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El-Atwani, Osman

Publications and source records attributed to El-Atwani, Osman.

17 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Down-selection of Innovative Fusion Materials

The goal of this proposal is to develop and fundamentally understand the microstructure of refractory multi-component alloys (MCA) as fusion-relevant plasma facing and structural materials, a step to down-select innovative fusion materials prior to the process of advancing their technology readiness level (TRL). Developing materials is paramount to enable fusion as energy source since no existing material is capable of coping with such extreme conditions. We will aim at understanding the role of chemistry and microstructure in these complex multi-component alloys systematically comparing their performance with pure tungsten materials in terms of mechanical properties and manufacturability. Such understanding will open opportunities to optimize MCAs for fusion applications and has a potential of attracting industry partnership.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Synthesis, microstructure and micro-mechanical characterization of metal (Nb, Ti) – MAX phase (Ti 2 AlC) nanolaminates

We utilize elevated temperature physical vapor deposition (PVD) techniques to design metal/MAX multilayered nanocomposite thin films with alternating nanoscale metallic (Nb, Ti) and MAX phase (Ti 2 AlC) layer thicknesses. These metal/MAX nanolaminate architectures attempt to exploit a unique hierarchical topology – as interfaces between the layers are expected to be in direct competition with the internal interfaces within the MAX layers, to drive their tunable macroscopic mechanical behavior. Two metal/MAX nanolaminates – Nb/Ti 2 AlC and Ti/Ti 2 AlC – were deposited. The Nb/Ti 2 AlC metal/MAX system showed highly diffused layer interfaces with distinct Ti – rich and Nb–Al – rich layers, with the presence of MAX phase alongside TiC and other Ti–Al and Nb–Al intermetallic phases. The Nb/Ti 2 AlC system possessed a layered architecture, though the MAX phases were not found to be continuously present in each alternating layer. The second Ti/Ti 2 AlC system showed a non-lamellar nanocomposite microstructure and the formation of mixed Ti n+1 AlC n phases (a mix of n = 1, 2), and no indication of layering. Diffusion occurring between the metal/MAX layers in both cases, likely due to the elevated temperatures during the deposition process, is speculated as the likely cause of these resultant microstructures. The mechanical properties of both systems were evaluated using micromechanical (nanoindentation and micro-pillar compression) techniques, which demonstrated high strengths for both systems (Nb system: yield and instability strengths of 4.88 ± 0.1 GPa and 5.57 ± 0.03 GPa, Ti system: yield and instability strength of 5.61 ± 0.28 GPa and 6.21 ± 0.25 GPa). In conclusion, this work highlights the promising mechanical properties of metal/MAX multilayered depositions and summarizes the challenges in PVD synthesis of metal/MAX multilayered nanolaminates.

MAX phase↗
DOE OSTI2024

Laser powder bed fusion of oxide dispersion-strengthened IN718 alloys: A complementary study on microstructure and mechanical properties

In this study, two new grades of oxide dispersion strengthened (ODS) Inconel 718 (IN718) alloys were designed by the thermochemical CALPHAD method and produced by laser powder bed fusion (LPBF) technique. Alloys designated as IN718-YF and IN718-YFH, that consist Y 2 O 3 –FeO and Y 2 O 3 –FeO–Hf, respectively, were fabricated with >99.9 % densification using optimized process parameters. CALPHAD calculations were highly consistent with experimental findings, highlighting the formation of Al-containing Y–Ti–O and Y–Hf–O nano-oxides in both alloy types. Texture analyzes revealed no significant texture development in as-built (AB) or heat-treated (HT) alloys. Heat treatment was applied at 1050 °C for 1 h to enhance nano-oxide density. Further, the nano-oxide number density remained similar in IN718-YF while it decreased in IN718-YFH alloy as a result of carbide formation after the heat treatment. Besides, formation of secondary γ' particles was observed in the IN718-YFH/HT alloy. Even though the yield strengths of IN718-YF and IN718-YFH alloys in both AB and HT conditions were similar, the ductility of IN718-YFH was ~50 % less in almost all conditions compared to the ductility of IN718-YF. This has been shown to be as a result of irregular shaped micron-sized Y-Hf-O oxides, martensite formation in AB condition, increased amount of carbides and existence of secondary γ' particles in HT condition in IN718-YFH. High density of stacking faults (SF) forming at the interface of the nano-oxides have been detected in IN718-YF alloys. Besides dislocation/nanoparticle interactions, SFs which are responsible for the delocalization of the deformation improve the ductility of IN718-YF alloys. Overall, high temperature mechanical tests exhibit that both alloys have higher strength with improved ductility compared to the standard IN718 alloys, indicating the contribution of the nano-oxides.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Microstructural evolution and transmutation in tungsten under ion and neutron irradiation

