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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 649 records · Page 36
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Parameterization of vacancy production rate in phase-field models of fission gas bubble evolution in nuclear fuel

Phase-field modeling has increasingly been used to study microstructural evolution in fission gas bubbles in nuclear fuel to improve understanding of fission gas release. To improve computational efficiency, often only vacancies and gas atoms are included as defect species. In this case, the net effects of vacancy and interstitial production, recombination, and biased sink absorption are included as a net vacancy source, or net vacancy source combined with an effective sink. However, there has been a lack of clarity on what parameter values should be used for these approaches to best match the more complete physical picture that includes interstitials and vacancies. Here, we compare a phase-field model of void growth to analytical models for the source-only and source plus sink approach to gain insight into how the phase-field models can be parameterized effectively. The source-only approach provides greater flexibility to match growth rates determined from the full vacancy-interstitial picture. A strategy was developed for determining the value of the net vacancy source term by comparing to an analytical model that includes vacancy and interstitial production, recombination, and biased sink absorption. Finally, this strategy can be used to parameterize phase-field models of fission gas bubble growth.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Hydrostatic pressure effect on mechanical behavior and texture evolution of Al and Brass

The effect of hydrostatic pressure on shear strength and microstructural evolution of polycrystalline FCC metals was investigated. Hydrostatic pressure of up to 5 GPa was imposed on commercial purity aluminum and 70/30 brass samples using a modified opposed-anvil apparatus (tri-anvil) that allows for measurement of shear strength in thin foil specimens. Similar to the previous investigations made in BCC metals (Ta and Mo), the shear strength of FCC metals increases significantly as the pressure rises. At 5 GPa, the shear strength of aluminum increased to 8 times its value at atmospheric pressure and 70/30 brass increased by a factor of 2.7. Here, EBSD analysis reveals an evident accumulation of dislocations in all sheared samples, with an approximately 50% decrease in grain diameter. Texture analysis suggests that, in addition to helping form selectively oriented dislocation walls, hydrostatic pressure also serves as a threshold to select certain favorable orientations in sheared metals. We propose that these hydrostatic pressure effects are intrinsically due to the excess volume associated with the cylindrical strain field of dislocation lines.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Microstructure effects on the recrystallization of low-symmetry alpha-uranium

We employ electron backscatter diffraction (EBSD) to investigate microstructural evolution of uranium during recrystallization. To understand the relationship between microstructure and recrystallization, we use measures of intra-granular misorientation within grains and near grain boundaries in both deformed (non-recrystallized) uranium and recrystallizing uranium. The data show that the level of intra-granular misorientation depends on crystallographic orientation. However, contrary to expectation, this relationship does not significantly affect the recrystallization texture. Rather, the analysis suggests that recrystallization nucleation occurs along high angle grain boundaries in the deformed microstructure. Specifically, we show that the nucleation of recrystallized grains correlates well with the spatially heterogeneous distribution of high angle boundaries. Due to the inhomogeneous distribution of high angle boundaries, the recrystallized microstructure after long times exhibits clustered distributions of small and large grains. Twin boundaries do not appear to act as recrystallization nucleation sites.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Neutron irradiation induced defects and clustering in NF616 and T91

NF616 is a third-generation ferritic martensitic steel, developed to have better creep resistance than the prior generation T91. Unlike relatively numerous studies of T91, there is a lack of understanding of the irradiation effect on the microstructural evolution and mechanical response of NF616. This work evaluated the microstructures and radiation hardening of NF616 irradiated up to 8.2 displacement per atom (dpa) at 292 °C – 431 °C, compared with T91 from two heats. Dislocation loops were observed in all investigated samples. NF616 exhibited comparable loop size but slightly lower loop density than those in the general T91 heat at 430 °C. Cavities were only observed in NF616 at 431 °C but absent at lower irradiation temperatures (292 °C and 359 °C). Ni-rich clusters were also observed in NF616 at 431 °C, while only weak Ni-clustering were observed at lower irradiation temperatures. Compared to the general T91 heat, NF616 demonstrated better swelling resistance (e.g., one third of swelling in the general T91 heat at 430 °C), a slightly higher number density of Ni-rich clusters, and slightly lower radiation hardening. The low-carbon T91 showed the greatest hardening with the largest swelling and loop sizes, despite its lowest irradiation temperature and intermediate dose. The calculated hardening from loops, cavities and Ni-rich clusters using the classic dispersed barrier-hardening model had reasonable agreement with the experiment-derived results, with the primary hardening contribution attributed to dislocation loops.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

