Dr. Nabil Daghbouj, Ph.D.

Cr coatings, as protective coatings of Zr-alloy fuel claddings, inevitably suffer from irradiation damage before they would possibly run into the accident condition. This study evaluates the radiation and oxidation tolerance of three Cr-based coatings with different microstructures (Cr, CrAlSi, and CrAlSiN) through He2 + ion irradiation and 1200 degrees C steam oxidation. The Cr and CrAlSi coatings experienced significant structural degradation, characterized by He bubble aggregation and amplified Kirkendall effects at elevated temperatures. In contrast, the irradiated CrAlSiN coating maintained structural integrity without measurable irradiation hardening. Following annealing at 800 degrees C for 30 min, approximately 40 % of injected He atoms were released, indicating a "self-healing" mechanism. The mechanism is attributed to uniformly distributed, low-density channels that act as sinks and release paths for irradiation-induced defects. Density functional theory simulations suggest that N atoms promote significant rearrangement of ions surrounding the free volume, inhibiting the formation of sites capable of trapping He atoms. Moreover, the CrAlSiN coating exhibited superior oxidation resistance compared to the Cr and CrAlSi coatings, even under high-temperature steam conditions. Notably, the irradiated CrAlSiN sample displayed a significantly thinner oxide scale compared to the pristine one (almost half), owing to a more protective oxide scale and rapid outward diffusion of Cr, Al, and Si through nanochannel veins. These findings illuminate the effects of structure and composition on irradiation and oxidation behavior in Cr-based coatings, offering insights for developing new-generation accident-tolerance fuel coatings for Zr-alloy claddings. (c) 2024 Published by Elsevier Ltd on behalf of The editorial office of Journal of Materials Science & Technology.

The wear behavior of two amorphous and polycrystalline forms of MoS2 prepared by magnetron sputtering is characterized in a combined nanoindentation and atomic force microscopy study supported by Raman and transmission electron microscopy (TEM) analysis. From the morphology of wear tracks estimated after scratching the surfaces with a Berkovich indenter and a loading force up to 2 mN, we conclude that, on the microscale, both forms follow the Archard’s wear equation, and the wear resistance is about four times higher on the amorphous MoS2. The coefficient of friction is much lower on the worn areas, which is associated to significant smoothing of the surfaces caused by the scratching process. With normal forces in the μN range, the analysis is made difficult by the fact that the linear dimensions of the wear tracks are comparable to those of the smallest surface features. Even if the Archard’s equation looses validity, the wear resistance is considerably larger on amorphous MoS2 also on the nanoscale. Our results concluded that the polycrystalline form of MoS2 has poor tribological properties at the micro/-nanoscale as compared to the amorphous form and hence less suited as a solid coating in ambient conditions.

First-principles calculations are used to investigate the effects of stacking faults (SFs) on helium trapping and diffusion in cubic silicon carbon (3C-SiC). Both extrinsic and intrinsic SFs in 3C-SiC create a hexagonal stacking sequence. The hexagonal structure is found to be a strong sink of a helium interstitial. Compared to perfect 3C-SiC, the energy barriers for helium migration near the SFs increase significantly, leading to predominant helium diffusion between the SFs in two dimensions. This facilitates the migration of helium towards interface traps, as confirmed by previous experimental reports on the nanocrystalline 3C-SiC containing a high density of SFs. This study also reveals that the formation of helium interstitial clusters near the SFs is not energetically favored. The findings from this study enhance our comprehension of helium behavior in faulted 3C-SiC, offering valuable insights for the design of helium-toleran SiC materials intended for reactor applications.