This study aims to compare the effects of neutron and self-ion irradiation on the mechanical properties and microstructural evolution in W. Neutron irradiation at the HFR reactor to 1.67 dpa at 800 °C resulted in the formation of large Re and Os rich clusters and voids. The post-irradiation composition was measured using APT and verfified against FISPACT modelling. The measured Re and Os concentration was used to create alloys with equivalent concentrations of Re and Os. These alloys were exposed to self-ion irradiation to a peak dose of 1.7 dpa at 800 °C. APT showed that self-ion irradiation leads to the formation of small Os clusters, wheras under neutron irradiation large Re/Os clusters form. Voids are formed by both ion and neutron irradiation, but the voids formed by neutron irradiation are larger. By comparing the behaviour of W-1.4Re and W-1.4Re-0.1Os, suppression of Re cluster formation was observed. Irradiation hardening was measured using nanoindentation and was found to be 2.7 GPa, after neutron irradiation and 1.6 GPa and 0.6 GPa for the self-ion irradiated W-1.4Re and W-1.4Re-0.1Os. The higher hardening is attributed to the barrier strength of large voids and Re/Os clusters that are observed after neutron irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Effects of transition metal carbide dispersoids on helium bubble formation in dispersion-strengthened tungsten

Abstract The formation of helium bubbles and subsequent property degradation poses a significant challenge to tungsten as a plasma-facing material in future long-pulse plasma-burning fusion reactors. In this study, we investigated helium bubble formation in dispersion-strengthened tungsten doped with transition metal carbides, including TaC, ZrC, and TiC. Of the three dispersoids, TaC exhibited the highest resistance to helium bubble formation, possibly due to the low vacancy mobility in the Group VB metal carbide and oxide phases. Under identical irradiation conditions, large helium bubbles formed at grain boundaries in tungsten, while no bubbles were observed at the interfaces between the carbide dispersoid and tungsten matrix. Moreover, our results showed the interfaces could suppress helium bubble formation in the nearby tungsten matrix, suggesting that the interfaces are more effective in trapping helium as tiny clusters. Our research provided new insights into optimizing the microstructure of dispersion-strengthened tungsten alloys to enhance their performance.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Predicting short-range order evolution in WTaCrVHf refractory high-entropy alloys

Short-range order (SRO) in multicomponent concentrated alloys affects their mechanical response. Hence, is paramount to understand how composition modifies the chemical ordering in the system to design materials with optimal properties. Here, in this work, we present a methodology to predict the SRO and thermodynamic properties in chemically complex systems and apply it to the WTaCrVHf quinary alloy. We observe that the addition of Hf significantly modifies the SRO, mainly at intermediate to low temperatures, matching experimental observations.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Grain boundary softening from stress assisted helium cavity coalescence in ultrafine-grained tungsten

The formation of helium cavities in coarse-grained materials produces hardening proportional to the number density and size of the cavities and due to the interaction of dislocations with intragranular helium defects. In nanostructured metals containing a high density of interfacial sinks, preferential cavity formation in the grain boundaries instead produces softening that is often attributed to enhanced interfacial plasticity. Here, employing two grades of ultrafine-grained tungsten, we explore this effect using targeted implantation studies to map cavity evolution as a function of the irradiation conditions and quantify its impact on the mechanical response through nanoindentation. Softening is reported at implantation temperatures above the threshold for preferential grain boundary cavity formation but at a sufficiently low fluence prior to the growth of intragranular cavities. Collective changes in the mean cavity size, density, and morphology beneath a residual impression on an implanted surface indicate that cavity coalescence accompanied the reduction in hardness. Complementary atomistic simulations demonstrate that, in tungsten grain structures exhibiting softening, grain boundary bubble coalescence is driven by stress concentrations that further act to localize strain in the grain boundaries through cooperative deformation processes involving local atomic shuffling and sliding, dislocation emission, and even the nucleation of unstable twinning events.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Perspectives on novel refractory amorphous high-entropy alloys in extreme environments

