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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Unveiling Atomistic Mechanisms Governing Additive Manufacturing Processability and Mechanical Behavior of a Refractory Complex Concentrated Alloy

Extending the concept of complex concentrated alloys (CCAs) to the refractory alloys (solidus temperature over 2000 °C) space potentially facilitates the design of lightweight structural alloys with service temperatures that exceed those of Ni and Co‐based alloys. However, the room and elevated temperature tensile properties of the current refractory‐CCAs (R‐CCAs) are inferior to those of the Ni/Co‐based alloys. Furthermore, the manufacturing scalability of R‐CCAs remains challenging, in that cracks are prevalent in all R‐CCAs when processed using near‐net shape manufacturing processes, such as fusion‐based additive manufacturing (F‐BAM). Still, mechanisms governing the poor F‐BAM processability of R‐CCAs remain unexplored. Here, to this end, this work unveils the atomistic mechanisms underlying F‐BAM process‐induced cracking in a NbTiTaMoHfZrC R‐CCA. The implications of light elements’ presence for intrinsic ductility and grain boundary cohesion, and subsequently for F‐BAM processability and mechanical behavior, are revealed. Leveraging the insights, we accomplish what is, to the best of the knowledge, the first instance of crack‐free F‐BAM processing of any R‐CCA. Additionally, the R‐CCA exhibits over 20% tensile ductility and ≈160 MPa tensile yield strength at 1200 °C. In addition to facilitating the design of lightweight R‐CCAs, findings enable scalable manufacturing of these ultra‐high temperature alloys for structural applications.

Refractory alloys↗
DOE OSTI2021

Critical cooling rates for amorphous-to-ordered complexion transitions in Cu-rich nanocrystalline alloys

Amorphous complexions in nanocrystalline metals have the potential to improve mechanical properties and radiation tolerance, as well as resistance to grain growth. In this study, the stability of amorphous complexions in binary and ternary Cu-based alloys is investigated by observing the effect of cooling rate from high temperature on the occurrence of amorphous-to-ordered complexion transitions. Bulk Cu-Zr and Cu-Zr-Hf alloy samples were annealed to induce boundary premelting and then quenched through a procedure that induces a gradient of local cooling rate through the sample height. Amorphous complexion thickness distributions were found to be invariant to local cooling rate in the Cu-Zr-Hf alloy, demonstrating enhanced stability of the amorphous complexion structure compared to the Cu-Zr alloy, which had thinner amorphous complexions in the regions that were slowly cooled. The experimental results are used to construct time-temperature-transformation diagrams for the amorphous-to-ordered complexion transition in both the binary and ternary alloys, enabling a deeper understanding of the influence of cooling rate and grain boundary chemistry on complexion transitions. In conclusion, the critical cooling rate necessary to avoid complexion transitions in the ternary alloy is found to be at least three orders of magnitude slower than that for the binary alloy.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Electrochemical grand potential-based phase-field simulation of electric field-assisted sintering

Here, an electrochemical grand potential functional was proposed to describe the sintering of an ionic ceramic green body. The resultant phase-field description enables simulation of the consolidation of an arbitrary number of granular particles and their interactions with the surrounding void phase. The model includes the effects of charged vacancies and the associated interactions between internal and applied electric fields. Defect segregation to grain boundaries is also accounted for, as well as enhanced interfacial defect mobilities. The model was parameterized for Y 2 O 3 . Simulations of two-particle systems showed that the applied electric field had an increasingly important impact on neck growth as particle size increased. A sudden rapid increase in temperature occurred for larger field strengths, which has been reported to be correlated to the onset of a flash event in flash sintering. Simulations of many particles showed that internal heat generation by Joule heating was localized at particle–particle contacts (grain boundaries), even though their conductivities were lower than nearby internal particle-void interfaces. A percolative path for ionic charge across the green body and the ceramic sintered solid was thus defined, accelerating the Joule heating process as the porosity of the green body is removed.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Unveiling the nucleation and growth of Zr oxide precipitates in internally oxidized Nb3Sn superconductors

