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W-(0, 0.1, 0.3, 0.5) wt.% Hf (mass fraction, wt.%) materials were fabricated by the powder metallurgy method and hot rolling. The microstructure, mechanical properties, and high-temperature stability of alloys with varying compositions were systematically studied. The active element Hf can react with the impurity O segregated at the grain boundary to form fine dispersed HfO2 particles, refining the grains and purifies and strengthening the grain boundary. The average size of the sub-grains in the W-0.3 wt.% Hf alloy is 4.32 μm, and the number density of the in situ-formed second phase is 6.4 × 1017 m−3. The W-0.3 wt.% Hf alloy has excellent mechanical properties in all compositions of alloys. The ultimate tensile strength (UTS) is 1048 ± 17.02 MPa at 100 °C, the ductile fracture occurs at 150 °C, and the total elongation (TE) is 5.91 ± 0.41%. The UTS of the tensile test at 500 °C is 614 ± 7.55 MPa, and the elongation is as high as 43.77 ± 1.54%. However, more Hf addition will increase the size of the second-phase particles and reduce the number density of the second-phase particles, resulting in a decrease in the mechanical properties of the tungsten alloy. The isochronal annealing test shows that the recrystallization temperature of W-Hf alloy is 1400 °C, which is 200 °C higher than rolling pure tungsten. [ABSTRACT FROM AUTHOR]

Helium-induced defect nucleation and accumulation in polycrystalline W and W0.5 wt%ZrC (W0.5ZrC) were studied in-situ using the transmission electron microscopy (TEM) combined with 40 keV He+ irradiation at 800 and 1000°С at the maximum damage level of 1 dpa. Radiation-induced dislocation loops were not observed in the current study. W0.5ZrC was found to be less susceptible to irradiation damage in terms of helium bubble formation and growth, especially at lower temperature (800 °C) when vacancies were less mobile. The ZrC particles present in the W matrix pin the forming helium bubbles via interaction between C atom and neighbouring W atom at vacancies. This reduces the capability of helium to trap a vacancy which is required to form the bubble core and, as a consequence, delays, the bubble nucleation.At 1000 °C, significant bubble growth occurred in both materials and all the present bubbles transitioned from spherical to faceted shape, whereas at 800 °C, the faceted helium bubble population was dominated in W.

ZrC dispersion-strengthened W exhibits high strength/ductility, low ductile-to-brittle transition temperature, and excellent thermal shock resistance, making it a promising candidate plasma-facing material for future fusion devices. In this study, surface modification of 0.5 wt.% ZrC dispersion-strengthened W (WZrC) under low energy and high fluence He plasma irradiation at high temperature was presented. Under the energy of 90 eV and fluence ranging from 6 × 1024 He·m−2–2 × 1026 He·m−2 He irradiation at 920 °C, a typical fuzz nanostructure appeared on the W matrix of WZrC. The thickness of fuzz layer is proportional to the square root of He irradiation fluence. The fuzz showed comparable thickness and structural features to pure W, indicating limited effects of the particle's addition on resistance to high fluence He irradiation at high temperatures. Under continuous He injection, the fuzz would grow extending onto the particle area, making the particle obscured. Besides, the erosion behavior of particles under He plasma irradiation has been investigated, which is thought to be dominated by a sputtering process. Under the He influence of 6 × 1024 He·m−2, only nanopores were observed in the surface region. With fluence increasing to 5 × 1025 He·m−2, the surface became relatively uneven with larger holes. W aggregated in spots and distributed on the surface of the particle, which might be the result of subthreshold sputtering and deposition. When fluence further increased to 2 × 1026 He·m−2, the particles were eroded completely and covered by the extended fuzz, forming cavities. In addition, distinctive layered nanotendrils were observed above the cavities, which were characterized to consist of inner W-riched skeletons and outer Zr-riched layers. It indicates that the layered nanotendrils should be the result of fuzz extension combined with particle sputtering/deposition. [ABSTRACT FROM AUTHOR]

The liquid Li coating on the plasma-facing materials (PFM) has been suggested and implemented as a viable and efficient solution to overcome the damage of plasma on PFM in Tokamak devices. However, the internal mechanisms and impacts of tungsten-based PFM are not yet fully understood. Here, we demonstrate the effects of the impurities with multiple Li atoms, the co-doped Li and He atoms, and the self-trapping induced by the He8/He12 and LiHe8/LiHe12 clusters on the mechanical properties and melting point of W, with the aid of the optimized structures, the formation energies, the charge density distribution, and the density of states obtained by the first-principles calculations. The bulk, shear, and Young moduli, together with the anisotropy factor and Poisson ratio, are calculated for each structure by using the calculated elastic coefficients to understand the effects of impurities on the mechanical properties of W. The results show that the substitutional site in the W matrix is more likely to be occupied by a single Li impurity atom and the impurity He atoms prefer the octahedral interstitial, at the same time, the additional Li impurity atoms are more likely to be located at tetrahedral interstitial sites. The charge density distributions and density of states exhibit that the bond energies involving the highly doped Li or He atoms decrease, which leads to a substantial decrease of elastic coefficients and then obviously reduces the melting points, and the shear and Young moduli, although the impurity of single Li and the self-trapping of the single He8/He12/LiHe8/LiHe12 clusters can maintain these physical quantities as PFM. It implies that the impurity concentration of Li or Li-He in the W matrix should be strictly controlled in the PFM application. Therefore, the present investigations can provide a helpful guide for developing PFM with tungsten.

W-(0, 0.1, 0.3, 0.5) wt.% Hf (mass fraction, wt.%) materials were fabricated by the powder metallurgy method and hot rolling. The microstructure, mechanical properties, and high-temperature stability of alloys with varying compositions were systematically studied. The active element Hf can react with the impurity O segregated at the grain boundary to form fine dispersed HfO2 particles, refining the grains and purifies and strengthening the grain boundary. The average size of the sub-grains in the W-0.3 wt.% Hf alloy is 4.32 μm, and the number density of the in situ-formed second phase is 6.4 × 1017 m−3. The W-0.3 wt.% Hf alloy has excellent mechanical properties in all compositions of alloys. The ultimate tensile strength (UTS) is 1048 ± 17.02 MPa at 100 °C, the ductile fracture occurs at 150 °C, and the total elongation (TE) is 5.91 ± 0.41%. The UTS of the tensile test at 500 °C is 614 ± 7.55 MPa, and the elongation is as high as 43.77 ± 1.54%. However, more Hf addition will increase the size of the second-phase particles and reduce the number density of the second-phase particles, resulting in a decrease in the mechanical properties of the tungsten alloy. The isochronal annealing test shows that the recrystallization temperature of W-Hf alloy is 1400 °C, which is 200 °C higher than rolling pure tungsten.