The surface coating technology, encompassing ceramics, refractory materials, metallic alloys containing Al or Si, and multicomponent composites, presents a viable approach to improve the corrosion resistance of ferritic/martensitic (F/M) steels with (9–12) wt.% Cr in liquid lead-bismuth eutectic (LBE) environment. Among these coating materials, chromium (Cr) coating emerges as a particularly noteworthy option. This study specifically focused on depositing a 3 μm thick Cr coating on on T91 and SIMP steels using magnetron sputtering. Subsequently, the corrosion behavior of the Cr coating was investigated in LBE at temperatures of 600 °C and 700 °C. The results revealed that, after 300 h at 600 °C, T91 and SIMP steels formed oxide scales with approximately 32.6 μm and 19.3 μm thicknesses, respectively. At 700 °C for 140 h, these oxide scales increased to about 82.4 μm and 73.1 μm for T91 and SIMP steels, respectively. However, the application of a Cr coating resulted in the formation of a dense layer of chromium oxide with a thickness of 4–5 μm. This layer effectively impeded oxygen diffusion and Fe migration leading to a significant reduction in the corrosion rate of the steel. Notably, the Cr coating maintained secure attachment to the steel even after exposure to high-temperature LBE corrosion. These findings underscore the capacity of coating to markedly enhance the corrosion resistance of T91 and SIMP steels in high-temperature LBE environments, providing robust protection against the detrimental effects of challenging conditions. Consequently, Cr coating emerges as a promising solution for future fission nuclear reactors.

In this study, single crystal (sc) and nanocrystalline (nc) 3C-SiC samples were subjected to 30 keV He ion irradiation across various doses while maintaining a temperature of 800 °C. Employing techniques including Raman spectroscopy, transmission electron microscopy (TEM), and nanoindentation, the alterations in microstructure and hardness resulting from He irradiation with various fluences were examined. In sc-SiC, irradiation prompted the formation of He platelets, resulting in a hardness increase of 7 GPa. In contrast, nc-SiC, characterized by a higher stacking fault density, exhibited the formation of bubbles, primarily at grain boundaries (GBs), with fewer occurrences within the grain interior, leading to a hardness increase of 1 GPa. Notably, in both sc- and nc-SiC, hardness reached saturation and subsequently stabilized or declined with increasing He fluence. Through molecular dynamics (MD) cascade simulations, we discerned that various planar defects do not uniformly contribute to enhancing radiation resistance. For example, intrinsic stacking faults (ISF) and twins in SiC played a substantial role in altering defect density and configurations, thereby facilitating point defect annihilation. Conversely, extrinsic stacking faults (ESF) and Σ3 GBs had a limited impact on defect production during a cascade. Furthermore, calculations of cluster diffusivity revealed an accelerated movement of He-vacancy towards GBs compared to bulk material and other planar defects. Moreover, the scarcity of point defects and constrained mobility of He atoms towards stacking faults in nc-SiC elucidated the marked tendency of He to form platelets in sc-SiC. Additionally, our findings established a correlation between the calculated indentation hardness and the geometry of He defects, consistent with experimental results from nanoindentation. These results significantly contribute to ongoing efforts to design SiC materials with heightened radiation tolerance.

This study investigates the intricate mechanisms that govern irradiation damage in high-entropy ceramic materials. Specifically, we synthesized (Hf0.2Zr0.2Ta0.2Nb0.2Ti0.2)C high-entropy carbide ceramics (HECC) with a single-phase rock-salt structure using spark plasma sintering. These ceramics were then subjected to irradiation with 1.08 MeV C ions, resulting in a dose of 7.2 dpa (dpa: displacements per atom) at both room temperature (RT) and 500 ◦C. To understand the resulting damage structure, we analyzed bulk irradiated HECC samples using Grazing Incidence X-ray Diffraction (GIXRD) and Transmission Electron Microscope (TEM) at both irradiation temperatures. GIXRD analysis revealed an average tensile strain out-of-plane of 0.16% for RT irradiation and 0.14% for irradiation at 500 ◦C. In addition, TEM analysis identified a buried damaged band, approximately 970 nm thick, under both irradiation temperatures. By employing the bright field TEM imaging technique under kinematic two-beam conditions, dislocation loops of both a/3 〈111〉{111} and a/2 〈110〉{110} types within the damaged band were observed. Furthermore, our analysis indicated an increase in the average size of the total dislocation loops within the band from 1.2 nm to 1.4 nm as the density decreased. Importantly, no amorphization, precipitates, or voids were detected in the damaged band under both irradiation temperatures. Density functional theory (DFT) simulations indicated that carbon predominantly resides in 〈110〉split interstitial sites causing lattice expansion, while vacancies, particularly Nb, induced compression along the c-axis. Carbon atoms tend to bond when collectively present in the <110> split interstitial sites, contributing to the formation of interstitial loops.

