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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 811 records · Page 45
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Two-zone heterogeneous structure within shear bands of a bulk metallic glass

Shear bands, the main plastic strain carrier in metallic glasses, are severely deformed regions often considered as disordered and featureless. Here we report the observations of a sandwich-like heterogeneous structure inside shear bands in Pd{sub 40.5}Ni{sub 40.5}P{sub 19} metallic glass sample after plastic deformation by high-resolution transmission electron microscopy. The experimental results suggest a two-step plastic deformation mechanism with corresponding microstructure evolution at atomic scale, which may intimately connected to the stability of the shear band propagation and the overall plastic deformability.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Development of Nano-Modified Concrete for Next Generation of Storage Systems

The goal of the research was to develop a superior concrete for the long-term storage of used nuclear fuel by engineering concrete at the nanoscale through the incorporation of nano-sized and nano-structured particles based on enhanced reactivity (not fiber reinforcement). The specific objectives were to determine for nano-modified concrete: 1. The effects of environmental weathering (i.e., wet/dry and ingress of reactive species such as chloride and sulfate from the marine environment or other sources) and exposure to radiation and thermal environments on the structural stability and microstructural evolution of the material; 2. How microstructural and morphological changes from environmental weathering and thermal and radiation damage affect the long-term, macroscale performance of the material; and, 3. Identify formulation envelopes for nanoparticle content and combinations to achieve superior concrete criteria for dry storage systems by integrating and extending existing predictive modeling approaches that couple reactive transport and thermo- chemo- and radiation induced damage. The major findings of this work are summarized as follows. The results are being processed and compiled in manuscripts for peer-reviewed publication. Due to the late access to the gamma irradiation facility, material characterization is continuing. Dispersion/agglomeration of nanoparticles in cement pastes. The dispersion/agglomeration of nanoparticles (NPs) of SiO 2 , TiO 2 , Al 2 O 3 , Fe 2 O 3 , and clay (bentonite and halloysite) in water, with and without polycarboxylate-based high range water reducer (P-HRWR), and in the environment of hydrating cement have been investigated. The intrinsic particle characteristics, surface charge, pH, monovalent and divalent ion concentrations in solution, and modes of interaction of the NPs with P-HRWR influenced the NP behavior in solution.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

FFTF HT9 Cladding Microstructure Characterization

The sodium-cooled fast reactor (SFR) is a promising candidate for next generation nuclear reactors, operating at extreme conditions which include high temperatures (>500?C core outlet temperature) and significant neutron damage. High-Cr martensitic HT9 steel is an excellent candidate for SFR cladding and duct material due to its compatibility with liquid sodium, good thermal conductivity, resistance to void swelling, and strong creep rupture strength [1-4].However, the harsh in-core environment of SFRs can cause complex microstructural changes and mechanical property degradation in HT-9. Ensuring the safe use of HT9 cladding for metallic fuel requires both a thorough understanding of its mechanical response to microstructure evolution as well as reliable microstructure-sensitive modeling predictions. Microstructure-sensitive modeling of high temperature creep behavior in HT9 cladding for SFR applications currently lack experimental data to model the phenomena accurately. To fill this need, methods to perform microstructural characterization have been developed and performed on HT9.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Microstructure and phase evolution in the U-10Zr fuel investigated by in situ TEM heating experiments

The development of U-Zr metallic nuclear fuel for fast spectrum reactors is impacted by a lack of mechanistic understanding of the fuel behavior evolution under thermal irradiation conditions, despite previous works providing substantial fuel performance data. This work uses in-situ transmission electron microscopy heating experiments to study phase and microstructural evolution in several unirradiated U-10Zr specimens during rapid heating ramps (from room temperature to 1000 °C). The starting a-U + bcc-(Zr, U) eutectic microstructure began to decompose above 600 °C. The decomposition initiated from the bcc (U,Zr) phase. Similar results were observed for all specimens even when fabricated by different routes (e.g., cold rolled or annealed). As a result, the impact of observed microstructure and phase evolutions at high temperatures on fuel fabrication and in-pile fuel transient test was also discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Microstructural and magnetic characterization of iron precipitation in Ni-Fe-Al alloys