We report on atomic-scale analyses of nucleation and growth of Zr oxide precipitates and the microstructural evolution of internally oxidized Nb3Sn wires for high-field superconducting magnet applications, utilizing atom probe tomography (APT), transmission electron microscopy (TEM), and first-principles calculations. APT analyses reveal that oxygen and zirconium are already segregated at grain boundaries (GBs) in the unreacted Nb-1Zr-4Ta (at%) alloy prior to forming Nb3Sn through reacting the Nb alloy with Sn and SnO2. After forming Nb3Sn, Zr oxide precipitates nucleate both at the Nb3Sn/Nb heterophase interfaces and in the Nb3Sn grains, driven by the small solubilities of Zr and O in Nb3Sn compared to their value in Nb. A high number density (Nv) of Zr oxide nanoprecipitates is observed in the Nb3Sn layers, ∼10 23 m −3 , with a mean diameter <10 nm for a heat treatment at 625 °C. Quantitative APT and TEM analyses of the Zr oxide precipitates in the reacted Nb3Sn layers elucidate details of the nucleation, growth, and coarsening processes of the Zr oxide precipitates in Nb3Sn. First-principles calculations and classical nucleation theory are employed to study the nucleation of Zr oxide precipitates in Nb3Sn and to estimate the maximum energy barrier and critical radius for nucleation. Our research unveils the kinetic pathways for nucleation and growth of Zr oxide precipitates and the microstructural evolution of Nb3Sn layers, which helps to understand and improve the superconducting properties of internally oxidized Nb3Sn wires for use in high-field superconducting magnets. •We investigated the nucleation and growth of Zr oxide nanoprecipitates in Nb3Sn superconductors for magnet applications.•APT and TEM analyses reveals the nucleation and growth process of Zr oxide precipitates in Nb3Sn layers.•Classical nucleation theory with DFT is employed to describe the nucleation process of Zr oxide precipitates in Nb3Sn.•We find that the interfacial reactions at Nb/Nb3Sn interface play a critical role in the nucleation of Zr oxide precipitates.•Current study provides a pathway to improve the properties of internally oxidized Nb3Sn superconducting wires.

43 PARTICLE ACCELERATORS↗
DOE OSTI2021

Atom Probe Tomography Characterization of High-Dose Ion Irradiated MA957

This paper presents a comprehensive characterization study of the microstructural responses to ion irradiation of the oxide dispersion strengthened (ODS) ferritic alloy MA957 using atom probe tomography (APT). The MA957 specimens were irradiated by 1.8 MeV Cr3+ ions to doses and temperatures relevant to advanced fast reactors (100 and 500 dpa at 400-500 °C). The average sizes of the YTiO particles decreased slightly after irradiation at all temperatures, with the most pronounced change at 400 °C. The particle number densities for 400 and 500 °C were slightly below number densities for the unirradiated condition, while the values were slightly higher at 420 and 450 °C. This behavior is consistent with the Nelson-Hudson-Mazey rate theory model that considers the competing effects between recombination and dissolution. An increase in the Y/Ti ratio of the YTiO particles was observed at the two lower temperatures. The formation of Cr-rich precipitates (a') was observed in specimens irradiated at temperatures up to 450 °C. Irradiation-induced segregation on grain boundaries was also investigated. Grain boundary element concentration profiles and ion maps demonstrate differences among the irradiation temperatures, with most datasets differing from the unirradiated condition. A key focus of this ion irradiation study is the high-dose response, and this quantitative assessment revealed relatively small but distinct effects of irradiation on microstructure evolution. Overall, these results showing minor second phase evolution suggest that MA957 should retain mechanical performance generally well up to 500 dpa for the irradiation temperatures studied.

Wang, Jing↗
DOE OSTI2020

Formation of multicomponent alloy particles in doped ceria under I 2+ ion irradiation and thermal annealing

Multicomponent alloys resembling the e-phase particles in nuclear fuel have the potential to immobilize 99 Tc in a stable waste form. Formation of such multicomponent alloys under extreme conditions is studied in a non-activated model system: CeO 2 doped with 2 wt% Mo, 1.5 wt% Ru, 0.75 wt% Pd, 0.5 wt% Re and 0.25 wt% Rh. The doped CeO 2 films were irradiated with I 2+ ions (610°C, 10 16 and 5×10 16 I 2+ /cm 2 ). For selected samples post-irradiation heat treatment was conducted (900°C, 1100°C). Analytical TEM revealed the formation of 5-20 nm precipitates containing Pd, Mo, and Re sometimes associated with cavities. Iodine was found to segregate to grain boundaries and cracks. After heat treatment at 1100°C, the CeO 2 matrix had recrystallized. Precipitates containing Pd, Mo, Re, and Ru were observed near the interface with the polycrystalline YSZ substrate. It follows that Pd is the most mobile element, followed by Mo and Re. Finally, while irradiation promotes precipitation, high-temperature effects are more significant.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Creep properties of advanced austenitic steel 709 determined through short experiments under in-situ neutron diffraction followed by TEM characterization