Layered carbides and borides, blending ceramic and metallic characteristics, present compelling prospects for future nuclear reactor applications due to their complex nanolaminate crystal structure. To explore the behavior of Fe2AlB2 under different conditions, helium irradiation is conducted across saturation fluences and under different annealing and irradiation temperatures. Room temperature (RT) irradiation causes some changes: a-lattice expands 0.3 %, c-lattice contracts 0.07 %, and an embedded amorphous layer into the matrix is formed. 400 °C annealing keeps an amorphous structure whereas 800 °C annealing fully recrystallizes with bubble growth. Increasing the irradiation temperature to 400 °C with the same fluence, smaller bubbles form without an amorphous layer, indicating defect recombination depends on helium attachment to vacancies. Under irradiation at 700 °C, faceted bubbles develop exclusively within Fe2AlB2, with no such bubbles in Al2O3. These faceted bubbles align themselves along (110) planes. Simultaneously, irradiation prompts tangled dislocations within the damaged layer, fostering the congregation of polygonal bubbles along grain boundaries (GBs) and creating denuded areas at GBs. Fe2AlB2 demonstrates remarkable oxidation resistance under high-temperature irradiation, maintaining surface stability without the cracks seen in ion-irradiated Ti3SiC2 and Ti2AlC at RT. Fe2AlB2 shows promise for use in reactors operating at temperatures above 500 °C due to its resilience to irradiation. Further investigation is warranted for its application in reactors operating in environments exceeding this threshold temperature.

This study investigates the effects of applied tensile and compressive stresses on the evolution of defects in He-irradiated silicon carbide (SiC) materials. By combining experimental studies with density-functional theory (DFT) simulations, we systematically analyzed the microstructural changes and defect formation mechanisms in SiC under stress. The research focused on He irradiation of SiC at 750 °C, with a fluence of 1 × 101⁷ He/cm2. The results revealed that the strain resulting from radiation damage depends on the applied stress during irradiation. Moreover, the formation of platelets is influenced by this applied stress: tensile stress promotes platelet growth, while compressive stress inhibits it. DFT simulations further supported these experimental findings by showing that under tensile strain, both carbon vacancies (Cv) and He atoms exhibit low energy barriers for migration. This phenomenon facilitates platelet growth, aligning well with the observations made in the experiments. However, when considering applications like semiconductor thin film transfer, such as the Smart-cut technique, He implantation under tensile stress can actually be advantageous. This approach enables the use of lower He fluence, which in turn reduces the overall cost of the operation. By strategically leveraging the effects of tensile stress on He implantation, it becomes possible to optimize processes like Smart-cut for more efficient and cost-effective thin film transfer in semiconductor manufacturing.