The angle of incidence dependence of the sputtering energy thresholds (Eth) for tungsten targets irradiated by Helium, Tritium, and Deuterium ions was studied using two different simulation programs (RDS-BASIC and SDTrimSP). For each of the studied irradiation systems, the sputtering energy threshold was found to slightly decrease (3 to 6 eV) by increasing the angle of incidence of the irradiating ion beam. This decrease was found to vary from one irradiation system to another and give a minimum Eth value which was found to occur at a certain angle range for each system. Departing from this minimal Eth angles toward higher angles was found to result in an abrupt and more dramatic increase in the Eth values which can exceed the normal incident Eth values, especially at grazing angles. Our results were also compared and discussed in light of some existing theories.

Tritium (T) is a costly radioactive element that, when retained in plasma-facing materials (PFMs), not only results in fuel loss but also raises issues of radioactive contamination. Hydrogen isotope exchange is a potential method for T removal in future fusion devices. However, in the nuclear fusion environment, PFMs will be subjected to low-energy and high-flux helium (He) plasma irradiation, forming a He bubble layer near the material surface. This greatly impacts the diffusion and retention behavior of hydrogen isotopes in PFMs. In this work, a multi-component hydrogen isotopes exchange model was developed considering the presence of He bubble layers. The influence of He bubbles location on hydrogen isotope retention and exchange behavior was investigated. It is found that when the hydrogen (H) particles implanted outside the He bubble layer, H entering the material would diffuse across the He bubble layer, resulting in a decrease of H amount entering the bulk. In other words, a thicker He bubble layer leads to lower H retention. As to isotope exchange, when there are deuterium (D) retained in the material, implanted H has a possibility to exchange with D, making D release from the materials. However, the D atoms exchanged out by H also need to diffuse across the He bubble layer, further reducing the D release rate. The results showed the barrier effect of the He bubble layer can have two distinct effects on hydrogen isotope exchange. One is that with enhancing the effect of the He bubble layer, the number of H entering the bulk will increase, resulting in an increase in the D removal. The other effect is that with strengthening the barrier effect of the He bubble layer, the D atoms exchanged out will be more inclined to diffuse into the bulk, leading to a decrease in the D removal. As a result, the D removal efficiency exhibits a peak considering the barrier effect of He bubble.

Tungsten fiber-reinforced tungsten composite (Wf/W) material is considered a plasma-facing material (PFM) with good application prospects. Commercial tungsten wire (fiber) prepared through forging and drawing processes has excellent mechanical properties, as well as a very high recrystallization temperature due to the unique texture of it grain structure. Commercial tungsten fiber is the most proper reinforcement for Wf/W. The change in the properties of tungsten fiber because of neutron irradiation makes it inevitable for Wf/W to be used as PFMs. However, there is very little research on the change in the properties of tungsten fiber caused by neutron irradiation. In this work, we used heavy ion irradiation to simulate the displacement damage generated by neutron irradiation to explore the alteration of the properties of a commercial tungsten fiber caused by neutron irradiation. The investigated subject was tungsten fiber with a diameter of 300 μm. The irradiation source was 7.5 MeV Au2+, which generated a maximum displacement damage of 60 dpa at a depth of 400 nm, and the irradiation influenced depth was 1000 nm. Because of the irradiation, significant lattice distortion occurred within the tungsten fiber, resulting in the transition from (110) texture to (100) texture at the fiber's cross-section. The results of the Schmidt factor and Taylor factor analysis indicate a decrease in the plasticity of the tungsten fiber after irradiation, but it did not completely lose its plasticity. The results of the nanoindentation test confirmed the radiation hardening. After irradiation, the hardness of the tungsten fiber increased by approximately 0.33 GPa, but this increase was relatively small compared to other tungsten-based materials. This indicates that commercial tungsten fiber is a low-cost and highly reliable reinforcement material for Wf/W composite materials. [ABSTRACT FROM AUTHOR]

Simulations for low-energy irradiation of nuclear Fusion plasma’s ion species on several plasma-facing materials were performed. This study is devoted to determining the minimum incident ion energy to cause sputtering or what is known as the “sputtering energy threshold” Eth. The simulations have been performed on various materials such as tungsten, graphite, and beryllium irradiated with various light ionic species such as helium, deuterium, and tritium. Two different Monte-Carlo-based simulation programs were used to perform this task. One of the latest versions of the state-of-the-art SDTrimSP simulation program was used in this study and complemented by the RDS-BASIC simulation program which was specifically designed to perform reasonably accurate and detailed calculations for the low–energy ions irradiation. The later simulation program also adopts a special and more realistic treatment accounting for the reduction in the atomic displacement threshold in the top surface monolayer. The simulation results were also compared with experimental data and theoretical data fittings.

With great power comes great challenges. For nuclear fusion, the holy grail of energy, taming the flame of a miniature star in a solid container remains one of the most fundamental challenges. A tungsten armour for the solid container marks a temporary triumphâ€"a solution adopted by the world’s largest fusion experiment, ITERâ€"but may be insufficient for future challenges. High-entropy alloys (HEAs), which are characteristic of a massive compositional space, may bring new solutions. Here, we explore their potential as plasma-facing materials (PFMs) with a prototype W57Ta21V11Ti8Cr3 HEA that was designed by exploiting the natural-mixing tendency among low-activation refractory elements. Revealed by x-ray diffraction analysis and energy-dispersive x-ray spectroscopy, it predominantly consists of a single bcc-phase but with V, Ti, and Cr segregation to grain boundaries and at precipitates. Its yield strength improves ∼60% at room temperature and oxidation rate reduces ∼6 times at 1273 K, compared with conventionally used W. The Tiâ€"Vâ€"Cr rich segregations and the formed CrTaO4 compound contribute to the improved oxidation resistance. However, the Tiâ€"Vâ€"Cr rich segregations, along with the decreasing valence-electron concentration of the matrix by the addition of Ta, V and Ti elements, considerably increase the deuterium retention of the W57Ta21V11Ti8Cr3 HEA to ∼675 multiples of recrystallized W. Moreover, its thermal conductivity decreases, being ∼40% of W at 973 K. However, the maximum tolerable steady-state heat load is still ∼84% of W because of its exceedingly high yield strength at elevated temperatures. Overall, despite being preliminary, we expect HEAs to play an important role in the development of advanced PFMs, for their disadvantages are likely to be compensated by their advantages or be overcome by composition optimization. [ABSTRACT FROM AUTHOR]