Here in this study, the creep mechanisms at play in Alloy 709 (Fe-20Cr-25Ni) are investigated by performing short-term creep-type tests under in-situ neutron diffraction experiments. Short tests are performed in the temperature range of 500 to 900 °C under constant load with a load ranging from 50 to 150 MPa. The creep exponent and activation energy are determined using the Bird-Mukherjee-Dorn relation and compared to that obtained from conventional longer creep tests from the literature. Scanning transmission electron microscopy (S/TEM) of the post-creep microstructure indicates that interaction of dislocations with precipitates are a dominant mechanism at play. Furthermore, local elemental mapping indicated chemical segregation at grain boundaries and formation of complex precipitates.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Distribution of Copper States, Phases, and Defects across the Depth of a Cu-Doped CdTe Solar Cell

Copper has been used as a p-type dopant in cadmium telluride (CdTe) for decades. However, the density of Cu atoms in the finished device is much higher than that of holes, which means that most Cu atoms are not activated as acceptors during incorporation. Furthermore, studies have demonstrated that the distribution of copper (Cu) atoms across the device is highly inhomogeneous, with reports citing Cu substitution on Cd sites and segregation to grain boundaries. Fast diffusion along these boundaries and Cu accumulation at the CdTe/CdS interface have also been observed and validated computationally. These levels of inhomogeneity make it difficult to accurately characterize and correlate the performance with the nature of the Cu atomic species present. To address this challenge, we utilize X-ray microscopy and, specifically, nanoscale fluorescence-mode X-ray absorption near-edge structure to resolve the atomic Cu environment throughout the depth of the CdTe layer. Our results suggest that the majority of Cu atoms are in the form of Cu x Te phases (or similar local environments) near the ZnTe|CdTe interface, Cu x O phases in the CdTe absorber, and present in various oxidation states, including Cu 1+ and Cu 2+ , near the CdS/CdTe junction. Here this work also provides experimental evidence for the first time of the presence of CuS around the ZnTe|CdTe interface and the hypothesized Cu Cd -Cl i complex in the CdTe absorber.

14 SOLAR ENERGY↗
DOE OSTI2020

Comparative study of helium bubbles in a Ti-Ta alloy and a Ti/Ta nanocomposite

Here, we investigated the size, density, and distribution of helium (He) bubbles in a Ti-Ta single-phase alloy and a Ti/Ta dual-phase nanocomposite using transmission electron microscopy. The Ti/Ta nanocomposite was fabricated via phase separation during high-temperature annealing of the single-phase Ti-Ta alloy. He ion implantation in the Ti-Ta single-phase alloy leads to the formation of nano-scale He bubbles (~1.7 nm in size, ~2.6×10 5 /μm 3 in volumetric density). He bubbles were found to segregate to grain boundaries. Under identical implantation conditions, numerous He bubbles also formed in the Ti/Ta nanocomposite. Closer inspection revealed that He bubbles in the Ti-rich phase are smaller and of higher volumetric density (~1.2 nm, ~8.4×10 5 /μm 3 ) than those in the Tarich phase (~3 nm, ~1.8×10 5 /μm 3 ) and the single-phase alloy. He bubbles were also observed to decorate phase boundaries, but they were generally smaller than those on grain boundaries. These results suggest that Ti/Ta phase boundaries are less conducive to the growth of large He bubbles than grain boundaries in the Ti-Ta single phase alloy.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Origin of micrometer-scale dislocation motion during hydrogen desorption

Hydrogen, while being a potential energy solution, creates arguably the most important embrittlement problem in high-strength metals. However, the underlying hydrogen-defect interactions leading to embrittlement are challenging to unravel. Here, we investigate an intriguing hydrogen effect to shed more light on these interactions. By designing an in situ electron channeling contrast imaging experiment of samples under no external stresses, we show that dislocations (atomic-scale line defects) can move distances reaching 1.5 μm during hydrogen desorption. Combining molecular dynamics and grand canonical Monte Carlo simulations, we reveal that grain boundary hydrogen segregation can cause the required long-range resolved shear stresses, as well as short-range atomic stress fluctuations. Thus, such segregation effects should be considered widely in hydrogen research.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Intermediate-Term Thermal Aging Effect Evaluation for Grade 92 and 316L at The LWR Relevant Temperature