This study explores two vital structural materials, T91 (Fe-9Cr) and SIMP (Fe-11Cr) steels, in the context of lead-cooled fast reactors and accelerator-driven sub-critical systems (ADS). Lead-bismuth eutectic (LBE) functions as a key coolant and spallation target material due to its impressive thermal conductivity, neutron yield, and chemical properties. Unfortunately, materials in contact with LBE are prone to severe corrosion at elevated temperatures (T>500 °C), compromising their integrity. To bolster corrosion resistance, we utilized laser remelting and laser cladding to apply FeCrAl/TiN coatings on the steel surfaces. Our study scrutinizes the corrosion behavior of steel in LBE saturated with oxygen at 700 °C and investigates the underlying causes. Following 240 h of exposure to corrosion, T91 and SIMP steels subjected to laser remelting displayed substantial oxide scale formation. Lead-bismuth atoms infiltrated the outer oxide layer (Fe3O4), diminishing adhesion between the inner and outer oxide layers, leading to the detachment of the outer oxide layer. The inner oxide layers, composed of Fe-Cr spinel, were approximately 123 μm thick for T91 steel and 77 μm for SIMP steel, underscoring SIMP steel's superior corrosion resistance. For T91 steel treated with laser cladding FeCrAl/TiN coating, a characteristic duplex oxide layer with a total thickness of around 83 μm was formed, with noticeable deposition of Pb-Bi atoms at the interface between the outer and inner oxide layers. Conversely, only a protective alumina layer safeguarded SIMP steel from LBE corrosion. This outcome emphasizes the efficacy of laser cladding FeCrAl/TiN coating in providing superior protection for SIMP steel over T91 steel. Our research significantly contributes to the development of anti-corrosion coatings for high-temperature LBE environments.

The effect of liquid lead-bismuth eutectic (LBE) on the corrosion evolution of iron-based materials with varying densities of edge dislocation defects was studied. Molecular dynamics simulations were conducted at temperatures ranging from 823 K to 1173 K, and the process of penetration of Pb and Bi atoms at the dislocation core was analyzed. The results indicated that Fe atoms near the dislocation core have the lowest substitution energy (-4.79 eV/atom), making them more likely to be replaced by Pb and Bi atoms. Moreover, it was found that the increase in dislocation density did not result in a high penetration depth of LBE atoms, attributed to the interaction of dislocations. These findings provide crucial insights into the LBE corrosion mechanism in iron-based materials with dislocation defects.

Zr/Nb6 multilayers of 6 nm periodicity, with well-composition-modulated structures, were prepared by magnetron sputtering. Their microstructure and scratch properties were investigated using HAADF-STEM, XRD, AFM, and triboindenter. The Zr/Nb interfaces have various orientations along the growth direction. The hardness (H) and reduced elastic modulus (E) are measured as 6.6 GPa, and 176.3 GPa, respectively, resulting in a high ratio of H/E, compared to other multilayer systems such as Ta/Co and Ag/Cu, which indicates superior tribological performance. The coefficient of friction (COF) was 0.27, and the elastic recovery was observed along the scratch path. Extensive large molecular dynamics simulations (MD) were conducted to investigate the impact of different Zr/Nb interface orientations on the friction/wear behavior of Zr/Nb6 multilayers. The primary cause of plastic deformation of the Nb layer was dislocations and BCC twinning, while Zr layers deform via dislocations and intrinsic stacking faults. The Zr/Nb6 exhibited better tribological properties, such as lower COF, higher scratch hardness, and improved wear resistance compared to their single-crystal counterparts. The Pitsch-Schrader interface showed the lowest COF value, whereas Rong–Dunlop and Zhang-Killy orientations exhibited better wear resistance. The interface structure was analyzed, and its blocking strength was discussed. These findings contribute to understanding the relationship between Zr/Nb interface and wear performance and tailoring them to achieve desired properties for specific applications.

