An, as cast, VCrFeW0.2 refractory medium entropy alloy (RMEA) was designed for fusion reactor divertor applications, focusing on reduced cost, low activation and compositional stability (low transmutation rates). The as-cast alloy was irradiated to a fluence of 5.6 x 10(17) ions/cm(2) at room temperature with 5 MeV helium ions whose energy have been uniformly attenuated to 0.4 MeV and 5 MeV via energy degradation device prior to sample irradiation. Pre and post irradiation, its mechanical properties were evaluated micro-tensile testing. Prior to irradiation, the VCrFeW0.2 alloy demonstrated good strength and ductility, with a yield strength of 1464 MPa and strain to UTS (epsilon(UTS)) of 4.6 %, maintaining comparable strength to pure tungsten (1403 MPa) but with greater strain to UTS (1.3 %). Post irradiation, the VCrFeW0.2 alloy exhibited remarkable damage resistance; its strength increased by only similar to 160 MPa, and it retained strain to UTS with a (epsilon(UTS)) of 2.9 %. It performed better than pure tungsten tested under identical irradiation conditions where there was similar to 1800 MPa increase in yield strength and a complete loss of plasticity. The micro-tensile results were supported by nanoindentation tests and Vickers hardness testing was also undertaken to show the yield strength values are representative of macro scale, bulk behavior. TEM and comparison with existing literature on RMEA/RHEA are presented here to understand the reason for difference in performance between VCrFeW0.2 alloy and pure tungsten.

Blistering of pure tungsten by MeV helium (He) bombardment was investigated by examination of the thickness of the blister skin. Polycrystalline tungsten (W) was fabricated by powder metallurgy and irradiated with, 4 MeV He2+, of fluence 1018 ions cm- 2, and a flux of 7 x 1013 ions cm- 2s- 1 at 332 K. We observed that the thickness of the blister skin was greater than anticipated with respect to the peak in the amount of He in W following focused ion beam (FIB) and a scanning electron microscopy (SEM) examination. He bubbles were observed at the position of the crack with transmission electron microscopy (TEM). Yield stress vs displacement per atom (DPA) was calculated. The stress increaseed from the surface to the maximum penetration depth and was shown to reach a minimum at a deeper position than the peak in the amount of implanted He.

The Ti6Al4V ELI (extra-low interstitial) alloy is a prominent choice for use in additively manufactured lattice structures due to its mechanical properties, low density and enhanced osseointegration. However, there is a lack of data on the impact behavior of these Ti6Al4V lattices produced by electron beam powder bed fusion (PBF-EB). Despite the significant progress in the field, printed titanium alloys still suffer from defects such as porosity and unfavorable brittle martensite located within coarse columnar beta grains. Using combined experimental and computational approaches, this study investigated the mechanical properties of PBF-EB Ti6Al4V ELI lattice structures under a wide range of strain rates with a focus on the effects of processing on the structure at the meso(design) and microscopic scales (defects, microstructure). Body-Centered Cubic (BCC) structures and BCC structures with vertical supports (BCCZ) were considered. Hot isostatic pressing was also applied to minimize PBF-EB-induced defects and to improve the microstructure. The effect of such is reported. Inclined struts exhibited poorer surface quality with distinct microstructures when compared to the vertical struts and this was attributed to their thermal histories during printing. Hot isostatic pressing significantly reduced porosity, refined surface quality, and coarsened the microstructure, leading to improved ductility and energy absorption but having little effect on strength. It was found that BCCZ microlattices showed superior mechanical properties over a range of compression velocities, with approximately double the compressive strength, Young's modulus and energy absorption. Both BCCZ and BCC microlattices demonstrated weak strain rate sensitivity.

In-situ TEM annealing and imaging were performed on tungsten samples pre-exposed at either 573 K or 1013 K in a pure helium plasma to a fluence of 1025 m−2. To investigate the thermal evolution of helium bubbles, the samples were imaged for annealing temperatures up to 998 K in steps of 100 K. It is found that annealing temperatures had little effect on the bubble dynamics for the lower temperature exposure of 573 K. Elongated bubble structures remained in the sample throughout the annealing process and there is minimal change in both the average bubble size and cross-sectional number density. However, annealing had a significant effect on the higher temperature exposure of 1013 K. Bubbles were found to be trapped in large clusters and increasing annealing temperature caused a significant increase in the average bubble size accompanied by a decrease in the cross-sectional bubble number density. The experimental observations are supported by theoretical calculations that show a simultaneous increase in the average number of He atoms per bubble in the higher exposure temperature sample but an overall decrease in He atoms retained within bubbles as the annealing temperature increased. These trends in bubble dynamics with annealing temperature suggest that Ostwald ripening dominates, whereby bubble growth is determined by the different pressures of different bubble sizes, rather than through bubble migration and coalescence.