Two new refractory amorphous high-entropy alloys (RAHEAs) within the W–Ta–Cr–V and W–Ta–Cr–V–Hf systems were herein synthesized using magnetron-sputtering and tested under high-temperature annealing and displacing irradiation using in situ Transmission Electron Microscopy. While the 14W-41Ta-26Cr-19V in at.% RAHEA (defined as WTaCrV RAHEA) was found to be unstable under such tests, additions of Hf in this system composing a new quinary 24W-40Ta-18Cr-5V-13Hf in at.% RAHEA (defined as WTaCrVHf RAHEA) was found to be a route to achieve stability both under annealing and irradiation. A new effect of nanoprecipitate reassembling observed to take place within the WTaCrVHf RAHEA under irradiation indicates that a duplex microstructure composed of an amorphous matrix with crystalline nanometer-sized precipitates enhances the radiation response of the system. Further, it is demonstrated that tunable chemical complexity arises as a new alloy design strategy to foster the use of novel RAHEAs within extreme environments. New perspectives for the alloy design and application of chemically-complex amorphous metallic alloys in extreme environments are presented with focus on their thermodynamic phase stability when subjected to high-temperature annealing and displacing irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Micromechanical properties of spherical and facetted He bubble loaded copper

Exploring new irradiation resistant materials requires understanding their mechanical responses to irradiation. Resistance to helium bubble formation and understanding bubble effects on the mechanical response of candidate materials are crucial factors to qualify materials as irradiation resistant. Here, in this paper, we explore the effect of spherical and facetted helium bubbles on the mechanical response of copper via in-situ micromechanical tensile testing at room temperature. Bubble formation and shape effects on strength and ductility, and their behavior on grain boundaries are discussed and compared to literature. Loading Cu with helium bubbles is shown here to increase strength but decrease ductility.

42 ENGINEERING↗
DOE OSTI2023

Assessing Manufacturability of the Oxide Dispersion Strengthened (ODS) 14YWT Alloy Fuel Cladding Tube Using SolidStir™ Technology

The 14YWT (Fe–14Cr–3W–0.4Ti–0.3Y2O3 (wt.%)) alloy is a potential candidatematerial for fabricating fuel cladding tubes and other structural components to be used in next-generation advanced nuclear reactors. Due to inherent limitations of current processing routes, this research explored a new variant of friction stir processing, commercially referred to as SolidStir™ Extrusion (patent pending) technology from Enabled Engineering to consolidate and extrude 14YWT powder into a fuel cladding tube. Several single-layer and multiple-layer consolidation experiments were carried out in an Inconel 625 die to determine the suitable processing parameters for SolidStir™ extrusion of the 14YWT alloy cladding tube. With these processing parameters, a 16–18 mm long tube of the 14YWT alloy with no visible defects was successfully extruded using this novel technology. Microstructural characterization results obtained using optical microscopy, scanning electron microscopy, energy dispersive spectroscopy, electron backscatter diffraction, and transmission electron microscopy will be presented.

Advanced Materials and Manufacturing Technology↗
DOE OSTI2022

Accelerated radiation tolerance testing of Ti-based MAX phases

MAX phases have recently attracted significant attention for potential nuclear applications due to their novel properties such as unique hexagonal-compact nanolayered crystal structure, high-machinability due to lower hardness levels compared to conventional ceramics, and high-chemical inertness. In order for MAX phases to be used in nuclear reactors, two aspects deserve detailed investigations: (i) their phase stability at high-temperatures and (ii) microstructural defect formation and recovery induced by energetic particle irradiation. To date, degradation mechanisms of MAX phases at high-temperatures and following irradiation are largely unexplored fields of research. This work focuses on the evaluation of two Ti-based MAX phases – Ti 2 AlC and Ti 3 SiC 2 – within the context of extreme environments. To accomplish this, a one-of-a-kind comparison between neutron irradiations, performed over a decade of research at the High Flux Isotope Reactor (HFIR), and heavy-ion irradiations, carried out in situ in a Transmission Electron Microscope (TEM) at the MIAMI-2 facility, has been conducted. The results show Ti-based MAX phases are prone to accelerated decomposition under the conditions investigated. This questions the hypothesis that MAX phases exhibit high phase stability, especially when used in future nuclear energy systems where energetic particle irradiation is a dominating degradation mechanism.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Implications of Microstructure in Helium-Implanted Nanocrystalline Metals