The study explored the microstructure evolution of 6H-SiC that underwent sequential iron and helium ion irradiation with energies of 2.5 MeV and 500 keV, respectively, at room temperature, followed by annealing at 1500°C for two hours. Following irradiation, the entire damaged layer underwent amorphization. However, during subsequent annealing, epitaxial recrystallization took place, resulting in the formation of defected polycrystalline 6H-SiC characterized by the presence of Fe-rich clusters, cavities, and stacking faults. Fe-rich cavities were found to predominantly form at the edges of the stacking faults, as revealed by XTEM. The interaction of microstructural defects is further investigated via first-principles calculations. The periphery of the stacking faults has been identified as the primary location for the emergence of vacancy clusters, serving as favorable sites for the accumulation of point defects, including Fe atoms. This behavior can be attributed to the combined effects of mechanical and electronic energy relaxation mechanisms. Mechanically, the presence of stacking faults allows for the release of elastic energy that had been stored at the boundary. Electronically, the energy relaxation arises from the saturation of C- and Si-dangling bonds. Both of these processes contribute to the observed behavior, highlighting the intricate interplay between mechanical and electronic factors in the system. The low point defect migration energy barriers in the vicinity of the stacking faults promise high recombination, which can limit cavity growth and enhance radiation resistance. The study not only offers valuable insights into the mechanism of cavity/stacking faults interaction, contributing to a better understanding of radiation damage in 6H-SiC but also demonstrates that 6H-SiC material containing stacking faults could serve as a viable alternative to 3C-SiC for nuclear application.

Microstructure of radiation-induced Iron phases were investigated in a 6H-SiC subjected to Iron and Helium bombardment with a damage level of 8 dpa. The microstructural evolution before and after annealing was investigated by combining transmission electron microscopy (TEM, STEM-EDS), automated crystal phase and orientation imaging (ACOM-TEM), secondary ion mass spectroscopy (SIMS), and atomic scale simulations. The irradiation amorphized the entire damaged layer which contains an embedded band of He bubbles located at peak damage concentration. After annealing, the amorphous layer recrystallized into a polycrystalline 6H-SiC where the Fe profile significantly changed to form Fe-rich clusters. ACOM-TEM reveals the formation of large cubic FeSi clusters and small bcc-Fe precipitates located at the 6H-SiC grain boundaries. The type and size distribution of the precipitates greatly depend on the Fe profile. Fe-Si compounds form around the Fe peak concentration, while, bcc Fe precipitates tend to be more homogeneously distributed. Density functional theory (DFT) calculations demonstrate that the formation of Fe dimers and trimers in the 1st nearest neighbor is energetically favorable. A combined Monte Carlo/Classical molecular dynamic (MMC/MD) technique reveals that the Fe atoms prefer to form large clusters in accordance with experimental results. MD annealing simulations reveal the formation of stable bcc Fe at high temperatures. The phase transition starts at the cluster-matrix interface around 620 K and the cluster is fully transformed at 700 K.

In this work, the radiation responses of Zr/Nb nanostructured metallic multilayers (NMMs) are studied. The nanostructures with different layer thicknesses were deposited on Si (111) substrate by using magnetron sputtering and were subjected to heavy-ion irradiation at room temperature with different fluences. Nanoindentation, XRD, DFT, SIMS, and Variable Energy Positron Annihilation Spectroscopy (VEPAS) techniques were used to study the type and distribution of defects, and strain within the material as well as the changes in the hardness of the structures as a function of damage. Our results suggest that the strain and the irradiation hardening are layer thickness- and damage-dependent while they are independent of the type of irradiated ions. The magnitude of hardening decreases with decreasing individual layer thickness indicating that the number of interfaces has a direct effect on the radiation tolerance enhancement. For thin layers with a periodicity of 27 nm (Zr/Nb27), a transition from hardening to softening occurs at high fluence, and a saturation point is reached in thick layers with a periodicity of 96 nm (Zr/Nb96). The as-deposited thin multilayers presented a significantly higher atomic-scale disorder which increases with ion irradiation compared to the thick multilayers. VEPAS reveals the vacancy defects before and after irradiation that contribute to the presented strain. Based on the findings, thin nanostructured Zr/Nb multilayered structures possess excellent radiation resistance due to the high density of interfaces that act as sinks for radiation-induced point defects.