The influence of annealing on the microstructural evolution and magnetic properties of Ni{sub 50}Fe{sub x}Al{sub 50-x} alloys for x = 20, 25, and 30 has been investigated. Solidification microstructures of as-cast alloys reveal coarse grains of a single B2 type {beta}-phase and typical off eutectic microstructure consisting of proeutectic B2 type {beta} dendrites and interdendritic eutectic for x = 20 and x > 20 at.% Fe respectively. However, annealing at 1073 K results in the formation of FCC {gamma}-phase particles along the grain boundaries as well as grain interior in x = 20 at.% Fe alloy. The volume fraction of interdentritic eutectic regions tend to decrease and their morphologies start to degenerate by forming FCC {gamma}-phase for x > 20 at.% Fe alloys with increasing annealing temperatures. Increasing Fe content of alloys induce an enhancement in magnetization and a rise in the Curie transition temperature (T{sub C}). Temperature scan magnetic measurements and transmission electron microscopy reveal that a transient rise in the magnetization at temperatures well above the T{sub C} of the alloys would be attributed to the precipitation of a nano-scale ferromagnetic BCC {alpha}-Fe phase. Retained magnetization above the Curie transition temperature of alloy matrix, together with enhanced room temperature saturation magnetization of alloys annealed at favorable temperatures support the presence of ferromagnetic precipitates. These nano-scale precipitates are shown to induce significant precipitation hardening of the {beta}-phase in conjunction with enhanced room temperature saturation magnetization in particular when an annealing temperature of 673 K is used. - Research Highlights: {yields} Evolution of microstructure and magnetic properties with varying Fe content. {yields} Transient rise in magnetization via the formation of ferromagnetic phase. {yields} Enhancements in saturation magnetization owing to precipitated ferromagnetic phase. {yields} Nanoscale precipitation of ferromagnetic BCC {alpha}-Fe confirmed by TEM. {yields} Hardness of the {beta}-phase correlates with room temperature saturation magnetization.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Microstructural characterization of the oxidation of nuclear graphite under chronic and accident conditions via XCT and synchrotron XCT

Graphite is a candidate material to moderate fast neutrons and for structural components in the US next-generation graphite-moderated reactors. A graphite core is conceived as a large formation of interconnected bricks that primarily serves as the moderator of fast neutrons, holds instrumentation, fuel elements, control rods, and is a receptacle for molten salts for Molten Salt Reactors (MSRs) designs. During the operation of a graphite-moderated power plant, graphite components might be subjected to chronic oxidation during normal operating conditions or aggressive oxidation as a result of accidental ingress of air that reacts vigorously with the graphite core. Using synchrotron, x-ray computed tomography (XCT), this research systematically characterized microstructural changes that accompanied these two oxidation scenarios. Chronic oxidation was studied by characterizing IG-110, PCEA, and NBG-18 specimens that were gradually oxidized in air at a low temperature (520°C). The accidental ingress of air into the graphite reactor was simulated by subjecting the grades mentioned above to high-temperature oxidation at approximately 750°C. This research is the first, in situ, systematic characterization of nuclear graphite microstructural evolution that can be associated with the two possible oxidation scenarios and provide insight into related repercussions. The results show that the microstructure and pore connectivity influence the rate of oxidation and evolution of the microstructure under the two oxidation regimes. These results are crucial to understanding which grades are more resilient to each type of oxidation and documenting the damage created in the graphite.

Arregui Mena, Jose'↗
DOE OSTI2025

HPC for Optimizing Process Parameters to Control Material Evolution in Seamless Induction Hardening of Wind Turbine Main Shaft Bearings

Work proposed in this project focused on understanding the effect of martensitic transformation in the steel on the potential for cracking during seamless induction hardening (SIH) as a function of process conditions to allow the process to optimally scale up. Large-scale, three-dimensional phase-field simulations of martensitic transformation were performed using MEUMAPPS-SS (Microstructure Evolution Using Massively Parallel Phase-field Simulations – Solid State) code developed at Oak Ridge National Laboratory. The simulations were guided by location-specific thermal history generated by experimental measurements of time-temperature history generated at The Timken Company. The simulations were able to capture the morphological evolution of the martensite variants in an Fe-1.0C-1.5Cr steel based on the Nishiyama-Wasserman (NW) orientation relationship. The simulations were also able to quantify the stress-state at the interface between impinging martensite variants. The simulations indicated that the magnitude of the various stress and strain components were dependent on the sizes of the impinging plates with a reduction in these quantities with reduced plate size in agreement with experimental findings. The results obtained from the simulations will be used to guide the optimization of the alloy thermal conditions to eliminate quench cracking during SIH of bearing steels.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2019