Refractory high‐entropy alloys (RHEAs) show potential for use in extreme environments, such as advanced nuclear reactors, owing to their high melting temperature, and often outstanding combinations of mechanical properties, corrosion resistance, and irradiation‐damage tolerance. This study evaluates the fracture toughness of a TiZrNbHfTa RHEA across different scales and microstructures, with a focus on the impact of He2+‐ion irradiation. Micro‐ and millimeter‐scale specimens with nanocrystalline (NC) microstructures are compared to existing ASTM standard sized coarse‐grained (CG) specimen data, with critical dimensions spanning over three orders of magnitude, from 10 μm to 12 mm. The ASTM standard sized CG specimens exhibit a fracture toughness 41‐fold greater than their NC microscale counterparts (210–5.1 MPa m1/2), while NC millimeter‐scale specimens show a 7.5‐fold higher fracture toughness than NC microscale specimens (38.1–5.1 MPa m1/2). He2+‐ion irradiation leads to a 27% decrease in fracture toughness in the NC microscale specimens. The results highlight the impact of sample dimensional scale, microstructure, and ion irradiation on the fracture toughness of the RHEA, indicating a need for thorough examination of such factors when investigating the mechanical properties of these materials.

A series of FeCr alloys, with 3 at.% - 18 at.% Cr, was irradiated in the Advanced Test Reactor (ATR) at the Idaho National Labs (INL), USA, up to a dose of ~6.7 dpa at a temperature of ~456 °C. Transmission electron microscopy (TEM) samples were extracted using a focused ion beam (FIB) instrument, and the resulting microstructural defects, such as voids, dislocation loops, network dislocations, Cr rich precipitates, etc., were characterized using a TEM. It was found that the size and number density of these defects varied widely over the different alloys with varying Cr content. As expected, there were no Cr rich precipitates in samples with Cr up to 9 %, and they started appearing only in samples with 12 at.% Cr and above. The particle size decreased from about 15 nm at 12 % Cr to 8 nm at 18 % Cr, while the number density increased from ~7e20 /m3 to 6e22 /m3 for the same Cr contents. Grain boundary segregation of Cr, along with a precipitate free zone, was observed in the cases where a boundary was present in the sample. Large voids (>1–2 nm) were almost invisible in the Fe-3at.%Cr sample, while the average void size remained almost constant between 15 and 18 nm for samples with 6–15 at.% Cr and increased slightly at 18 at.% Cr to ~22 nm. Fe-3 %Cr showed a high density of small voids(<2 nm), estimated to be about 1e23/m3. The number density of large voids increased from ~0 at 3 % Cr to a peak of ~6.4e20 /m3 at 12 % Cr, then decreased to about 1.1e20 /m3 at 18 % Cr. The dislocation loops, which appeared in linear arrays in the Fe-9 %Cr sample, were analysed in detail using both invisibility criteria and image simulations using the Oxford University TEMACI software, and it was found that they are most likely to be ½ 〈111〉 type loops on {100} planes. These loops seem shaped like sections of helices in some places and are most likely formed by loops near screw dislocations climbing into a helix and then collapsing into a loop array. Similar dislocation analysis was performed in the other samples as well wherever feasible, and it was found that there is a mixture of ½ < 111 > and [100] dislocation loops.

A series of Fe-Cr alloys, with 3 at.% - 18 at.% Cr, was irradiated in the Advanced Test Reactor (ATR) at the Idaho National Labs (INL), USA, up to a dose of similar to 6.7 dpa at a temperature of similar to 456 degrees C. Transmission electron microscopy (TEM) samples were extracted using a focused ion beam (FIB) instrument, and the resulting microstructural defects, such as voids, dislocation loops, network dislocations, Cr rich precipitates, etc., were characterized using a TEM. It was found that the size and number density of these defects varied widely over the different alloys with varying Cr content. As expected, there were no Cr rich precipitates in samples with Cr up to 9 %, and they started appearing only in samples with 12 at.% Cr and above. The particle size decreased from about 15 nm at 12 % Cr to 8 nm at 18 % Cr, while the number density increased from similar to 7e20 /m(3) to 6e22 /m(3) for the same Cr contents. Grain boundary segregation of Cr, along with a precipitate free zone, was observed in the cases where a boundary was present in the sample. Large voids (>1-2 nm) were almost invisible in the Fe-3at.%Cr sample, while the average void size remained almost constant between 15 and 18 nm for samples with 6-15 at.% Cr and increased slightly at 18 at.% Cr to similar to 22 nm. Fe-3 %Cr showed a high density of small voids(<2 nm), estimated to be about 1e23/m(3). The number density of large voids increased from similar to 0 at 3 % Cr to a peak of similar to 6.4e20 /m(3) at 12 % Cr, then decreased to about 1.1e20 /m(3) at 18 % Cr. The dislocation loops, which appeared in linear arrays in the Fe-9 %Cr sample, were analysed in detail using both invisibility criteria and image simulations using the Oxford University TEMACI software, and it was found that they are most likely to be <1/2> < 111 > type loops on {100} planes. These loops seem shaped like sections of helices in some places and are most likely formed by loops near screw dislocations climbing into a helix and then collapsing into a loop array. Similar dislocation analysis was performed in the other samples as well wherever feasible, and it was found that there is a mixture of 1/2 < 111 > and [100] dislocation loops.