ExaAM: Metal additive manufacturing simulation at the fidelity of the microstructure

Additive manufacturing (AM), or 3D printing, of metals is transforming the fabrication of components, in part by dramatically expanding the design space, allowing optimization of shape and topology. However, although the physical processes involved in AM are similar to those of welding, a field with decades of experimental, modeling, simulation, and characterization experience, qualification of AM parts remains a challenge. The availability of exascale computational systems, particularly when combined with data-driven approaches such as machine learning, enables topology and shape optimization as well as accelerated qualification by providing process-aware, locally accurate microstructure and mechanical property models. We describe the physics components comprising the Exascale Additive Manufacturing simulation environment and report progress using highly resolved melt pool simulations to inform part-scale finite element thermomechanics simulations, drive microstructure evolution, and determine constitutive mechanical property relationships based on those microstructures using polycrystal plasticity. We report on implementation of these components for exascale computing architectures, as well as the multi-stage simulation workflow that provides a unique high-fidelity model of process–structure–property relationships for AM parts. In addition, we discuss verification and validation through collaboration with efforts such as AM-Bench, a set of benchmark test problems under development by a team led by the National Institute of Standards and Technology.

3D printing↗
DOE OSTI2020

Modeling the role of local crystallographic correlations in microstructures of Ti-6Al-4V using a correlated structure visco-plastic self-consistent polycrystal plasticity formulation

This paper presents a multi-level crystal plasticity-based simulation framework for modeling mechanical response and microstructure evolution of Ti-6Al-4V with α-lath/lamellar microstructures. Here, the model is a correlated structure visco-plastic self-consistent (CS-VPSC) formulation linking three scales: a single crystals micro-scale, a lath/lamellar colony meso-scale, and a lath/lamellar aggregate macro-scale. A selected hardening law for the evolution of critical resolved shear stress per slip system used in CS-VPSC is phenomenological. However, it adjusts the resistances of basal and prismatic slip systems based on the geometry of slip transfer between adjacent lamellae. Consistent with experimental evidences, the resolved shear stress on the pyramidal slip planes is dependent not only on the stress in the direction of slip but also on the two orthogonal shear stress components and the three normal stress components (non-Schmid effects). Electron backscatter diffraction (EBSD) data in conjunction with a procedure relying on α→β phase transformation is used to construct paired variants of α-laths/lamellae satisfying their local crystallographic correlations. The procedure fits volume fractions of individual laths/lamellae with the experimental EBSD data and selects a distribution of habit planes between adjacent variants with respect to the loading direction. The simulation framework is applied to interpret the deformation behavior in tension and compression along two sample directions of Ti-6Al-4V fabricated via laser powder bed fusion. Moreover, the model is used to simulate texture evolution during rolling of the material to large strains. It is demonstrated that the model is capable of predicting plastic anisotropy/asymmetry and the concomitant texture evolution. While the model reveals a significant effect of habit plane inclination with respect to the loading direction on yield stress, the comparison of the data and model predictions shows that a random distribution of habit planes fits the flow response. It is further inferred that the tension-compression asymmetry arises from the non-Schmid effects.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Effect of wheel speed and annealing temperature on microstructure and texture evolution of Ni{sub 45}Mn{sub 36.6}In{sub 13.4}Co{sub 5} ribbon