Life extension of the existing nuclear reactors imposes accumulated damages, such as higher fluences and longer period of corrosion, to structural materials, which would result in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) on the Advanced Radiation Resistant Materials (ARRM) program, aiming to develop and test degradation resistant alloys from current commercial alloy specifications by 2021 to a new advanced alloy with superior degradation resistance in light water reactor (LWR)-relevant environments by 2024. Based on a comprehensive microstructure and property screening performed in Phase-1 of the ARRM program, a total of five alloys, together with 316L and X-750 as references, were down-selected for Phase-2 neutron irradiation studies. Because thermal aging could exert a synergistic effect on neutron irradiation due to the low neutron damage rate on the order of 10 –7 displacements per atom per second (dpa/s), Grade 92 (one of the five down-selected alloys) and two heats of 316L were selected in this task to study the effect of aging at 350°C for 12.5–12.7 kh on microstructure and mechanical properties. The aging time is approximately corresponding to 5 dpa neutron irradiation, which can be served as a reference for the 5-dpa-irradiated samples of the alloys to help understand the independent neutron damage influence on microstructure and mechanical properties. Optical microscopy and scanning electron microscopy were used for microstructural characterization. Hardness, tensile, Charpy impact toughness, and fracture toughness in the ductile regime of the aged samples were examined. The aging of Grade 92 led to the formation of many Laves phase in sizes of ~100–200 nm and resulted in some reduction in hardness and yield/ultimate tensile strength with some increases in uniform and total plastic elongations, which may have helped the enhancement of Charpy impact toughness, e.g., ~4 J increase in upper-shelf energy and 20.5°C reduction in ductile-brittle transition temperature compared with the unaged condition. The fracture toughness of the aged Grade 92 showed decent toughness of ~303 MPa√m (K Jq ) with ~87 tearing modulus at 22°C, which reduced to ~241 MPa√m (~20% reduction) with ~75 tearing modulus at 300°C. The two heats of 316L are differentiated by their amounts of δ-ferrite, i.e., ~1 vol% in heat T1103 and ~4 vol% in heat N5B8. The two heats of 316L were subjected to 15% cold work (CW) as this condition is often used in nuclear reactors to help trapping radiation-induced defects and thus postpone the steady state swelling stage. The aging did not result in any noticeable microstructure changes, except for possible segregations at grain boundaries that need to be further investigated. The aging resulted in slight reduction in hardness of 316L-T1103 and slight increase in hardness of 316L-N5B8, which is consistent with their tensile testing results. The 15%CW might have introduced some inhomogeneity, resulting in large standard deviations in hardness, which were reduced after the aging. The aged 316L followed the same trend as the unaged condition for the strength and elongation results from the tensile testing. Minimum elongations appeared within ~300–600°C, above which the elongations seem to increase. Unlike the negligible or minor changes in microstructure, hardness and tensile results of the two heats of 316L, their fracture toughness showed noticeable difference. The aged 316L-T1103 showed good toughness of ~350 MPa√m with ~80 tearing modulus at 22°C, which reduced to ~271 MPa√m (~23% reduction) with ~83 tearing modulus at 300°C. In contrast, the aged 316L-N5B8 showed lower toughness of ~244 MPa√m with ~53 tearing modulus at 22°C, which reduced to ~180 MPa√m (~26% reduction) with ~31 tearing modulus at 300°C. The preliminary results indicate that the presence of high volume of δ-ferrite would noticeable impair fracture toughness although it may not influence hardness and tensile properties.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Copper Defects and Charge Collection in Cadmium Telluride Photovoltaic Devices

In CdTe photovoltaic devices, the CdTe layer must be doped with Cu to reduce its resistivity, increase carrier lifetime, and improve hole transport. However, the Cu concentration is typically very low around the interfaces and within the CdTe layer, making it difficult to determine the influence Cu has on charge transport at the nanoscale; Cu concentrations are difficult to detect due to low ionization probability of Cu atoms [1]. Here, we use synchrotron X-ray microscopy to probe the nanoscale distribution of Cu and correlate it to local charge collection in CdTe photovoltaic devices. We demonstrate Cu segregation around grain boundaries, and, using cross-section charge collection measurements and transport modelling, show recombination center concentration in the CdTe layer dictates the interface at which charge collection occurs. The work gives insights on how Cu distributes in CdTe photovoltaic systems, and an understanding of how these distributions affect charge transport.

Walker, Trumann↗
DOE OSTI2025

Grain boundary energy control in zinc aluminate nanoceramics

This study investigates the grain boundary energy dependence on segregated dopants in nanocrystalline zinc aluminate ceramics. Atomistic simulations of Σ3 and Σ9 grain boundaries showed that trivalent ions of varying ionic radii [Sc 3+ (74.5 pm), In 3+ (80.0 pm), Y 3+ (90.0 pm), and Nd 3+ (98.3 pm)] have a tendency to segregate to both interfaces, with Y 3+ presenting the highest segregation potentials. The connection between segregation and the reduction of interfacial energies was explored by measuring the grain boundary energy on nanoceramics fabricated via high‐pressure spark plasma sintering (HP‐SPS) using differential scanning calorimetry (DSC). The results revealed that Y 3+ doping at 0.5 mol% reduces the grain boundary energy in zinc aluminate nanoceramics from 1.1–1.3 J/m 2 to 0.6–0.8 J/m 2 ; the range correlates with the observed size dependence of the excess energy, with higher values observed for the smaller grain sizes (∼17 nm). The noted decrease in interfacial energies for doped samples suggests it is indeed possible to alter the stability of zinc aluminate grain boundaries via dopant segregation.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Triple junction solute segregation in Al-based polycrystals