Additive manufacturing (AM) enables the rapid fabrication of complex shapes using engineering materials such as austenitic stainless steels and can imbue them with high irradiation resistance for use in reactor components. This is attributed to their refined grain structure and high grain boundary area. In this study, austenitic stainless steels (type 304) fabricated via direct energy deposition (DED) and powder bed fusion (PBF) techniques were irradiated with 5 MeV He ions to an approximate dose of 0.6 dpa at 300°C; subsequently, they were characterized through electron microscopy and the micro tensile testing. The results revealed that austenitic stainless steels manufactured using AM methods exhibited outstanding mechanical performance. The high performance of austenitic stainless steels fabricated through the DED technique can be attributed to their high tensile strength and excellent ductility elongation. This excellent performance is believed to be caused by the low stacking fault energy and the corresponding martensite formation during deformation. In particular, it was found that better mechanical properties were maintained even after helium irradiation, which is an important result obtained from the micro-tensile test. Even a small variation in the chemical composition and sub-microstructure of AM materials could result in improved irradiation tolerances.

Refractory high‐entropy alloys (RHEAs) are candidate structural materials for nuclear applications due to their promising high‐temperature mechanical performance and irradiation tolerance. However, most body‐centered cubic (BCC) RHEAs form additional phases depending on their thermal history, with few studies assessing their effect on irradiation tolerance. This study characterizes the impact of phase transformations on the room‐temperature irradiation tolerance of a nanocrystalline TiZrNbHfTa RHEA by assessing its microstructure and micromechanical properties before and after thermal treatments between 500 and 800 °C. The alloy demonstrates exceptional irradiation tolerance before and after 500 °C treatments for 1–100 h, which induce BCC to hexagonal close‐packed (HCP) phase transformation, with excellent microstructural stability and minimal irradiation‐induced hardening. Conversely, 800 °C treatment for 1 h forms two major BCC phases and a minor HCP phase, negatively impacting both pre‐ and post‐irradiation mechanical performance and causing significant irradiation‐induced hardening and embrittlement. Additionally, this research identifies a second HCP phase in the 500 °C, 100 h‐treated condition, marking its first mention in the literature. This study emphasizes the importance of assessing temperature and phase formation effects on the irradiation tolerance of RHEAs for future nuclear reactors.

Fusion reactors are designed to operate at extremely high temperatures, which causes the plasma-facing materials to be heated to 500 °C to 1000 °C. Tungsten is one of the target design materials for the plasma-facing diverter components in Tokamak designs, such as ITER, because of its excellent high-temperature strength and creep properties. However, recrystallization due to high temperatures may be detrimental to these superior mechanical properties, while exposure to He plasma has been reported to influence the recrystallization behaviour. This influence is most likely due to the Zener effect caused by He bubbles formed near the surface, which retard the migration of grain boundaries, while at the same time modifying the surface microstructure. This paper reports a study of the effect of plasma exposure at different sample temperatures on the recrystallization behaviour of W at different annealing temperatures. The characterization after plasma exposure and annealing is pursued through a series of post-exposure annealing, followed by scanning electron microscopy (SEM), electron backscatter diffraction (EBSD) characterization and nanoindentation to determine the mechanical properties. Here, it is shown that the hardness is closely related to the recrystallization fraction, and that the plasma exposure at a sample temperature of 300 °C slows down the recrystallization more than at higher sample temperatures of 500 °C and 800 °C. Atomic force microscopy (AFM) was subsequently used to determine any changes in pile-up height around the nanoindents, to probe any indication of changes in hardenability. However, these measurements failed to provide any clear evidence regarding this aspect of mechanical behaviour.