Ni{sub 45}Mn{sub 36.6}In{sub 13.4}Co{sub 5} magnetic shape memory alloy was successfully produced as preferentially textured ribbon by melting spinning with different wheel speed. X-ray diffraction (XRD) and electron back scatter diffraction (EBSD) were used to study structure and texture evolution of these melt-spun ribbons. The thickness of melt-spun ribbon is 42 μm, 65 μm and 30 μm depending on wheel speed of 1 0 m/s, 15 m/s and 20 m/s, respectively. Density of α fiber texture (〈100〉//ND) vary with wheel speed changes, and is most intensive in the ribbon with wheel speed of 15 m/s. Grains of the ribbons grow after being annealed at 873 K, 973 K, 1073 K and 1173 K, recrystallization was not observed in ribbons after being annealed at 873 K but occurred in ribbons after being annealed at higher temperatures. The α fiber texture becomes weaker to some extent after annealing at different temperatures, due to new recrystallization texture formed at the process of annealing. - Highlights: •Sectional part of shape memory ribbon is firstly investigated by EBSD method. •Thickness and texture of ribbons vary with wheel speed. •Annealing temperature affect texture and microstructure evolution greatly. •Recrystallization textures were observed in ribbons after being annealed.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Mobility assessment of the BCC and carbide phases in the C-Nb, C-U and Nb-U systems

Uranium carbides with refractory metal additions are considered for Gen IV nuclear reactors and nuclear thermal propulsion as fuels for their high-temperature and corrosion resistant properties. Understanding kinetic effects that dictate microstructural evolution during fabrication and operating conditions is essential to advance technological development of these fuels. This work presents the development of an atomic mobility database for C-Nb-U systems based off available experimental data supported with ab-initio methods. The mobility assessments and uncertainty quantification (using Markov chain Monte Carlo) were conducted in the Kawin software. Carbon diffusion is considered dominant, as metal diffusion is much slower, with niobium diffusion being even slower and rate limiting than uranium metal. We provide a comprehensive and self-consistent thermo-kinetic database that is validated by diffusion couple simulations through Kawin. In conclusion, this enables prediction of microstructural and phase evolution critical for the development and lifetime assessment of next generation nuclear fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Excellent high cyclic fatigue properties of a novel ultrafine-grained medium entropy alloy

Fatigue is a major cause of failure and requires dedicated investigation to validate the integrity of potential structural materials. As high and medium entropy alloys (HEAs and MEAs) are garnering significant attention from scientific communities, it is important to test the structural integrity of these novel alloys in cyclic loading environment. This study analyzed both monotonic and cyclic loading properties of a novel medium entropy alloy (MEA) CoCr 1.3 Ni in the ultrafine-grained (UFG) and fine grain (FG) conditions. The UFG MEA performs better than the FG MEA during cyclic loading. Microstructural evolution after cyclic loading via electron back-scattered diffraction shows microstructural stability of both FG and UFG microstructures. Transmission electron microscopy suggests formation of stacking faults and fine-scale HCP transformation during cyclic loading.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Predicting structural material degradation in advanced nuclear reactors with ion irradiation

Swelling associated with the formation and growth of cavities is among the most damaging of radiation-induced degradation modes for structural materials in advanced nuclear reactor concepts. Ion irradiation has emerged as the only practical option to rapidly assess swelling in candidate materials. For decades, researchers have tried to simulate the harsh environment in a nuclear reactor in the laboratory at an accelerated rate. Here we present the first case in which swelling in a candidate alloy irradiated ~ 2 years in a nuclear reactor was replicated using dual ion irradiation in ~ 1 day with precise control over damage rate, helium injection rate, and temperature and utilize physical models to predict the effects of radiation in reactors. The capability to predict and replicate the complex processes surrounding cavity nucleation and growth across many decades of radiation dose rate highlights the potential of accelerated radiation damage experiments. More importantly, it demonstrates the capability to predict the swelling evolution and the possibility to predict other features of the irradiated microstructure evolution that control material property degradation required to accelerate the development of new, radiation-tolerant materials.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

MARMOT Modeling of Microstructure-Dependent Thermal Conductivity in Nuclear Fuels