Solute segregation is a crucial means of stabilizing nanostructured alloys, and at very small grain sizes, this requires consideration of triple junctions (TJs), which attain a meaningful volume fraction and thus become relevant for bulk material behavior. Here, in this study, the solute segregation spectra for grain boundary (GB) and TJ sites are calculated for a large number of dilute Al-based binary systems with available interatomic potentials. A defect-identifying algorithm is applied to quantify the average GB thickness and classify the intergranular site spectra into GB and junction subspectra. The algorithm is also applied to a hybrid electronic-level database for GB segregation of various solutes in Al, yielding polycrystalline TJ solute segregation spectra from first principles. The results suggest that TJ segregation is alloy or interatomic potential dependent and can exhibit either boundary or junction preference. With these spectra as inputs, the spectral GB segregation model gives quantitative predictions of segregation as a function of grain size, temperature, and total solute concentration, suitable for alloy screening and design.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2025

Understanding the effect of minor alloying elements on helium bubble formation in ferritic-martensitic steels

Ferritic-martensitic steels are promising structural materials for advanced nuclear reactors. To minimize long-term radioactivity, reduced-activation ferritic-martensitic steels have been developed by substituting high-activation elements like Ni and Mo with low-activation elements such as W. However, the impact of these alloying modifications on helium bubble formation, which plays a key role in material swelling, remains unclear. Here, in this study, we compared helium bubble formation in ferritic-martensitic steel T91 and reduced-activation ferritic-martensitic steel F82H. Both materials were irradiated with sequential 100 keV, 150 keV, and 200 keV helium ions to a dose of 0.5 dpa and a helium concentration of 9,000 appm at 500°C. The helium bubbles in F82H exhibited a larger average size and a lower density than those in T91, suggesting differences in minor alloying elements may influence the bubble growth. Here, to investigate the effects of these alloying elements, we characterized radiation-induced segregation near bubbles and grain boundaries. Prominent Ni-Mn-Si enriched clusters were found near bubbles in T91, while only Mn-Si enriched clusters were found near bubbles in F82H. In addition, the obvious Cr enrichment near grain boundaries was absent around bubbles in both steels. The different segregation trends among elements revealed the variations in element diffusion mechanisms and the different sink biases between bubbles and grain boundaries. Cr enrichment near grain boundaries is mostly driven by interstitial-mediated diffusion. However, since bubble growth relies on net vacancy flux, vacancy-mediated diffusion plays a dominant role in controlling element segregation near bubbles. Therefore, Cr enrichment was not found near bubbles. Because of preferential vacancy-drag diffusion for Ni, Si and Mn, these elements were enriched near bubbles. Due to the strong binding energies of vacancies with these solute atoms, the vacancy diffusivity can be reduced near these solutes. Therefore, the more prominent Ni-Si-Mn clustered near helium bubbles in T91 lead to stronger suppression of helium bubble growth compared to F82H.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Examination Of IASCC In 304 Stainless Steel Core Shroud From A Commercial Boiling Water Reactor (BWR)

The 304 stainless steel core shroud and the accompanying weld material from a commercial boiling-water reactor (BWR) experienced an estimated dose of ~3.5 dpa neutron irradiation leading to irradiation-assisted stress corrosion cracking (IASCC). The resulting IASCC crack tips are examined with high-resolution analysis techniques, including scanning electron microscopy (SEM), scanning transmission electron microscopy (STEM), and atom probe tomography (APT), with select examples of directly correlated STEM/APT observations. Analytical observations reveal the crack morphology, composition, microstructure, and any associated corrosion-induced or radiation-induced segregation in the local metal grain boundaries ahead of the propagating SCC cracks. The observed oxide composition and structure were in-line with expectations, consisting of a Cr-rich inner oxide and a Fe-rich outer oxide, with the leading crack tip oxide being Cr-rich. In addition, the leading grain boundary exhibited typical radiation-induced segregation signatures that were largely unperturbed by the accompanying corrosion. Finally, the quantitative capabilities and limitations of STEM and APT in the context of corrosion fronts and grain boundary chemistry from the directly correlated measurements are discussed.

Olszta, Matthew J.↗