• Studies of co-deposition of plasma-facing materials. • Plasma-wall interactions and material erosion. • Molybdenum, tungsten, and carbon plasma-facing components. • Characterization of surface properties of plasma-exposed samples. Material erosion, migration, mixing, dust formation, and co-deposition were investigated on plasma-facing materials in the experimental advanced superconducting tokamak (EAST). The test samples (TSs) of molybdenum (Mo), tungsten (W), and carbon (C) were irradiated during EAST operations. All exposed TS surfaces are modified by erosion and deposition processes resulting in the formation of thin layers containing a mixture of PFCs (Mo, W, Cu, Cr, Fe) and light elements (Li, C, Ca, N, O). The majority of chemical species are in a layer of thickness <650–900 nm. The particle size and structure of co-deposition were observed on each TS. Depth profiling by laser-induced breakdown spectroscopy (LIBS) suggests the presence of local and global impurities on the TSs resulting from the plasma-wall interaction (PWI). The processes involved in the PWI that caused erosion of EAST materials are heat flux and high energy excited impurity particles. The material is sputtered from the first wall plasma-facing components. Erosion and deposition processes occur simultaneously in the EAST. The migrated impurities get deposited globally on remote surfaces. A crack was observed in one of the W TSs due to a high heat load. The spectral intensity of co-deposition and substrate continuously changed during successive laser shots. [ABSTRACT FROM AUTHOR]

Physical sputtering caused by particle bombardment is believed to be the main erosion mechanism of W materials in fusion devices, in which W atoms are the sputtering products. However, the tungsten deuteride molecule (WD) spectra have been observed in both TEXTOR and ASDEX Upgrade, which was believed to be the product of chemically assisted physical sputtering (CAPS), a new sputtering mechanism that has been proposed in recent years. In this paper, we report the spectroscopic observation of WD molecules in the EAST W divertor. The behaviors of WD molecules sputtering are compared with W atoms sputtering via the spectral measurements of the ro-vibrational band emission of WD 6Π → 6Σ+ in the spectral range between 673 nm and 678 nm and the WI line emission at 400.9 nm. The physical sputtering characters and chemical sputtering characters of WD molecule sputtering were confirmed in EAST. The measurements in EAST showed that there is an energy threshold for WD molecules sputtering and that the sputtering energy threshold of WD molecules is smaller than that of W atoms. Furthermore, the dependence of WD molecule sputtering efficiency (the absolute WD photon flux (PhWD) normalized to the particle flux (Γion) reaching the target with a constant Te) on impact energy and heat flux were studied, presenting significantly different behaviors compared with W atom sputtering. In addition, it is found that the decrease of WD sputtering efficiency with the heat flux is accompanied by the rise of the photon flux of Dδ (410.06 nm) normalized to the particle flux (Γion) at the divertor target, which may imply the enhanced deuterium desorption at the W surface. The further increase of heat flux hitting the target surface could elevate the surface temperature and benefit the deuterium release from the surface, which may influence the formation of WD molecules at the surface layer.

Tungsten fiber-reinforced tungsten composite (Wf/W) material is considered a plasma-facing material (PFM) with good application prospects. Commercial tungsten wire (fiber) prepared through forging and drawing processes has excellent mechanical properties, as well as a very high recrystallization temperature due to the unique texture of it grain structure. Commercial tungsten fiber is the most proper reinforcement for Wf/W. The change in the properties of tungsten fiber because of neutron irradiation makes it inevitable for Wf/W to be used as PFMs. However, there is very little research on the change in the properties of tungsten fiber caused by neutron irradiation. In this work, we used heavy ion irradiation to simulate the displacement damage generated by neutron irradiation to explore the alteration of the properties of a commercial tungsten fiber caused by neutron irradiation. The investigated subject was tungsten fiber with a diameter of 300 μm. The irradiation source was 7.5 MeV Au2+, which generated a maximum displacement damage of 60 dpa at a depth of 400 nm, and the irradiation influenced depth was 1000 nm. Because of the irradiation, significant lattice distortion occurred within the tungsten fiber, resulting in the transition from (110) texture to (100) texture at the fiber’s cross-section. The results of the Schmidt factor and Taylor factor analysis indicate a decrease in the plasticity of the tungsten fiber after irradiation, but it did not completely lose its plasticity. The results of the nanoindentation test confirmed the radiation hardening. After irradiation, the hardness of the tungsten fiber increased by approximately 0.33 GPa, but this increase was relatively small compared to other tungsten-based materials. This indicates that commercial tungsten fiber is a low-cost and highly reliable reinforcement material for Wf/W composite materials.

Due to the excellent resistance to radiation damage, high-entropy alloys (HEAs) have been considered as important candidates for nuclear materials. However, to realize the great potential of HEAs in fusion energy, it is important to consider the hydrogen (H) behaviors, which remain to be elucidated. Here, we systematically investigate the dissolution, diffusion and desorption of H in equiatomic WTaVCr using the first-principles calculations. It is found that the H solution energy in WTaVCr is lower than that in pure W, which can be clarified by the H affinity of metallic elements and the available volume of interstitial sites. Accordingly, a possible diffusion path of interstitial H is selected, containing 26 metastable sites, and the corresponding energy barrier is 0.46 eV. Besides, the maximum trapping energy and accommodation number of H at a mono-vacancy in WTaVCr is only 0.30 eV and 3 ∼ 5 H, respectively, which is much lower than that in pure W (1.28 eV and 12 H). More importantly, because of the low trapping energy and moderate diffusion energy barrier, the desorption temperature of H from a mono-vacancy is lower than the room temperature, implying that vacancies are not efficient trapping centers for interstitial H atoms in WTaVCr. Therefore, although the equilibrium interstitial H concentration in WTaVCr is higher than that in pure W, H retention at high temperatures and H-induced blistering are suppressed due to the weak H-defect interactions. Our results will provide a good reference for understanding the H behaviors in HEAs. [ABSTRACT FROM AUTHOR]

Tungsten (W) is one of best candidate metal for plasma-facing materials (PFM), especially due to its high melting temperature and neutron absorption capability. However, converting W into bulk PFM is hard because of its high melting point. This problem can be solved by adding metallic sintering aids with low melting points. In this study, W matrix with 1 wt% Ni aid was reinforced by adding NdB 6 particles (1, 5, and 10 wt%). It can be introduced as a novel potential PFM, thanks to its low volatility and high neutron absorbability. The ceramic and composite powders produced via mechanochemical synthesis and mechanical alloying were examined in terms of composition, particle size, crystallite size, and lattice strain. Samples sintered via pressureless sintering (PS) and spark plasma sintering (SPS) were microstructurally analyzed by using an X-ray diffractometer (XRD), a scanning electron microscope (SEM) attached with an energy dispersive spectroscope (EDS), and mechanically analyzed in terms of microhardness and wear behavior. Based on the results, W 2 B and WB phases emerged in the SPS'ed W1Ni-5NdB 6 and PS'ed./SPS'ed W1Ni-10NdB 6 composites. SPS'ed W1Ni-10NdB 6 composite had the highest hardness value and the lowest specific wear rate. The SPS'ed W1Ni-5NdB 6 composite showed fewer surface damages and higher irradiation resistance as compared with other samples after exposure of He+ irradiation. [Display omitted] • NdB 6 ceramic particulates were reinforced to W1Ni matrix with mechanical alloying. • W1Ni-NdB 6 composites were sintered with pressureless and spark plasma methods. • W1Ni-10NdB 6 composite had the highest hardness value and lowest specific wear rate. • W1Ni-5NdB 6 composite showed less surface damage and high resistance to irradiation. • Potential plasma-facing material was produced by adding NdB 6 particulates into W. [ABSTRACT FROM AUTHOR]