Uranium dioxide (UO{sub 2}) is the primary nuclear fuel in light water reactors. The thermal conductivity of UO{sub 2} is an important fuel performance metric as it affects nuclear energy conversion efficiency and safety. Unfortunately UO{sub 2} has a very low thermal conductivity, even in the single crystal form. In reactor environment, the radiation-induced defects, fission products, and complex microstructural evolution degrade the thermal conductivity of UO{sub 2} significantly. These defects and microstructures in UO{sub 2} fuels include point defects, gas bubbles, metallic precipitates, etc.. In the rim region of a fuel pellet, high burnup structure also forms which contains many nanograins and thus high density of grain boundaries. Therefore, understanding how different irradiation-induced microstructures affect the thermal conductivity in nuclear fuels is important for predicting the overall thermal transport properties in nuclear fuels. In this work, the unit effect of each microstructure on the thermal conductivity of UO{sub 2} is evaluated, using grain boundary as an example. The obtained unit mechanism can be used as input for the multiplier model that describes the combined effect of various microstructures. In this work we use phase field simulations coupled with heat conduction calculations to study the co-evolution of grain size and thermal conductivity. The simulation data are used to calibrate an analytical model. Based on the sufficient statistics from simulations, we conclude that the analytical model can well predict the grain-size-dependent thermal conductivity over a wide range of grain sizes if the input parameters are calibrated. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Quantitative investigation into the influence of temperature on carbide and austenite evolution during partitioning of a quenched and partitioned steel

Here, the influence of partitioning temperature on microstructural evolution during quenching and partitioning was investigated in a 0.38C-1.54Mn-1.48Si wt.% steel using Mössbauer spectroscopy and transmission electron microscopy. η-carbide formation occurs in the martensite during the quenching, holding, and partitioning steps. More effective carbon partitioning from martensite to austenite was observed at 450 than 400°C, resulting in lower martensite carbon contents, less carbide formation, and greater retained austenite amounts for short partitioning times. Conversely, greater austenite decomposition occurs at 450°C for longer partitioning times. Lastly, cementite forms during austenite decomposition and in the martensite for longer partitioning times at 450°C.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Evolution of B2 and laves phases in a ferritic steel under Fe 2+ ion irradiation at 475 °C

Microstructural evolution in a novel ferritic steel (Fe–12Cr–3W–3Ni–3Al–1Nb, in wt.%) computationally designed to contain B2 and Laves phases after 4 MeV Fe 2+ ion irradiation up to 220 dpa at 475 °C was characterized using transmission electron microscopy in conjunction with x-ray energy dispersive spectroscopy. The ferritic matrix phase exhibited dislocation loops and tangled dislocations, but our focus was on stability of two types of intermetallic precipitates. The B2–NiAl precipitates ~13 nm in size remained crystalline and appeared to have slightly lower Al concentration after irradiation. The Laves phase, (Fe,Cr) 2 (Nb,W), were present in two size ranges: coarse micron-scale precipitates which were amorphized with a slight composition change at irradiation doses above~30 dpa, while the finer precipitate particles~100 nm in size were partially disintegrated with a noticeable composition change at doses above~70 dpa. Meanwhile, many Nb/Cr-enriched particles ~8 nm in size formed within a few hundreds of nanometers from the disintegrated particles. The understanding of the phase stability would help design advanced steels and engineer microstructures that are stable against high irradiation doses, while retaining good high temperature strength.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Identifying multiple synergistic factors on the susceptibility to stress relaxation cracking in variously heat-treated weldments

The 347H austenitic stainless steel has been widely used for pressure vessels and pipeline (PVP) applications due to its excellent creep and corrosion resistance, which fit ideally to the harsh conditions in petrochemical industries, fossil fuel or nuclear power plants, and modern energy storages. However, a failure mode has been commonly observed with cracks emerging at the heat affected zone (HAZ) of weldments during post-weld heat treatment (PWHT) or under intermediate to high temperature service conditions. This phenomenon is termed as Stress Relaxation Cracking (SRC) since the purpose of PWHT is to relieve the welding-induced residual stress fields, or as Stress Age Cracking (SAC) if failure happens during service. A leading literature explanation of this failure suggests that the residual stress relaxation and the precipitation dissolution and/or re-precipitation occur in the same temperature range, which can lead to locally high strains and thus to crack at the grain boundaries. Since in situ spatial measurements of residual stress fields, microstructural evolution, and failure processes are nearly infeasible, this work recourses to a micromechanical finite element framework that models the high temperature failure as the nucleation and growth of grain boundary cavities, whereas various parameters such as thermomechanical loading history and its evolution, the competition of grain-interior dislocation creep and grain-boundary diffusion in failure lifetime, and microstructural heterogeneities (such as the precipitate free zone near grain boundaries) can be quantitatively incorporated. It can be concluded from these microstructure-explicit simulations that an accurate knowledge of residual stress evolution and a carefully calibrated set of material constitutive parameters are the essential prerequisites for lifetime predictions. The understanding of individual governing factors also leads to a mechanistic interpretation of the observed SRC susceptibility C-curves. In conclusion, these results suggest that the criticality of residual stress evolution, but not the precipitation-induced local strains, be the leading factor for SRC.