Carbon fibre-reinforced polyetheretherketone (CF-PEEK) composites have gained significant usage across diverse industries like automotive and aerospace due to their desirable characteristics. These properties encompass recyclability, low density, high strength, wear resistance and thermal stability. The components made from CF-PEEK composites for space applications will be subjected to a high radiation environment due to the incoming cosmic rays, comprising protons, α particles, electrons, γ rays, etc., once they escape the Earth’s atmosphere. The ion irradiation of CF-PEEK is accompanied by radiation-induced effects, which drastically change the structure and properties of irradiated material. Since the resistance of CF-PEEK to radiation damage has not been studied extensively, this study aims to understand the effect of high-energy He 2+ ions on the microstructure and properties of CF-PEEK composites manufactured using automated fibre placement (AFP) under different processing conditions. The samples have been radiated with 5 MeV He 2+ ions using an energy degrader wheel to create a layer with relatively uniform damage. Then, were characterized using optical and scanning electron microscopy and their hardness was evaluated using nanoindentation. It was observed that, irradiation increases the hardness of the fibres in all cases. Also, fibre orientation affects the hardness in a statistically significant manner in both unirradiated and irradiated conditions.

In this paper, the deformation mechanisms of a Fe–2.25

GH3535, a candidate Ni-Mo-Cr alloy for GEN IV molten salt reactors, is thermally aged at 650 degrees C for 2000 h and exposed to low density so they ion irradiation. The ion energy is attenuated via means of a degrader wheel to vary between 0.1 to 5MeV, resulting in a near -uniform level of radiation damage over a 10 mu m depth to a dose of 0.25 dpa and helium concentration 1250 appm at room temperature. A focused ion beam (FIB) milling machine was used to fabricate micro-tensile samples (5 x 5 mu m in cross sectional area and 17 mu m in length) from pre- and post- irradiated GH3535 samples. Some of the micro-tensile samples feature a single crystal whilst others feature a grain boundary. Both single crystal and grain boundary micro-tensile results showed helium irradiation had a negligible effect on tensile strength and elongation illustrating the radiation damage resistance of thermally aged GH3535. Additionally, the grain-boundary micro-tensile samples did not fail along the grain boundary before irradiation nor after irradiation which suggests GH3535 continues to fail trans - granularly even after helium ion irradiation. TEM and EBSD analysis were performed on the unirradiated and irradiated aged GH3535 samples and the results of the analysis are discussed here in relation to the micro-tensile results.

Micro-tensile testing has been used to study the response of pure tungsten and two tungsten alloys to helium ion irradiation. Commercially supplied plates of W, W-5Ta and W-5Re were irradiated using 6 MeV helium ions at room temperature. The ion energy was attenuated with an energy spreading device such that a uniform level of damage at 0.6 dpa (and 11,000 appm He) was deposited at the 3–9 µm depth. Focused ion beam milling was used to fabricate dog-bone shaped, micro-tensile samples 5 × 5 µm in cross-sectional area and 17 µm in length from the unirradiated and irradiated samples. All micro-tensile samples were tested at a quasi-static strain rate and the stress–strain curves were analysed to determine the mechanical properties. A close correlation was found between micro-tensile results and the bulk mechanical properties reported in the literature. Comparison between the unirradiated micro-tensile properties of W-5Re and W-5Ta with W showed that, as expected, W-5Re was softer than W whilst W-5Ta had only minor differences in micro-tensile properties compared with W. The micro-tensile results of the irradiated W, W-5Ta and W-5Re showed an increase in strength and an almost complete loss of ductility compared to the unirradiated samples. In comparing micro-tensile results to nanoindentation measurements, it was found that micro-tensile offers comparable level of precision in measurement of irradiation hardening amongst W, W-5Ta and W-5Re. The implications of the results with respect to the future performance of tungsten-based materials in the divertors in fusion reactors are discussed in detail.

The rapid progress in space exploration, mining, and tourism has been fuelled by both public and private sector investments. The latter has led to the need to reduce manufacturing and launch cost of space hardware to create a competitive and sustainable space economy. A major step in making space accessible is to develop affordable power systems for “commercial space” use. Photovoltaics has in the past and will in the future be a key component. Metal halide perovskite solar cells show the greatest potential of all emerging technologies for low‐cost space photovoltaics. They have demonstrated the highest rate of power conversion efficiency improvement. Compared to the triple junction III–V compound semiconductor cells commonly used for space applications, perovskite cells have a higher power to weight ratio and are significantly cheaper to be manufactured. They have high radiation tolerance and can be fabricated onto flexible substrates for expand‐on‐demand solar panels. This paper outlines the major space markets for photovoltaics, and research and development opportunities for perovskite space solar cells in the context of their recent progress.