Helium bubbles are known to form in nuclear reactor structural components when displacement damage occurs in conjunction with helium exposure and/or transmutation. If left unchecked, bubble production can cause swelling, blistering, and embrittlement, all of which substantially degrade materials and—moreover—diminish mechanical properties. On the mission to produce more robust materials, nanocrystalline (NC) metals show great potential and are postulated to exhibit superior radiation resistance due to their high defect and particle sink densities; however, much is still unknown about the mechanisms of defect evolution in these systems under extreme conditions. Here, the performances of NC nickel (Ni) and iron (Fe) are investigated under helium bombardment via transmission electron microscopy (TEM). Bubble density statistics are measured as a function of grain size in specimens implanted under similar conditions. While the overall trends revealed an increase in bubble density up to saturation in both samples, bubble density in Fe was over 300% greater than in Ni. To interrogate the kinetics of helium diffusion and trapping, a rate theory model is developed that substantiates that helium is more readily captured within grains in helium-vacancy complexes in NC Fe, whereas helium is more prone to traversing the grain matrices and migrating to GBs in NC Ni. Our results suggest that (1) grain boundaries can affect bubble swelling in grain matrices significantly and can have a dominant effect over crystal structure, and (2) an NC-Ni-based material can yield superior resistance to irradiation-induced bubble growth compared to an NC-Fe-based material and exhibits high potential for use in extreme environments where swelling due to He bubble formation is of significant concern.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Stable, Ductile and Strong Ultrafine HT-9 Steels via Large Strain Machining

Beyond the current commercial materials, refining the grain size is among the proposed strategies to manufacture resilient materials for industrial applications demanding high resistance to severe environments. Here, large strain machining (LSM) was used to manufacture nanostructured HT-9 steel with enhanced thermal stability, mechanical properties, and ductility. Nanocrystalline HT-9 steels with different aspect rations are achieved. In-situ transmission electron microscopy annealing experiments demonstrated that the nanocrystalline grains have excellent thermal stability up to 700 °C with no additional elemental segregation on the grain boundaries other than the initial carbides, attributing the thermal stability of the LSM materials to the low dislocation densities and strains in the final microstructure. Nano-indentation and micro-tensile testing performed on the LSM material pre- and post-annealing demonstrated the possibility of tuning the material’s strength and ductility. The results expound on the possibility of manufacturing controlled nanocrystalline materials via a scalable and cost-effective method, albeit with additional fundamental understanding of the resultant morphology dependence on the LSM conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Limitations of Thermal Stability Analysis via In-Situ TEM/Heating Experiments

This work highlights some limitations of thermal stability analysis via in-situ transmission electron microscopy (TEM)-annealing experiments on ultrafine and nanocrystalline materials. We provide two examples, one on nanocrystalline pure copper and one on nanocrystalline HT-9 steel, where in-situ TEM-annealing experiments are compared to bulk material annealing experiments. The in-situ TEM and bulk annealing experiments demonstrated different results on pure copper but similar output in the HT-9 steel. The work entails discussion of the results based on literature theoretical concepts, and expound on the inevitability of comparing in-situ TEM annealing experimental results to bulk annealing when used for material thermal stability assessment.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Comparison of void swelling of ferritic-martensitic and ferritic HT9 alloys after high-dose self-ion irradiation

In this work, ferritic-martensitic (F/M) HT9 along with fully ferritic HT9 specimens were irradiated using self-ions to 600 peak displacement-per-atom (dpa) at 450 °C. To investigate and minimize the carbon contamination effect on void swelling, two identical F/M HT9 specimens were prepared: one without a coating and the other one with a tungsten and silver coating on the surface prior to irradiation. The specimen without a coating showed carbide formation in the matrix, resulting from carbon contamination during irradiation, with no voids observed. The coated sample did not form matrix carbides, and voids were observed within the irradiated region. The same coating was applied to the ferritic HT9, and void swelling results of both the F/M HT9 and the ferritic HT9 alloys were compared after irradiation. The F/M HT9 exhibited a maximum of 2.6% swelling, while the ferritic HT9 showed a maximum of 0.8% swelling at a 400–500 nm depth, equivalent to 334 local dpa. The results were correlated with the instability of carbides in the ferritic HT9 under irradiation. Carbides in the F/M HT9 and the ferritic HT9 are both M23C6, but the matrix carbides in the ferritic HT9 were less stable under the irradiation, compared to the ones on the grain boundaries in the F/M HT9. The study shows the impact of chemistry on the void swelling which overpowers the microstructural factors such as grain size and phase.

36 MATERIALS SCIENCE↗