In fusion reactors, plasma-facing materials will be exposed to low-energy and high-flux plasma of the hydrogen isotopes fuel, i.e., deuterium (D) and tritium (T), along with helium (He), the product of nuclear fusion reactions. He irradiation is likely to trigger the formation of He bubbles beneath the material surface, exerting complex effects on the migration patterns of hydrogen isotopes. In this work, molecular dynamics simulations were employed to investigate the influence of nano-sized He bubbles on the D migration behavior in tungsten (W) during the continuous implantation and diffusion processes, which is more consistent with the fusion environment for plasma-facing materials. It is found that the influence of He bubbles on D migration behavior can be classified into barrier and trapping effects. These effects show a competitive relationship at different processes. During the implantation process, the barrier effect takes precedence over the trapping effect, leading to the reduced D retention. Conversely, during the diffusion process, the trapping effect dominates over the barrier effect, resulting in increased D retention. Due to the surface of He bubbles serves as the trapping sites for hydrogen isotopes, augmenting the size and density of He bubbles notably enriches the strong trapping sites in W. When the location of He bubble is closer to the surface, a higher frequency of interaction events occurs between the implanted D atoms and He bubbles, consequently increasing D retention. Additionally, it was also found that the lattice distortion in pure W is more pronounced than that in He bubble-containing W at high temperatures due to the robust interaction between D and He, which diminishes collisions between D atoms and the W lattice. This study gives a valuable insight into the influence of He bubbles on hydrogen isotopes migration behavior in the fusion environment. [ABSTRACT FROM AUTHOR]

• Conducted MD simulations to analyze the interaction of W dust with W walls at varying velocities (1000–4500 m/s), benchmarked against experimental data. • Observed that increased impact velocities enhance dust absorption by the wall and escalate the ejection of wall material. • Demonstrated excellent agreement between MD results and experimental findings, affirming the reliability of the simulation approach. • Proposed theoretical models to extrapolate the results, facilitating broader applicability and understanding of dust-wall interactions in PFMs. This research investigates the interaction between high-speed tungsten (W) dust and plasma-facing components (PFCs) in fusion reactors, particularly focusing on W walls. Through molecular dynamics (MD) simulations, the study covers a broad spectrum of W dust velocities to evaluate their effect on wall materials with various crystal orientations. We found that high-speed impacts cause considerable damage, including sputtering, degradation, and deformation. The study introduces a damage model derived from experimental and simulation data that reveals the patterns and mechanisms of damage caused by dust impacts. The proposed model significantly improves our understanding of dust-wall interactions and underscores the importance of MD simulations as a reliable technique for exploring such phenomena in the challenging conditions of fusion devices. These insights are crucial to predict and mitigate damage to PFCs, helping to develop more resilient and efficient components. Overall, the research offers valuable knowledge on the atomic-level dynamics of dust impacts and represents a notable advancement in the durability and efficiency of materials used in fusion energy technologies. [Display omitted] [ABSTRACT FROM AUTHOR]

The plasma-facing material is one of the most critical issues in fusion devices, which is important for the improvement of the life time of divertor targets, the control of impurities and the main plasma parameters. The high- energy plasma results in the high particle flux and heat flux deposition on the divertor target, leading to the erosion o f target plates. The re fore, the materials of target plates have an effect on the edge plasma parameters. In this work, the distributions of electron density, electron temperature, and neutral particle density with different divertor materials ( lithium, boron, carbon, and tungsten) are investigated with the help o f the three-dimensional edge transport program EMC3-EIRENE. It is found that the downstream electron density with tungsten targets is significantly lower than that of the other three low-Z target mate ria ls, and the downstream electron temperature of the tungsten targets is higher. Similar distributions of the electron density and temperature are also observed in the mid- p lane region. [ABSTRACT FROM AUTHOR]

Hydrogen glow discharge cleaning (H-GDC) is a routine conditioning procedure for the modern tokamaks and the future fusion machines, including ITER. Due to the low energy of hydrogen ions in glow discharge plasmas, the probability of any considerable damage to the plasma-facing components was mainly ignored. In this work, Tungsten, as the primary candidate for the plasma-facing materials in tokamaks, is considered for studies regarding effects of the H-GDC procedure on the material during a routine vacuum vessel conditioning in Damavand tokamak. After performing routine H-GDC using pure hydrogen, the formation of loosely attached nanostructured bundles (NSBs) on the surface of the tungsten samples was observed. The presence of the NSBs, which can be a source of dust, would be significant due to their possible effects on the functionality of the future fusion plasma devices. The NSBs are observed on the tungsten samples with the surface temperature of <370 K after 2.5–4 hours of hydrogen ion bombardment with incident energy of ≈120 eV and fluence of ≈(2–3.5) × 1022 m–2. The surface modifications of specimens exposed to H‑GDC were examined using a material probe experiment and several surface analysis techniques such as SEM, EDX, XRD, and ERDA. Therefore, the formation of loose nanostructures on the wall of plasma confinement vessels due to H-GDC draws attention to probable damaging effects of this phenomenon upon functionality and outcomes of tokamaks. [ABSTRACT FROM AUTHOR]

To overcome the erosion of the plasma-facing materials (PFMs) inside a fusion reactor, liquid metals such as lithium or tin are being explored as potential PFMs in fusion reactors due to their several advantages such as self-repair or high-power exhaust capability. Lithium can reduce the oxygen and carbon impurity contamination by chemically binding these species to the wall. Thus, experiments of power handling performance with a special liquid metal (Li) divertor module based on the capillary porous system technology were performed in the tokamak COMPASS. The elemental analysis of 14 screw samples located around the vacuum vessel provides information about the migration of evaporated/redeposited liquid elements. The main goal of this work is to characterize and quantify screws containing Li-based coating by LIBS by measuring the quantitative elemental composition. The studied samples were Li-coated Ni-Cr-based screws. The laser ablation was performed in air at atmospheric pressure using a 5 ns pulse Nd:YAG laser at 1064 nm. Suitable Li I, Ca I-II, and Fe I spectral lines were employed for the evaluation of the electron temperature of the plasma using multi-elemental Saha-Boltzmann plots. The electron density was obtained from the Stark broadening of the Hα spectral line. The elemental content of each sample was then obtained by calibration-free LIBS (CF-LIBS).