In this study, alpha particle radiation damage through transmutation of Ni anticipated in a Molten Salt Reactor (MSR) core candidate material is simulated through irradiation of nickel-based alloy, GH3535, with 5 MeV helium ion dose of 1.5 dpa at room temperature via use of a degrader wheel to ensure near-uniform radiation damage throughout a 10 mu m depth. For both the unirradiated and irradiated samples, micro tensile specimens 5 x 5 mu m in cross sectional area and ~ 17 mu m in length are fashioned (using focused ion beam milling) out of regions having single crystals and those including grain boundaries and subsequently tested in tension to failure. Surprisingly, micro-tensile results reveal that voids and displacement damage introduced via irradiation did not embrittle the GH3535 appreciably as there was not a statistically significant difference in their elongation to failure, despite an increase of the Critical Resolved Shear Stress (CRSS) from 200 MPa to 361 MPa. It was also observed that all samples including both unirradiated and irradiated grain boundary micro-tensile samples failed within a grain rather than along the grain boundary proving GH3535 fails trans-granularly even after helium ion irradiation. The implication of the results and comparison with existing literature data are discussed further in this paper. Crown Copyright (C) 2022 Published by Elsevier B.V. All rights reserved.

The effects of hole shape and orientation on the mechanical properties of micro-scale 2D honeycomb structures, fabricated using a focused ion beam equipment, have been investigated using an in situ micro-mechanical testing machine inside the scanning electron microscope . The material used was single-crystal Ni oriented in the < 100 > direction, with the plane of the 2D micro-lattice having a {001} normal direction. The hole shapes explored were hexagonal and circular, while two different orientations of the hexagonal holes were also compared. One of these orientations had a horizontal arm (designated 0° orientation), while the other had a vertical arm (30° orientation) in each hexagon. The results indicate that there is substantial change in strength and ductility depending on the orientation and shape of the holes with respect to the tensile axis. The samples with 30° oriented hexagonal holes had the lowest strength and highest ductility, while the samples with circular holes showed the greatest yield and tensile strength. The samples with the 0° orientated hexagonal holes had much higher strength and lower ductility than the 30° orientated ones. Moreover, the samples with 0° orientated hexagonal holes, which had a similar hole pattern arrangement to the ones with circular holes, had a similar strength to those of the latter type. Thus, it is apparent from this study that the orientation or arrangement of the holes is more important in determining the properties of the 2D microlattice than the shape of the holes. Finite element simulation of the lattice structures utilised the GTN (Gurson, Tvergaard and Needleman) model to evaluate the failure modes under uniaxial tension. The lattice structure has been shown, in a previous paper by the authors, to exhibit composite like behaviour with strength differences in various parts arising from size effects. These size effect variations were incorporated into the model, and a generalised formulation for the GTN parameters was proposed on the basis of one of the experimental configurations and subsequently applied to the other geometries. The models were in good quantitative agreement with the experimental results with accurate representation of the flows stress and failure modes.

The effect of ageing time and temperature on the deformation mechanism and corresponding tensile properties are investigated in a gamma/gamma ' Ni-based superalloy, HAYNES 282. Through a systematic variation in duration (24 h and 216 h) and temperature of ageing (650 degrees C and 760 degrees C), a significant variation in strengthening microstructural features was achieved in the resulting microstructures. While in one case (650 degrees C, 24 h ageing condition), a microstructure with no gamma ' was observed, for the rest of the cases, gamma ' precipitates of varied sizes were observed. In one specific case (760 degrees C, 216 h ageing condition), very fine (4-6 nm) MC carbides were observed along with gamma ' precipitates; in which these tiny MC carbides became the deciding factor for strength over gamma ' precipitates. While common knowledge of smaller gamma ' precipitates giving higher strength could explain the higher strength at 760 degrees C, 24 h ageing condition as compared to 650 degrees C, 216 h ageing condition, the further increase in strength in case of 760 degrees C, 216 h ageing condition, even with much larger gamma ' precipitates was something quite interesting and counterintuitive. This paper elucidates this puzzling observation. The variation of tensile properties of these wide varieties of microstructures will be described in this paper in light of the underlying deformation mechanisms. A change in deformation mechanism from planar slip (microstructure with no gamma ') to twinning (microstructure with gamma ') and Orowan-looping (microstructure with gamma ' along with fine nano carbides) as a function of microstructure explains the difference in strength, ductility, and strain hardening phenomena among the varied ageing conditions.(c) 2022 Elsevier B.V. All rights reserved.