347H stainless steel weldments↗
DOE OSTI2018

Lattice distortion in a strong and ductile refractory high-entropy alloy

Here, the maximization of the mixing entropy with the optimal range of enthalpy in high-entropy alloys (HEAs) can promote the formation of a stable single solid-solution phase with the absence of competing intermetallic compounds. The resultant effects, such as lattice distortion, can contribute to excellent mechanical properties, which has motivated numerous efforts to develop and design single-phase HEAs. However, challenges still remain, particularly on quantifying the lattice distortion and relating it to materials properties. In this study, we have developed a NbTaTiV refractory HEA with a single body-centered-cubic (BCC) structure using an integrated experimental and theoretical approach. The theoretical efforts include thermodynamic modeling, i.e., CALculation of PHAse Diagram (CALPHAD). The microstructural evolutions have been investigated by systematic heat-treatment processes. The typical dendrite microstructure was observed, which is caused by the elemental segregation during the solidification in the as-cast condition. The structural inhomogeneity and chemical segregation were completely eliminated by the proper homogenization treatment at 1200 °C for 3 days. The homogeneous elemental distribution was quantitatively verified by the Atom Probe Tomography (APT) technique. Importantly, results indicate that this HEA exhibits the high yield strength and ductility at both room and high temperatures (up to 900 °C). Furthermore, the effects of the high mixing entropy on the mechanical properties are discussed and quantified in terms of lattice distortions and interatomic interactions of the NbTaTiV HEA via first-principles calculations. It is found that the local severe lattice distortions are induced, due to the atomic interactions and atomic-size mismatch in the homogenization-treated NbTaTiV refractory HEA.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Consumable development to tailor residual stress in parts fabricated using directed energy deposition processes

Distortion and residual stresses are major challenges that limit the ability to fabricate large scale structures using Additive Manufacturing (AM). Researchers worldwide are evaluating techniques to induce compressive residual stress in the parts via intermittent rolling. While reasonable success has been documented, the idea of lowering the martensite start temperature to induce compressive stresses has not been evaluated in the context of AM, despite demonstrated success by the welding community. This study validates the hypothesis that, by a proper selection of materials and process parameters, one may effectively reduce distortion and induce a compressive residual stress in AM parts. Using neutron diffraction to measure residual stresses in parts, we demonstrate that, in addition to selection of the correct materials, the inter pass temperature plays a major role in controlling the residual stress evolution. The observations relating to the residual stresses are rationalized based on a microstructural evolution in these samples. Based on this preliminary study, a strategy to fabricate large structures with minimal distortion and residual stress is outlined.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Characterization of a minor actinides bearing metallic fuel pin irradiated in EBR-II

The X501 transmutation experiment irradiated in Experimental Breeder Reactor II (EBR-II) was intended to evaluate the safety and performance of the addition of minor actinides (Am, Np) into the metallic nuclear fuel system. The irradiation of this experiment is also of paramount importance for comparisons between true fast spectrum irradiations (e.g. EBR-II) and shrouded irradiation capsules performed at the thermal neutron spectrum Idaho National Laboratory Advanced Test Reactor. The X501 experimental fuel pins were irradiated to a peak measured burnup of 6.2% fission per initial heavy metal atom. This paper reports and discusses Postirradiation Examination results for one of two minor actinide bearing pins from the EBR-II X-501 experiment (X501-G591) with a focus on fuel microstructure evolution. Additionally, fuel swelling, fission product distribution, cladding strain, fission gas release, fuel microstructure and fuel cladding chemical interaction were all evaluated. The fission gas release was 79.9% and the microstructure along the fuel pin presented several characteristics very similar to other EBR-II U-Pu-Zr ternary metallic fuel experience. Minor actinides seem to not dramatically affect the overall performance of this candidate transmutation fuel. Scanning Electron Microscopy was also performed and an in-depth characterization of the different regions and phases is presented. Performance data from these irradiations can be used to inform the feasibility of minor actinide transmutation in future reactor systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