Nanolayered metallic alloys are promising materials for nuclear applications thanks to their resistance to radiation damage. Here, we investigate the effect of ion (C, Si, and Cu) irradiation at room temperature with different tluences into sputtered Zr/Nb metallic multilayer films with periods 27 nm (thin) and 96 nm (thick). After irradiation, while a high strain in the entire thin nanoscale metallic multilayer (NMM) is observed, a quite small strain in the entire thick NMM is established. This difference is further analyzed by a semianalytical model, and the reasons behind it are revealed, which are also validated by local strain mapping. Both methods show that within a thick layer, two opposite distortions occur, making the overall strain small, whereas in a thin layer, all the atomic planes are affected by the interface and are subjected to only a single type of distortion (Nb-tension and Zr-compression). In both thin and thick NMMs, with increasing damage, the strain around the interface increases, resulting in a release of the elastic energy at the interface (decrease in the lattice mismatch), and the radiation-induced transition of the Zr/Nb interfaces from incoherent to partially coherent occurs. Density functional theory simulations decipher that the inequality of point defect diffusion flux from the inner to the interface-affected region is responsible for the presence of opposite distortions within a layer. Technologically, based on this work, we estimated that Zr/NbSS with thicknesses around Zr = 24 nm and Nb = 31 nm is the most promising multilayer system with the high radiation damage resistance and minimum swelling for nuclear applications.

Tailoring interfaces is a powerful way of reducing the accumulation of radiation defects. Understanding strain evolution induced by ion bombardment in nuclear materials with high interface density is crucial for next-generation reactors since induced defects are responsible for volumetric swelling and catastrophic failure. X-ray and selected-area diffraction patterns (SADPs) measurements reveal, after Cu implantation, that a relatively high out-of-plane strain is created in thin Zr/Nb-6 multilayers, while thick Zr/Nb96 is barely distorted. The absence of layer deformation in Zr/Nb-96 is explained by local TEM strain mapping showing the presence of two oppositely distorted regions (inner and interface-affected regions) within one layer producing only a small overall strain, whereas the whole individual layers of Zr/Nb-6 are affected by the interface manifesting high strain. Using MD simulations, the types of defects responsible for layer distortion are identified. The opposite distortion within the layer is attributed to the inequality of the defect flux from the inner to interface-affected region due to the difference in migration energy barriers of the point defects. Moreover, the interface sink efficiency (defect annihilation) is determined for Zr/Nb as an illustration which provides a strategy for designing new derivate structures of multilayers with high radiation damage resistance.

SiC is considered a perspective material in advanced nuclear systems as well as for electronic or spintronic applications, which require an ion implantation process. In this regard, two sets of 6H-SiC samples were implanted with i) 2.5 MeV Fe ions and ii) 2.5 MeV Fe ions and co-implanted 500 keV He ions at room temperature and then annealed at 1500 degrees C for 2 h. The microstructure evolution and Fe diffusion behavior before and after annealing were characterized and analyzed. After annealing, Fe concentration is enhanced close to the surface in the Fe-implanted sample, whereas in the co-implanted system, Fe atoms are redistributed into two distinct, spatially separated regions (close to the surface, and around the He-induced defects). The reason behind this finding is explained from an energetic point of view by using ab initio simulations. Technologically, the preexisting cavities can be used to control the Fe diffusion.

The blistering efficiency in He-H-ions co-implanted and annealed InP has been found to peak and vanish in a narrow range of ion fluence ratio (?H/?He = 1.5?3.5) with a fixed He fluence of 2 ? 1016 He+/cm2. The blisters are formed at low fluence (?H/?He = 1.5), peaked in the middle (?H/?He = 2.5), and disappeared at the high fluence ratio (?H/?He = 3.5). To get a fundamental understanding of blister formation in nanoscale, the defect profiles were studied by various experimental techniques combined with FEM and ab-initio simulations. Crosssection TEM images showed that at a low fluence ratio, He and H are stored in microcracks and bubbles whereas, at a high fluence ratio, the ions are trapped only inside bubbles. These atomic processes that occur during and after co-implantation and annealing are presented together with detailed scenarios in an attempt to explain our results. Based on DFT simulations, the de-trapping of He atoms from the small clusters is energetically cheaper compared to the migration of He from the large clusters formed at high fluence. Moreover, at a high fluence ratio, the presence of large clusters inhibits the He diffusion to the small clusters (precursor of blisters) by capturing migrating He atoms.