Mock-ups of the ITER divertor plate are tested with the combination of steady-state plasma and e-beam loads: (1) thermocyclic tests with powerful electron beam load of up to 49 MW/m2 in the e-beam facility; and then (2) subsequent testing in the PLM plasma device with steady-state plasma loads of 0.5–1 MW/m2 and more. Such tests simulate the variable load on divertor plates in the ITER. Tungsten samples are irradiated with helium plasma in experiments on the PLM plasma device with discharge duration of up to 200 min. The tests of ITER-grade tungsten VM-P with the combination of steady-state plasma and thermocyclic e-beam loads led to erosion, cracking, and nanostructured “fuzz” structure growth on the material surface. Post-mortem scanning electron microscopy revealed a stochastic nanostructured surface with dimension of structural elements less than 100 nm. The growth of a nanostructured surface with a “fuzz”-type structure and high porosity is observed. The results of such tests are of interest to estimate the erosion of tungsten in fusion reactors, including ITER, fusion neutron source FNS and DEMO.

The lithium and tin capillary-porous systems (CPSs) were tested with steady-state plasma in the PLM plasma device which is the divertor simulator with plasma parameters relevant to divertor and SOL plasma of tokamaks. The CPS consists of tin/lithium tile fixed between two molybdenum meshs constructed in the module faced to plasma. Steady-state plasma load of 0.1 – 1 MW/m2 on the CPS during more than 200 min was achieved in experiments on PLM which is a modeling far scrapeoff- layer and far zone of divertor plasma of a large tokamak. The heating of the CPS was controlled remotely including biasing technique which allows to regulate evaporated metal influx to plasma. After exposure, the materials of the tin and lithium CPSs were inspected and analyzed with optic and scanning electron micriscopy. Experiments have demonstrated sustainability of the tin and lithium CPSs to the high heat steady state plasma load expected in a large scale tokamak. The effect of evaporated lithium and tin on the plasma transport/radiation was studied with spectroscopy to evaluate changes of plasma properties and plasma-surface interaction.

Self-passivating tungsten based alloys for the first wall armor of future fusion reactors are expected to provide an important safety advantage compare to pure tungsten in case of a loss-of-coolant accident with simultaneous air ingress, due to the formation of a stable protective scale at high temperatures in presence of oxygen preventing the formation of volatile and radioactive WO3. In this work, Zr is added to self-passivating W-10Cr-0.5Y alloy, manufactured by mechanical alloying and HIP, in view of improving its mechanical strength and thus, its thermal shock resistance. The as-HIPed W-10Cr-0.5Y-0.5Zr exhibits a nanocrystalline microstructure with the presence of an extremely fine nanoparticle dispersion. After heat treatment at 1555 °C for 1.5 h, the grain size growths from less than 100 nm to 620 nm and nanoparticles are present both at the grain boundaries and inside the grains. Oxidation tests at 1000 °C revealed that the alloy with Zr exhibits also a strong oxidation reduction compared to pure W. The long-term oxidation rate is similar to that of the alloy without Zr. Under thermal shock loading simulating 1000 ELM-like pulses at the divertor, the heat treated Zr-containing alloy did not present any damage.

In the future fusion devices, ELMs-induced transient heat flux may lead to the surface cracking of tungsten (W) based plasma-facing materials (PFMs). In theory, the cracking is related to the material fracture toughness and the thermal stress-strain caused by transient heat flux. In this paper, a finite element model was successfully built to realize a theoretical semi infinite space. The temperature and stress-strain distribution as well as evolution of W during a single heating-cooling cycle of transient heat flux were simulated and analyzed. It showed that the generation of plastic deformation during the brittle temperature range between room temperature and DBTT (ductile to brittle transition temperature, ∼400 °C) caused the cracking of W during the cooling phase. The cracking threshold for W under transient heat flux was successfully obtained by finite element analysis, to some extent, in consistent with the similar experimental results. Both the heat flux factors (FHF = P·t0.5) and the maximum surface temperatures at cracking thresholds were almost invariant for the transient heat fluxes with different pulse widths and temporal distributions. This method not only identified the theoretical conclusion but also obtained the detail values for W with actual temperature-dependent properties.

Plasma sprayed W-based coatings have the potential application as joining interlayers for the plasma-facing components of future fusion reactors. Particularly in the form of tungsten-steel functionally graded materials (FGMs), they would reduce the stress concentration between the W-based plasma-facing armor and steel-based construction. For mechanical integrity, good adhesion is essential. For this, conventional grit blasting treatment of the substrate may not be sufficient. Therefore, alternative treatments capable of reaching significantly higher roughness without damaging the substrate are sought. In this study, the effects of laser texturing on the adhesion of plasma sprayed W and W-steel FGMs on W and steel substrates are investigated. A variety of surface profiles were achieved by controlled laser ablation in several types of patterns. Their geometric features were systematically varied and measured, and the extent of their filling by the coating material was evaluated. On the most promising patterns, coating adhesion tests were performed in shear loading mode and the best performing patterns were identified. Significant improvement over grit-blasted surface was obtained on patterned surfaces with a high effective contact area, promoting mechanical interlocking. On steel substrates, generally higher adhesion was observed, surpassing already the coating cohesive strength; this is attributed to a combination of mechanical interlocking and metallurgical bonding. [ABSTRACT FROM AUTHOR]

• The crack tip is blunted and the crack propagation rate is reduced in W at high temperatures. • The crack propagation rate decreases as the strain rate increased due to the emission of dislocations from the crack tip in W, regardless of the presence of H. • H atoms in the vicinity of the crack tip can hinder the propagation of the crack at low temperatures, while they can promote the propagation of the crack at high temperatures. Tungsten (W) is regarded as a viable choice for plasma-facing materials in nuclear fusion reactors. However, its mechanical properties are significantly degraded by hydrogen (H) atoms during irradiation, of which the mechanism is still elusive. In this study, we conduct molecular dynamics (MD) simulations to study the impact of H atoms on the propagation of a crack in single crystal W. The results show that the propagation rate of the crack slows down with increasing temperature. This is due to the enhanced plastic deformation, leading to blunting of the crack tip. A pre-existing crack in W is then considered at various temperatures and uniaxial applied tensile strain conditions. The propagation rate of the crack decreases with the increase of the applied tensile strain rate. This phenomenon occurs due to the relaxation of the stress around the crack tip following the emission of the dislocation at high strain rates. After introducing H atoms, it can be observed that at low temperatures, H impedes the propagation of the crack, while at high temperatures, H promotes it. This is primarily due to the formation of voids at the slip traces of dislocations and the reduction in surface energy. Additionally, the crack tip becomes blunted and its propagation rate decreases with increasing strain rate. These results indicate that providing sufficient time for H atoms to migrate is a key factor affecting the mechanical properties of W. The current results provide valuable insights into understanding the interaction mechanism of a crack and H atoms in W. [ABSTRACT FROM AUTHOR]

The paper is the result of investigations the authors of initial stages of hydrogen isotope retention and migration, formation of radiation damages of lattice in the constructive materials irradiated wich H+/D+ ions. The studies methods were thermal desorption spectroscopy (TDS) combined with electron microscopy (TEM and SEM) and electrical resistance measurements. Deuterium retention was studied as a function of ion fluence, implantation temperature, incident ions energy and pre-implantation by He+ ions. Special attention was given to the investigation of thin films of constructive metals irradiated with D+ ions with the range greater than the film thickness. Also the trapping of deuterium in thin metallic films was investigated at D+ ions implantation (the range was comparable with the film thickness). Possible mechanisms are proposed to describe the observed processes.