Detailed Analysis of RUS Insertion Experiment and Scoping Studies for Performing Next Experiment using an Enriched Fuel Sample

Microstructure evolution due to irradiation in a nuclear reactor can have a dramatic effect on material properties. A better understanding of this evolution is necessary for developing improved nuclear fuels and materials. The ability to measure such changes in real time is extremely challenging due to high temperatures, high radiation fields, and limited access of the reactor environment. Through carefully designed experiments, measurement of elastic properties can be tied directly to microstructure. We present an instrument that has been developed to monitor in-pile changes in grain microstructure. The measurement approach involves optically exciting and detecting flexural waves in a thin cantilever beam. An instrument capsule based on this technique was fabricated and underwent an irradiation test in the TREAT reactor in May 2019. Analyses of the test results are presented in this report. Scoping studies on the expected impact of radiation on similar tests are also presented.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Effect of laser melt schedule on the microstructure of additively manufactured IN718 Superalloy

Laser powder bed fusion (L-PBF) has enabled the fabrication of geometrically complex metallic structures and components that are challenging to producing using conventional manufacturing approaches. The site-specific and far from equilibrium thermal conditions of L-PBF offer the potential to facilitate multi-length scale design of structure and properties across the atomic-through macro-levels. However, L-PBF systems face scalability challenges due to throughput constraints. Laser rotary powder bed fusion (L-RPBF) systems are being investigated as a solution to enhance the deposition rates compared to conventional L-PBF. Rotary systems also offer additional flexibility for controlling the time structure of melting through laser interleaving on alternating layers. Here, in this study, IN718 test samples were printed using single-laser or interleaved dual-laser configuration in a L-RPBF system to investigates the effect of process settings and melt-interleaving on as-fabricated microstructure. The microstructural evolution, such as grain size and crystallographic texture, was assessed by determining variations in the melt-pool shapes. Laser interleaving leads to a reduction in average grain size compared to single laser by ∼ 40 % at high power (400 W) and by ∼36 % at medium power (370 W). Results presented here identify key challenge for obtaining uniform microstructures and barriers for the broader adoption of high-deposition rate L-RPBF.

Dual-laser↗
DOE OSTI2012

Electron Backscatter Diffraction (EBSD) Characterization of Uranium and Uranium Alloys

Electron backscatter diffraction (EBSD) was used to examine the microstructures of unalloyed uranium, U-6Nb, U-10Mo, and U-0.75Ti. For unalloyed uranium, we used EBSD to examine the effects of various processes on microstructures including casting, rolling and forming, recrystallization, welding, and quasi-static and shock deformation. For U-6Nb we used EBSD to examine the microstructural evolution during shape memory loading. EBSD was used to study chemical homogenization in U-10Mo, and for U-0.75Ti, we used EBSD to study the microstructure and texture evolution during thermal cycling and deformation. The studied uranium alloys have significant microstructural and chemical differences and each of these alloys presents unique preparation challenges. Each of the alloys is prepared by a sequence of mechanical grinding and polishing followed by electropolishing with subtle differences between the alloys. U-6Nb and U-0.75Ti both have martensitic microstructures and both require special care in order to avoid mechanical polishing artifacts. Unalloyed uranium has a tendency to rapidly oxidize when exposed to air and a two-step electropolish is employed, the first step to remove the damaged surface layer resulting from the mechanical preparation and the second step to passivate the surface. All of the alloying additions provide a level of surface passivation and different one and two step electropolishes are employed to create good EBSD surfaces. Because of its low symmetry crystal structure, uranium exhibits complex deformation behavior including operation of multiple deformation twinning modes. EBSD was used to observe and quantify twinning contributions to deformation and to examine the fracture behavior. Figure 1 shows a cross section of two mating fracture surfaces in cast uranium showing the propensity of deformation twinning and intergranular fracture largely between dissimilarly oriented grains. Deformation of U-6Nb in the shape memory regime occurs by the motion of twin boundaries formed during the martensitic transformation. Deformation actually results in a coarsening of the microstructure making EBSD more practical following a limited amount of strain. Figure 2 shows the microstructure resulting from 6% compression. Casting of U-10Mo results in considerable chemical segregation as is apparent in Figure 2a. The segregation subsists through rolling and heat treatment processes as shown in Figure 2b. EBSD was used to study the effects of homogenization time and temperature on chemical heterogeneity. It was found that times and temperatures that result in a chemically homogeneous microstructure also result in a significant increase in grain size. U-0.75Ti forms an acicular martinsite as shown in Figure 4. This microstructure prevails through cycling into the higher temperature solid uranium phases.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Microstructural investigation of plastically deformed Ti{sub 20}Zr{sub 20}Hf{sub 20}Nb{sub 20}Ta{sub 20} high entropy alloy by X-ray diffraction and transmission electron microscopy