Sputter-deposited Zr/Nb nanoscale metallic multilayers with a periodicity of 27 (thin) and 96 nm (thick) were subjected to Cu + implantation with low and high fluences and then studied using various experimental techniques in combination with DFT calculations. After Cu + implantation, the thinner multilayer exhibited a tensile strain along c-axis in Nb layers and a compressive strain in Zr layers, while the thicker multilayer showed a compressive strain in both layers. The strain is higher in the thin multilayer and increases for higher fluences. We developed a mathematical method for the fundamental understanding of the deformation mechanisms in metallic multilayers subjected to radiation damage. In the model, the cumulative strain within a layer is described as the combination of two contributions coming from the interfacial region and the inner region of the layers. The semi-analytical model predicts that the interfacial strain is dominant and extends over a certain region around the interface. Predictions are well supported by ab-initio calculations which show that in the vicinity of the interface and in the Zr side, vacancies and interstitials (low energy barriers) exhibit high mobility compared to the Nb side, thus resulting in a high recombination rate. As a consequence, less strain occurs in the Zr side of the interface compared to the Nb side. The density and distribution of various types of defects along the ion profile (low and high damaged regions) are obtained by combining DFT results and the predictions of the model. (c) 2020 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

The defect evolution in InP with the 75 keV H+ and 115 keV He+ implantation at room temperature after subsequent annealing has been investigated in detail. With the same ion implantation fluence, the He+ implantation caused much broader damage distribution accompanied by much higher out-of-plane strain with respect to the H+ implanted InP. After annealing, the H+ implanted InP did not show any blistering or exfoliation on the surface even at the high fluence and the H-2 molecules were stored in the heterogeneously oriented platelet defects. However, the He molecules were stored into the large bubbles which relaxed toward the free surface, creating blisters at the high fluence.

As for a newly developed tempered martensitic steel, SIMP with 1.4 wt.% Si, used in the Chinese Initiative Accelerator Driven System (CiADS), so it is worth investigating the stress corrosion property of SIMP steel in liquid lead-bismuth eutectic at the system operating temperature. The compatibility of the structural materials with the proposed operational conditions was investigated by performing corrosion-mechanical testing. The stress corrosion was performed in static lead-bismuth eutectic at 300 degrees C, 450 degrees C, and 500 degrees C with different loading stresses. The effect of stress on corrosion rate at different temperatures was investigated by scanning electron microscopy and transmission electron microscopy. The testing of SIMP steel showed that the corrosion rate in the presence of LBE was strongly temperature-dependent. At 300 degrees C, only a very thin oxide scale was formed which inhibits the Pb and O from penetrating inside matrix steel, and hence keeps it ductile (no crack). On the other hand, at 450 and 500 degrees C, the stress can significantly enhance the corrosion rate at 450 degrees C and 500 degrees C, but not at 300 degrees C. Reasons are investigated and discussed based on the available space model.

Microstructural evolution and deformation mechanisms of magnetron sputtered V-Si-N coatings with various Si contents are investigated by transmission electron microscopy, X-ray absorption spectroscopy, and ab initio calculations. A small amount of Si atoms was dissolved into the cubic VN lattice, locally reducing the neighboring V-N p-d hybridization near the Si site. The Si content was found to impact the architecture of coating significantly. With increasing Si content, the microstructure evolved through three different architectures: (i) highly textured columnar grains, (ii) refined columnar grains, and (iii) nanocomposite structures where elongated grains were bounded by vein-like boundaries. Enhanced damage tolerance was observed in the nanocomposite structure, where multiple toughening mechanisms become active. Ab initio calculations revealed that the incorporation of Si monolayer in the (111) -oriented VN resulted in the formation of weaker Si-N bonds compared to V-N bonds, which allowed a selective response to strain and shear deformations by assisting the activation of the slip systems. (c) 2021 Published by Elsevier Ltd. This is an open access article under the CC BY-NC-ND license (http:// creativecommons.org/licenses/by-nc-nd/4.0/).