This study combines both transmutational changes and accelerated damage by simulation of irradiating tungsten with 16, 30 and 45 MeV protons. Comparative results indicate 30 MeV to be most optimal amongst the three, for uniformity of combined damage. Finally, the results were compared against fission reactor calculations and DEMO relevant compositional changes. Using 30 MeV protons, for damages of 1 dpa equivalent, the rhenium content is calculated as 401 appm. This compares well against appm induced within a DEMO reactor and is better than estimated 50,000 appm for a fission reactor. Using higher energy protons, the recoils are expected to behave similar to neutron displacement damage creation. Additionally, the study suggests near constant and comparable damage for sample thickness’s upto 500 µm.

W, W-TaC, and W-TiC materials were subjected to heat–load tests in an electron beam facility (10keV, 8kW) at 100 pulses. After heat loading, severe cracks and plastic deformation were detected on the surface of pure W materials. However, plastic deformation was the primary change on the surfaces of W-TaC and W-TiC alloys. This phenomenon was due to the second-phase (TaC and TiC) particles dispersed in the W matrix, which strengthened the grain boundaries and prevented crack formation and propagation. In addition, the microhardness of W and W-TiC obviously decreased, whereas that of W-TaC did not change considerably before and after heat loading.

Tungsten-based materials are the most prospective candidates for plasma-facing components of future fusion devices, such as DEMO. W-based composites and graded layers can serve as stress-relieving interlayers for the joints between plasma-facing armor and the cooling or structural parts. Coating/cladding techniques offer the advantages of eliminating the joining step and the ability to coat large areas, even on nonplanar shapes. In this work, W + Cu and W + Ni composites were prepared by pulsed plasma transferred arc (PTA) cladding on several different substrates. Optimization of the process was carried out with respect to powder mixture composition and process parameters like arc current, plasma gas composition, and traverse velocity. Dense claddings of several millimeters thickness and various W content were achieved. Moreover, multilayers with W content gradually varying from 47 to 92% were formed. The structure, compositional profiles, and thermal properties of the claddings were characterized.

The present work proposes a brazing procedure to join a self-passivating tungsten alloy (W-10Cr-0.5Y) with Eurofer. The results indicated the achievement of high continuity W-Eurofer joints using 80Cu-20Ti filler material, which involves operative brazeability. The resulting microstructure of the braze changes considerably compared to the brazing attempts with pure tungsten, which is associated to the reactive character of chromium. A high interaction between molten filler and W base material close to the braze was detected, with preferential grain boundary penetration of Cu-Ti into W alloy. It gave rise to a loss of hardness at the W base material near the joint. Regarding to the strength of the joints, shear strength of ∼90 MPa was obtained. [ABSTRACT FROM AUTHOR]

Compared to pure tungsten, self-passivating tungsten based alloys for the first wall armor of future fusion reactors shall provide a major safety advantage in case of a loss-of-coolant accident with simultaneous air ingress, due to the formation of a stable protective scale at high temperatures in presence of oxygen preventing the formation of volatile and radioactive WO 3. Recently developed W-Cr-Y alloys produced by mechanical alloying and hot isostatic pressing (HIP) exhibit a strong reduction of oxidation rate compared to pure W and high mechanical strength. A heat treatment after HIP at 1555 °C results in a one-phase material with a high thermal shock resistance. Nevertheless, the microstructure is metastable and its thermal stability under operational conditions has to be assessed. In this work results of thermal stability tests on heat treated W-10Cr-0.5Y alloy subjected to temperatures of 650, 700, 800 and 1000 °C for times ranging from 50 to 1000 h are presented. After 1000 h at 650 °C and 100 h at 700 °C no visible change of the microstructure is detected. After 100 h at 1000 °C a complete decomposition takes place with the formation of a uniform, fine-scale mixture of W- and Cr-rich phases, typical for spinodal decomposition. [ABSTRACT FROM AUTHOR]

• Diamond/W matrix composites were prepared by SPS process. • Interdiffusion of tungsten and carbon occurs at the interface between diamond and tungsten matrix. • The relative density and thermal conductivity of coated composites are higher than those of uncoated composites. • A high thermal conductivity is achieved because of the formation of the diamond/WC, W 2 C/W. Future fusion reactors need materials that can withstand high heat loads. This study introduced a production process of high thermal conductivity diamond/W composite materials via spark plasma sintering (SPS). The microstructures, structure of the interface, and thermal properties of the diamond/W composites are investigated. The results indicated that both the interfacial bonding and the thermal conductivity of the composite with a W coated layer are strengthened compared to the non-coated composites. The volume fraction of the diamond particles in the composites is between 10% and 50%. Moreover, thermal conductivity of 30% by volume of coated composites reaches 204 W/(mK), an increase of 18% over pure tungsten. A high thermal conductivity is achieved because of the formation of the diamond/WC、W 2 C/W structure at the interface between diamond and tungsten. [ABSTRACT FROM AUTHOR]

Neutron shielding materials are a critical area of development for nuclear fusion technology. In the compact spherical tokamak, shielding efficiency improvements are particularly needed because of severe space constraints. The most spatially restricted component is the central column shield. It must protect the superconducting magnets from excessive radiation-induced degradation, but also from associated heating, so that energy consumption of the cryogenic systems is kept to an acceptable level. Recent simulations show that tungsten carbide and its composites form an attractive class of neutron-attenuating materials. In this paper, the key structure-property relationships of these materials are assessed, as they relate to generic materials challenges for plasma-facing materials. We first consider some fundamental materials properties of monolithic tungsten carbide including thermal transport, mechanical properties and plasma interaction. WC is found to have generally favourable properties compared to metallic tungsten shields. We then report progress on the development of a new candidate cermet material, WC-FeCr. Recent results on its accident safety, thermo-mechanical properties, and irradiation behaviour are presented. This review also highlights the need for further study, particularly in the areas of irradiation damage and hydrogen trapping. [ABSTRACT FROM AUTHOR]