The microstructure evolution in body-centered cubic (bcc) Ti{sub 20}Zr{sub 20}Hf{sub 20}Nb{sub 20}Ta{sub 20} high entropy alloy during quasi-static compression test was studied by X-ray line profile analysis (XLPA) and transmission electron microscopy (TEM). The average lattice constant and other important parameters of the microstructure such as the mean crystallite size, the dislocation density and the edge/screw character of dislocations were determined by XLPA. The elastic anisotropy factor required for XLPA procedure was determined by nanoindentation. XLPA shows that the crystallite size decreased while the dislocation density increased with strain during compression, and their values reached about 39 nm and 15 × 10{sup 14} m{sup −2}, respectively, at a plastic strain of ~ 20%. It was revealed that with increasing strain the dislocation character became more screw. This can be explained by the reduced mobility of screw dislocations compared to edge dislocations in bcc structures. These observations are in line with TEM investigations. The development of dislocation density during compression was related to the yield strength evolution. - Highlights: • Ti{sub 20}Zr{sub 20}Hf{sub 20}Nb{sub 20}Ta{sub 20} high entropy alloy was processed by arc-melting. • The mechanical was evaluated by RT compression test. • The microstructure evolution was studied by XLPA and TEM. • With increasing strain the dislocation character became more screw. • The yield strength was related to the development of the dislocation density.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Scoping Studies to ascertain the change in the phase transition temperature of metallic fuels

Microstructure evolution due to irradiation in a nuclear reactor can have a dramatic effect on material properties. A better understanding of this evolution is necessary for developing improved nuclear fuels and materials. In the current scoping study we evaluate the influence of irradiation induced defects on the recrystallization temperature of a single component fuel surrogate.

36 MATERIALS SCIENCE↗
DOE OSTI2012

Multiscale Simulation of Thermo-mechanical Processes in Irradiated Fission-reactor Materials

This report contains a summary of progress made on the subtask area on phase field model development for microstructure evolution in irradiated materials, which was a part of the Computational Materials Science Network (CMSN) project entitled: Multiscale Simulation of Thermo-mechanical Processes in Irradiated Fission-reactor Materials. The model problem chosen has been that of void nucleation and growth under irradiation conditions in single component systems.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Cavity evolution and void swelling in dual ion irradiated tempered martensitic steels

Microstructural evolutions in two variants of the F82H 8 wt%Cr tempered martensitic steel under dual (He + plus Fe 3+ ) ion irradiations at 500 °C have been characterized by transmission electron microscopy. Emphasis is on understanding the combined effects of displacements per atom (dpa) and the helium/dpa ratio (He/dpa) on cavity evolution and void swelling (f v ). The dual ion database includes 231 alloy-dpa-He/dpa conditions for irradiations up to ~ 82 dpa and 3700 appm helium. Here, we hypothesize that the large “scatter” observed in the f v data is primarily due to local microstructural variations, especially those that shield regions at shallower depths from the effects of injected interstitial atoms, which would otherwise reduce f v . More generally, our hypothesis is that the highest f v data, at a specified dpa and He/dpa, provides the most appropriate estimate of swelling for fusion neutron irradiation conditions. The dose dependence of f v can be described by an incubation dpa i , marking the onset of significant void growth, and a post-incubation swelling rate, f v ’ = Δf v /Δdpa. Both dpa i and f v ’ decrease with increasing He/dpa. The f v ’ data are quantitatively consistent with a simple model for defect partitioning between evolving bubble, void, and dislocation sinks strengths and biases. Most notably, the model correctly predicts both the incubation dpa i , and the subsequent f v ’, in spite of the fact that they reflect somewhat different physics. Our analysis also suggests that defect recombination is modest at 500 °C. Along with other considerations, this supports the hypothesis that, even at high dpa rates, dual ion irradiations can be used to reasonably emulate void nucleation and growth for a range of fusion-relevant neutron irradiation service conditions. A companion paper incorporates the results of this study in a reduced-order, data-driven void swelling model.