The microstructure evolution of hydrogen-implanted 6H-SiC at different temperatures and fluences is investigated by using various experimental techniques. In H-implanted samples with relatively low fluence at RT, dense blister cavities are observed after annealing at 1100 degrees C, while no visible blister cavities appear after annealing at 1100 degrees C in the sample implanted at RT with high fluence. The absence of blister cavities is due to the loss of elastic energy during the crystalline-to-amorphous transition. With a further increase of implantation temperature to 450 and 900 degrees C, amorphization did not occur and H-containing microcracks grew laterally below the surface. Thus, blisters appeared on the surface of the samples implanted at 900 degrees C even without annealing. The results are compared to the microstructural evolution of He-implanted 6H-SiC which was explored in our previous work. The behavior of hydrogen and helium ions in 6H-SiC lattice was rather different. For He implantation, regardless of the fluence and implantation temperature, blisters did not form. The mechanism of migration and coalescence of nanoscale bubbles that are responsible for blistering were studied via density functional theory calculations, which well-supported the presented results. We found that both mechanisms (migration and coalescence) are energetically cheaper in the case of H compared to He. (C) 2020 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

Sputter-deposited Zr/Nb nanometric multilayer films with a periodicity (L) in the range from 6 to 167 nm were subjected to carbon, silicon and copper ion irradiation with low and high fluences at room temperature. The ion profiles, mechanical proprieties, and disordering behavior have been investigated by using a variety of experimental techniques (Secondary Ion Mass Spectrometry - SIMS, nanoindentation, X-ray diffraction - XRD, and scanning transmission electron microscopy - STEM). On the STEM bright field micrographs there is damage clearly visible on the surface side of the multilayer; deeper, the most damaged and disordered zone, located close to the maximum ion concentration, was observed. The in-depth C and Si concentration profiles obtained from SIMS were not affected by the periodicity of the nanolayers. This is in accordance with SRIM simulations. XRD and electron diffraction analyses suggest a structural evolution in relation to L. After irradiation, Zr (0002) and Nb (110) reflexions overlap for L=6 nm. For the periodicity L> 6 nm the Zr (0002) peak is shifted to higher angles and Nb (110) peak is shifted to lower angles.

The microstructural phenomena occurring in 6H–SiC subjected to different irradiation conditions and annealing temperatures were investigated to assess the suitability of 6H–SiC as a structural material for nuclear applications. To this aim, a single crystal of 6H–SiC was subjected to He+ irradiation at 300 keV with different fluences and at temperatures ranging from 25 to 750 °C. Rutherford backscattering/channeling (RBS/C), X-ray diffraction (XRD) and transmission electron microscopy (TEM) analyses were combined to shed light on the microstructural changes induced by irradiation and subsequent annealing (750 to 1500 °C). At room temperature, amorphization starts to occur at a fluence of 2.5 × 1016 cm−2 (0.66 dpa). On the contrary, amorphization was prevented at high irradiation temperatures and fluences. Furthermore, a thin and highly strained region located around the maximum He concentration (Rp) formed. This region results from the accumulation of interstitial atoms which are driven toward the highly damaged region under the actions of a strain gradient and high temperature. Regardless of the fluence and irradiation temperature, the material stores elastic energy, which leads to the trapping of He in dissimilar defect geometries. For irradiation temperatures below 750 °C, helium was accumulated in bubbles which coarsened after annealing. On the other hand, for an irradiation temperature of 750 °C, helium was trapped in platelets (even for medium fluence), which evolved into a homogeneous dense array of cavities during annealing. DFT calculations show that the bubbles are under high pressure and contribute to developing the overall tensile strain in the single crystal 6H–SiC.