Highlights • Low-energy and high-flux He+ ions were produced using a radio frequency (RF) ion source in present irradiation experiments. • The effect of O 2 impurity in He plasma on W surface microstructure was studied in the present work. • When R is larger than 5.2 × 10−2, the erosion yield increased obviously, accompanied by the disappearance of nano-fuzz. • The thickness of nano-fuzz layer was reduced with increasing R from 4.0 × 10−6 to 1.8 × 10-2. Abstract The interaction between the impurities (such as carbon, nitrogen, oxygen) and the plasma-facing materials (PFMs) can profoundly influence the performance and service of the PFMs. In this paper, we investigated the influence of oxygen (O 2) impurity in the helium radio frequency (RF) plasma on the surface morphology of polycrystalline tungsten (W) irradiated at the surface temperature of 1450 ± 50 K and the ion energy of 100 eV. The pressure ratio of O 2 to He (R) in RF source varied from 4.0 × 10−6 to 9.0 × 10-2. The total irradiation flux and fluence were ˜1.2 × 1022 ions·m-2·s-1 and ˜1.0 × 1026 ions·m-2, respectively. After He+ irradiation, the specimen surface morphology was observed by scanning electron microscopy. It was found that with increasing R from 4.0 × 10−6 to 9.0 × 10-2 the thickness of nano-fuzz layer at the W surface was thinner and thinner, accompanied by the formation of rod-like structures. The erosion yield increased from 5.2 × 10-4 to 2.3 × 10-2 W/ion when R varied from 4.0 × 10-6 to 9.0 × 10-2. The X-ray diffraction analysis shows that tungsten oxides were formed at the near surface of specimens when R exceeded 1.8 × 10-2. The erosion yield measurements revealed that in addition to surface physical sputtering process, the chemical erosion process could occur due to the interaction between oxygen-containing species and W at the surface. The results indicated that the presence of O 2 impurity in He plasma can obviously affect the surface microstructure of W. The study suggested that O 2 impurity can effectively reduce the growth of nano-fuzz structures. [ABSTRACT FROM AUTHOR]

Tungsten (W) is considered a leading candidate for structural and functional materials in future fusion energy devices. The most attractive properties of tungsten for magnetic and inertial fusion energy reactors are its high melting point, high thermal conductivity, low sputtering yield, and low long-term disposal radioactive footprint. However, tungsten also presents a very low fracture toughness, primarily associated with inter-granular failure and bulk plasticity, limiting its applications. In recent years, several families of tungsten-based alloys have been explored to overcome the aforementioned limitations of pure tungsten. These include tungsten-based high-entropy alloys (W-HEAs) and tungsten-based Self-passivating Metal Alloys with Reduced Thermo-oxidation or "SMART alloys" (W-SAs). Given their proximity to the plasma, it is crucial to understand how the exposure of these candidate plasma-facing materials (PFMs) to the neutron fluxes expected in fusion reactors impacts their material behavior over time. In this work, we present a computational approach that combines inventory codes and first-principles DFT electronic structure calculations to understand the behavior of transmuting tungsten-based PFMs. In particular, we calculate the changes in the chemical composition, production uncertainties, the elastic and ductility properties, and the density of states for five tungsten-based PFMs when exposed to EU-DEMO fusion first wall conditions for ten years. [ABSTRACT FROM AUTHOR]

• For the first time the retention of noble, seeded, and rare isotope tracer gases in plasma-facing components from JET-ILW has been quantitatively determined. • The retrieval of a significant number of specimens from the JET torus after three consecutive ILW campaigns made this study possible. • Helium was found and quantified on the divertor tile 5 and on the limiters, with locally up to 44·1015 3He cm−2 and 12·1015 4He cm−2. • Nitrogen-15 was found on limiter tiles from three distinct poloidal positions (inner wall, outer wall and top of the main chamber) after exposure to ILW-3. • A correlation of N retention with the N seeding rates for each campaign could be shown. Plasma edge cooling, ion cyclotron wall conditioning and disruption mitigation techniques involve massive gas injection (by puffs or pellets) to the torus. A certain fraction remains in plasma-facing components (PFC) due to co-deposition and implantation. An uncontrolled release/desorption of such retained species affects the stability of plasma operation. The aim of this work was to determine the lateral and depth distribution of noble (3He, 4He, Ne, Ar), seeded (N 2 , Ne, Ar) and tracer gases (15N, 18O) in PFC retrieved from the JET tokamak with the ITER-Like Wall (JET-ILW) after three experimental campaigns (ILW-1, ILW-2, ILW-3). Results regarding the retention of those gases are shown as well as a comparison to the deuterium retention in the studied areas. Heavy ion elastic recoil detection analysis was used, being the only technique capable of detection and quantitative assessment of all elements, especially low-Z isotopes. Helium was found on the divertor Tile 5, locally up to 44·1015 3He cm−2 and 12·1015 4He cm−2, and on the limiters as well. Neon was found in two positions on the limiters, with up to 10·1015 Ne cm−2 and the 15N tracer on Be limiters exposed to ILW-3. A correlation of N retention with the N seeding rates for each campaign has also been found. [ABSTRACT FROM AUTHOR]

The experimental study of damage to tungsten (W), molybdenum (Mo), and silicon carbide (SiC) surfaces induced by focused extreme ultraviolet laser radiation (λ ~ 47 nm/~1.5 ns/21–40 µJ) is presented. It was found that W and Mo behaved similarly: during the first shot, the damaged area is covered by melted and re-solidified material, in which circular holes appear – residua of just opened pores/bubbles, from which pressurized gas/vapors escaped. Next cracks and ruptures appear and the W has a tendency to delaminate its surface layer. Contrary, single-crystalline SiC has negligible porosity and sublimates; therefore, no escape of "pressurized" gas and no accompanying effects take place. Moreover, SiC at sublimating temperature decomposes to elements; therefore, the smooth crater morphology can be related to local laser energy density above ablation threshold. When more shots are accumulated, in all three investigated materials, the crater depth increases non-linearly with number of these shots. The surface morphology was investigated by an atomic force microscope, the surface structure was imaged by a scanning electron microscope (SEM), and the structure below the surface was visualized by SEM directed into a trench that is milled by focused ion beam. Additionally, structural changes in SiC were revealed by Raman spectroscopy. [ABSTRACT FROM AUTHOR]

Tungsten/steel composites might play an important role in plasma-facing components, especially in joining of tungsten armor to structural parts made of steel. In this work, Young's moduli and shear moduli of a set of tungsten/P91 steel composites fabricated by spark plasma sintering were determined by resonant ultrasound spectroscopy. It was observed that the sintering temperature and the volume fractions of the individual phases have strong effects on the macroscopic elastic constants. The results were interpreted by means of a finite elements numerical model, showing that the regions of imperfectly bonded tungsten particles appearing in the microstructure of some of the composites act effectively as inclusions with very low elastic stiffness. A good correlation between the number of these regions in the microstructure and the elastic constants was observed. [ABSTRACT FROM AUTHOR]

