutonium concentrations but various U/Zr ratios. Due to sluggish kinetics and narrow ranges of phase stability, complex peaks required the use of a Frazier-Suzuki peak fitting algorithm to deconvolute and calculate transition peak temperatures and enthalpies. We also identified trends of phase transition behavior by critically comparing our DSC data with previous phase transition measurements as well as historical and calculated phase equilibrium diagrams. In conclusion, this provides a critical approach for benchmarking and assessing the quality of new U-Pu-Zr phase equilibria data prior to its incorporation into nuclear material databases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Microstructural features and deuterium diffusion in lithium penta-aluminate pellets under He + and D + ion irradiation

Lithium (Li) penta-aluminate (LiAl 5 O 8 ) is investigated as a potential tritium (T) breeding material, with a focus on microstructural response to ion irradiation and deuterium (D) diffusion behavior. Under high-fluence ion irradiation (2 x 10 17 (He + +D + )/cm 2 ) at 773 K, LiAl 5 O 8 exhibits significant disorder on the Li sublattice, as revealed by atomic-resolution scanning transmission electron microscopy, while the Al and O sublattices remain stable, demonstrating strong resistance to structural amorphization. Irradiation induces the formation of platelet-shaped antiphase boundaries (APBs), which may serve as effective D trapping sites. Atom probe tomography suggests the presence of 6 LiD clusters in the mass spectra, though definite conclusions regarding APB composition are hindered by signal overlap and limited data statistics. Time-of-flight secondary ion mass spectrometry reveals that D retention approaches to saturation at 3 x 10 17 (He + +D + )/cm 2 . Isothermal and isochronal annealing studies determine an average diffusivity of 1.6 x 10 -13 at 773 K and an effective activation energy of 0.8 ± 0.1 eV for D migration. Compared to γ-LiAlO 2 , LiAl 5 O 8 demonstrates superior irradiation resistance, minimal Li loss, and enhanced D retention, underscoring its potential as a durable breeder material for T production. In conclusion, these findings provide key insights into the microstructural evolution, defect dynamics, and D retention mechanisms in LiAl 5 O 8 under reactor-relevant conditions.

42 ENGINEERING↗
DOE OSTI2025

Microstructural evolution and hardness changes in ion irradiated nickel-based Haynes 282 superalloy

This study investigated the crucial aspects of thermal and irradiation stability in precipitation hardened Haynes 282 Ni-based superalloy. The Haynes 282 Ni-based alloy was irradiated by 8 MeV Ni 3+ ions to assess its resistance to phase instabilities and mechanical property alterations. The mid-range doses (at a depth of ∼1 µm) were 1 and 10 displacements per atom (dpa) at temperatures of 600°C and 750°C. Nanoindentation tests provided insights into bulk equivalent hardness of the irradiated and pristine regions, while scanning transmission electron microscopy (STEM) and energy dispersive X-ray spectroscopy (EDS) were used to examine the microstructural evolution of irradiation-induced defects and γ′ -Ni 3 (Al, Ti) precipitates and defect structure under irradiation. These precipitates, with an average diameter of 29 nm and a number density of 5 x 10 21 /m 3 , acted as robust dispersion strengthening agents with high radiation point defect sink strength. Remarkably, irradiation did not significantly alter the size or number density of the γ′ precipitates, indicating exceptional thermal stability and radiation resistance of these precipitates at the examined conditions. Radiation-induced dislocation loops were observed at 600°C, albeit without a substantial impact on mechanical properties due to the dominance of γ′ precipitates on the overall alloy strength. The superior stability of γ′ precipitates observed in this study contrasts with several previous research findings on Ni-based alloys that reported poor precipitate stability. Plausible reasons for this difference are discussed. Moreover, this work explicitly outlines a physically grounded approach to ensure accurate microstructure-hardness correlations and clarifies the hardening model by addressing common misapplications of superposition in prior studies.

36 MATERIALS SCIENCE↗
DOE OSTI2026

High energy X-ray characterization of the microstructure of PuGa alloy samples at macroscopic depths and non-ambient conditions

High energy (95 keV) X-rays were utilized to characterize the microstructure of eight PuGa alloy samples at room temperature and during cooling to ∼10 K. The samples had different Ga content (nominally 0-3.4 at.%), age, and history. A ninth sample, similarly characterized with neutron diffraction, was included to extend the Ga range to 7 at.%. The samples span the range from single phase α (monoclinic) to δ (face-centered cubic) PuGa, as well as a two-phase α′/δ sample. The crystallographic textures, phase fractions, lattice parameters and dislocation densities of each sample were evaluated through distinct analysis techniques. The textures of the samples were modest. In each case, the dislocation densities were relatively high, comparable to cold worked metals and metals exposed to similar radiation dose. At room temperature, the lattice parameters determined in the predominantly single-phase samples were larger than expected based on the nominal Ga concentration. The observed lattice thermal expansions of the single-phase δ samples to 10 K are self-consistent across the samples and consistent with previous measurements reported in the literature. Significantly larger thermal lattice strain is observed in the δ phase of the two-phase material due to mechanical constraint from the α′ phase, which has a three times larger thermal expansion and is much harder than the δ phase. In conclusion, the results demonstrate the unique value of high-energy X-rays for relatively routine microstructural characterization of Pu alloy samples.

36 MATERIALS SCIENCE↗