36 MATERIALS SCIENCE↗
DOE OSTI2012

A calibrated Monte Carlo approach to quantify the impacts of misorientation and different driving forces on texture development

A calibrated Monte Carlo (cMC) approach, which quantifies grain boundary kinetics within a generic setting, is presented. The influence of misorientation is captured by adding a scaling coefficient in the spin flipping probability equation, while the contribution of different driving forces is weighted using a partition function. The calibration process relies on the established parametric links between Monte Carlo (MC) and sharp-interface models. The cMC algorithm quantifies microstructural evolution under complex thermomechanical environments and remedies some of the difficulties associated with conventional MC models. After validation, the cMC approach is applied to quantify the texture development of polycrystalline materials with influences of misorientation and inhomogeneous bulk energy across grain boundaries. The results are in good agreement with theory and experiments.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2017

Atomic-level heterogeneity and defect dynamics in concentrated solid-solution alloys

Performance enhancement of structural materials in extreme radiation environments has been activelyinvestigated for many decades. Traditional alloys, such as steel, brass and aluminum alloys, normallycontain one or two principal element(s) with a low concentration of other elements. While these existin either a mixture of metallic phases (multiple phases) or in a solid solution (single phase), limited orlocalized chemical disorder is a common characteristic of the main matrix. Fundamentally different fromtraditional alloys, recently developed single-phase concentrated solid-solutio n alloys (CSAs) contain mul-tiple elemental species in equiatomic or high concentrations with different elements randomly arrangedon a crystalline lattice. Due to the lack of ordered elemental arrangement in these CSAs, they exhibit sig-nificant chemical disorder and unique site-to-site lattice distortion. While it is well recognized intraditional alloys that minor additions lead to enhanced radiation resistance, it remains unclear inCSAs how atomic-level heterogeneity affects defect formation, damage accumulation, and microstruc-tural evolution. These knowledge gaps have acted as roadblocks to the development of future-generation energy technology. CSAs with a simple crystal structure, but complex chemical disorder,are unique systems that allow us, through replacing principal alloying elements and modifying concen-trations, to study how compositional complexity influences defect dynamics, and to bridge the knowl-edge gaps through understanding intricate electronic- and atomic-level interactions, mass and energytransfer processes, and radiation resistance performance. Recent advances in defect dynamics and irradi-ation performance of CSAs are reviewed, intrinsic chemical effects on radiation performance arediscussed, and direction for future studies is suggested.

Zhang, Yanwen↗
DOE OSTI2025

Microstructurally validated stable and predictable swelling in low-enriched uranium monolithic U-10Mo fuel mini-plates

Qualification of the low-enriched uranium (LEU) monolithic U-10 wt%Mo (U-10Mo) plate-type fuel system requires a demonstration of a stable and predictable fuel swelling behavior over the anticipated operating conditions of the United States high-performance research reactors (USHPRRs) selected for conversion to LEU operation. This will allow each reactor to develop appropriate safety margins that will retain fuel element lifetime coolability. Additionally, the fuel system must maintain performance attributes when fabricated at a commercial scale. The Mini-plate 1 experiment represents the first irradiation test of commercially fabricated miniaturized monolithic LEU U-10Mo fuel plates. Here, the swelling behavior within this experiment was compared against that of historical fuel developmental tests to reveal that the commercially fabricated fuel performed within the current recommended U-10Mo swelling model's predictions. Additionally, the fuel microstructural evolution was evaluated to link initial conditions to subtle variations detected in the swelling response, providing validation and confidence that the fuel system is robust.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Features of the microstructure development under conditions, reproducing the process of friction stir welding. Molecular-dynamics study

Friction stir welding is a recently developed technology which is used in various branches of modern engineering. The basis of this technology is the friction of the rotating cylindrical or specially shaped tool between two metal plates brought together either to meet their ends of one above another with the overlap. When applying the FSW process in various economical sectors, the important task is to study the mechanisms and identify the physical laws and factors leading to formation of structural inhomogeneities and discontinuities in the weld seam. This paper analyzes the basic mechanisms behind the structural state generation in the material subjected to severe plastic deformation and heating. To investigate the atomic mechanisms of structural changes in FSW, the modeling at atomic scale has been carried out. Results of work can be a basis for new knowledge about the microstructure evolution in FSW.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