Self-passivating tungsten based alloys for the first wall armor of future fusion reactors are expected to provide a major safety advantage compared to pure tungsten in case of a loss-of-coolant accident with simultaneous air ingress, due to the formation of a stable protective scale at high temperatures in presence of oxygen which prevents the formation of volatile and radioactive WO 3 . This work analyses the oxidation and thermal shock resistance of W-Cr-Y alloys obtained by mechanical alloying followed by HIPing. Alloys with different Cr and Y contents are produced in fully dense form with nanocrystalline or ultrafine-grained microstructure and a dispersion of Y-rich oxide nanoparticles located mainly at the grain boundaries. Isothermal oxidation experiments confirm an excellent oxidation resistance due to the formation of protective oxide scales at the very surface. These layers mainly consist of Cr 2 O 3 and mixed Y-W and Cr-W oxides. The superior oxidation resistance of these alloys is confirmed by tests simulating accident-like conditions. The thermal conductivity of these alloys at 600–1000 °C is 2–3 times higher than standard Ni-base superalloys like Inconel-718. The material also exhibits outstanding thermal shock resistance: 1000 pulses of 0.19 GW/m 2 power density and 1 ms duration at 400 °C base temperature resulted in no damage, while an increased power density of 0.38 GW/m 2 resulted in the formation of a crack-network and slight surface roughening. An additional thermal treatment at 1550 °C improves slightly the oxidation resistance and significantly the thermal shock resistance of the alloy. [ABSTRACT FROM AUTHOR]

The synthesis and characterization of Ti x WTaVCr alloys for high-temperature and fusion plasma-facing applications are presented in this study. Specimens of Ti x WTaVCr containing 0 to 7 at% Ti were prepared via spark plasma sintering of an elemental powder mixture at 1500 °C. Characterization of sintered samples via X-ray diffraction and scanning electron microscopy with energy dispersive spectroscopy revealed increased formation of a BCC solid solution due to the addition of up to 7 at% Ti. Mechanical characterizations of the Ti x WTaVCr alloys indicate that they are harder and stronger than several reported high-entropy alloys. The alloy that contained 7 at% Ti also exhibited improved oxidation resistance in air at 1000 °C. Its improved strength, hardness, and oxidation resistance suggest future applications for this high-entropy alloy in high-temperature and nuclear applications. [ABSTRACT FROM AUTHOR]

The results of studying the trapping and retention of hydrogen and deuterium in MPG-8 graphite under sequential irradiation with ions of deuterium and hydrogen plasma are presented. The effect of the energy and density of the hydrogen-ion flux, the impurity of oxygen in the plasma, and the temperature of the irradiated samples on a decrease in the deuterium concentration and hydrogen trapping in graphite is studied. Based on the analysis of experimental data, it is suggested that the observed effect of deuterium removal can be explained by the chemical sputtering of graphite. [ABSTRACT FROM AUTHOR]

Tungsten (W) as one of the potential plasma-facing materials (PFMS) in the nuclear fusion reactors would suffer severe irradiation which can result in deterioration of properties. Understanding the irradiation effects on PFMS is thus critical to improve their performances. Here, the He irradiation effects is studied by focusing on microstructures of the W surfaces. As the irradiation fluence increases, surface morphologies of W become rough with abundant nano-structures, namely nanotendrils. Transmission electron microscope (TEM) images show these nanotendrils contain lots of He voids, which reveals good deformation behavior albeit brittle. [ABSTRACT FROM AUTHOR]

• The research status of tungsten as plasma-facing materials (PFMs) is introduced. • The main methods to improve the toughness of tungsten-based materials are summarized. • The effects of different strengthening mechanisms on the properties of materials are compared. • It is pointed out that the existing plasma material irradiation experiment and post-irradiation performance test system need to be improved. • The research trend of tungsten-based composites in the future will focus on composite strengthening. Tungsten-based materials are considered to be one of the most promising plasma-facing materials, but there are still many problems in practical application. Plasma-facing materials are subjected to the multi-field coupling effect of thermal shock and multiple radiations, which requires tungsten-based materials to have not only good mechanical properties but also certain resistance to irradiation. This paper analyzes the damage condition of tungsten and introduces the research on tungsten material strengthening by researchers in recent years from the aspects of solid solution alloying, dispersion strengthening, and fiber strengthening, and the future research trends. So far, fiber strengthening has just started and still needs a lot of research, especially its ability to resist irradiation. [ABSTRACT FROM AUTHOR]

• The formation of surface blisters and deuterium retention in tungsten preliminarily irradiated by argon ions and then exposed to deuterium plasma were studied in this work. • An increase in depth and an increase in the level of argon-induced damage lead to a decrease in the area density of blisters and an increase in deuterium retention. • The reduced surface blisters are mainly in diameter below 10 µm, suggesting that deuterium aggregation at the grain boundaries in the damaged region was suppressed due to the increase in the number of lattice defects created by argon ion pre-irradiations. • Thermal desorption spectra in similar shapes implicate that the type of irradiation-induced defects was rarely affected by the ion energy as well as fluence used in this work. Heavy ions have been widely used to simulate the effect of neutron irradiation-induced crystal defects on the plasma-wall interaction in the future fusion device. In this work, energetic argon ions were utilized to damage tungsten with the different pre-damage distributions. The pre-damage distribution is controlled by tuning ion energy and fluence. After exposure to deuterium plasma, the pre-damaged tungsten targets were analyzed to explore the impact of pre-damage distribution on deuterium-induced surface blistering and retention. Surface observation results reveal that increasing pre-damage depth and peak dpa significantly reduce the surface blister's area density, especially in diameters below ten micrometers. Meanwhile, thermal desorption spectra indicate that the given change in the pre-damage distribution barely change the type of irradiation-induced defects. The role of pre-damage increases in deuterium trapping and aggregation is discussed based on the experimental results. [ABSTRACT FROM AUTHOR]

Materials exposed to plasmas in magnetic confinement nuclear reactors will accumulate radiation-induced defects and energetically implanted gas atoms (from the plasma and transmutations), of which insoluble helium (He) is likely to be the most problematic. The large surface-area-to-volume ratio exhibited by nanoporous materials provides an unsaturable sink with the potential to continuously remove both point defects and He. This property enhances the possibilities for these materials to be tailored for high radiation-damage resistance. In order to explore the potential effect of this on the individual ligaments of nanoporous materials, we present results on the response of tungsten (W) nanoparticles (NPs) to 15 keV He ion irradiation. Tungsten foils and various sizes of NPs were ion irradiated concurrently and imaged in-situ via transmission electron microscopy at 750 °C. Helium bubbles were not observed in NPs with diameters less than 20 nm but did form in larger NPs and the foils. No dislocation loops or black spot damage were observed in any NPs up to 100 nm in diameter but were found to accumulate in the W foils. These results indicate that a nanoporous material, particularly one made up of ligaments with characteristic dimensions of 30 nm or less, is likely to exhibit significant resistance to He accumulation and structural damage and, therefore, be highly tolerant to radiation.