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This study presents a systematic investigation of the synergistic effect of deuterium and helium on the microstructural evolution of W–Y2O3 material during plasma exposure. The effect of single D/He and sequential D/He, He/D plasma exposure on the stability of Y2O3 nanoparticles and the microstructural evolution of the W matrix were investigated using scanning electron microscopy and transmission electron microscopy. The results indicate that Y2O3 particles undergo a significant size change during plasma exposure, and this process was significantly influenced by the synergistic effects of hydrogen and helium. The size evolution of Y2O3 particles is controlled by three mechanisms: the etching-controlled process at the initial stage, followed by deposition-controlled process or mixed-controlled processes. Interestingly, the fuzz structure on the surface of Y2O3 particles is proved to be W instead of Y2O3, which results from the deposition during plasma exposure. Significant synergistic effects of deuterium and helium were observed, in which the strong combination of H and He-vacancy composites suppressed the bubble formation and growth, leading to a different evolutionary dynamic.

A wet-chemical method combined with spark plasma sintering was used to prepare a W–Y 2 O 3 alloy. High-temperature tensile tests and nano-indentation microhardness tests were used to characterize the mechanical properties of the alloy. After He-ion irradiation, fuzz and He bubbles were observed on the irradiated surface. The irradiation embrittlement was reflected by the crack indentations formed during the microhardness tests. A phase transformation from α-W to γ-W was investigated by X-ray diffraction (XRD) and transmission electron microscopy (TEM). Polycrystallization and amorphization were also observed in the irradiation damage layer. The W materials tended to exhibit lattice distortion, amorphization, polycrystallization and phase transformation under He-ion irradiation. The transformation mechanism predicted by the atomic lattice model was consistent with the available experimental observations. These findings clarify the mechanism of the structural transition of W under ion irradiation and provide a clue for identifying materials with greater irradiation resistance.

The extremely harsh environment of the high temperature plasma imposes strict requirements on the construction materials of the first wall in a fusion reactor. In this work, a refractory alloy system, WTaVTiZrx, with low activation and high entropy, was theoretically designed based on semi-empirical formula and produced using a laser cladding method. The effects of Zr proportions on the metallographic microstructure, phase composition, and alloy chemistry of a high-entropy alloy cladding layer were investigated using a metallographic microscope, XRD (X-ray diffraction), SEM (scanning electron microscope), and EDS (energy dispersive spectrometer), respectively. The high-entropy alloys have a single-phase BCC structure, and the cladding layers exhibit a typical dendritic microstructure feature. The evolution of microstructure and mechanical properties of the high-entropy alloys, with respect to annealing temperature, was studied to reveal the performance stability of the alloy at a high temperature. The microstructure of the annealed samples at 900 °C for 5–10 h did not show significant changes compared to the as-cast samples, and the microhardness increased to 988.52 HV, which was higher than that of the as-cast samples (725.08 HV). When annealed at 1100 °C for 5 h, the microstructure remained unchanged, and the microhardness increased. However, after annealing for 10 h, black substances appeared in the microstructure, and the microhardness decreased, but it was still higher than the matrix. When annealed at 1200 °C for 5–10 h, the microhardness did not increase significantly compared to the as-cast samples, and after annealing for 10 h, the microhardness was even lower than that of the as-cast samples. The phase of the high entropy alloy did not change significantly after high-temperature annealing, indicating good phase stability at high temperatures. After annealing for 10 h, the microhardness was lower than that of the as-cast samples. The phase of the high entropy alloy remained unchanged after high-temperature annealing, demonstrating good phase stability at high temperatures.

A common issue encountered in metal powder injection molding is the separation of the powder and binder during the injection process, which can give rise to a number of defects. In order to investigate the phenomenon of phase separation in stainless-steel injected flat parts, a numerical simulation methodology was employed by using the Moldex3D R14.0 finite element software to simulate the injection process. Then, the impact of injection parameters on separation was analyzed by comparing key performance indicators such as powder volumetric concentration and density. Furthermore, the extent of the separation was minimized by the optimization of process parameters. The simulation results indicate that the separation of binder and powder is most severe near the gate, where the binder percentage is the highest. The application of elevated mold temperatures, augmented injection rates and reduced injection temperatures can effectively mitigate the two-phase separation phenomenon and enhance the uniformity of powder distribution. This provides a crucial theoretical foundation and technical support for the enhancement of the quality and performance of metal powder injection products.

Plasma-facing components (PFCs) were simulated by ANSYS, and the influence of gradient layer number and composition distribution index on the distribution of temperature field and stress field was analyzed. The simulation results show that a gradient structure with four gradient layers and a component distribution index of 1 makes the PFC assembly have lower overall temperature and lower thermal stress. Tungsten–copper functionally graded materials (W–Cu FGMs) (W-20 vol% Cu/W-40 vol% Cu/W-60 vol% Cu/W-80 vol% Cu) were fabricated by a selective laser melting (SLM) process based on finite element simulation results. The effects of microstructure on the hardness, internal stresses, thermal conductivity, and thermal expansion coefficient of the W–Cu FGMs were evaluated. The results show that hardness increases from 196 to 1173 HV0.3 with increasing W content. The internal stresses of W-20 vol% Cu, W-40 vol% Cu, W-60 vol% Cu, and W-80 vol% Cu are about 191.7 MPa, 627 MPa, 1049.5 MPa, and 561.9 MPa, respectively. The thermal conductivity of the W–Cu FGM is 23 W/m·K and the thermal diffusion coefficient is 10 mm2/s at 25 °C, and the thermal conductivity rises to 70 W/m·K and the thermal diffusion coefficient rises to 18.5 mm2/s at 800 °C. After 100 thermal shock cycles, the internal defects increased, but the interface between the gradient layers remained well bonded.

In the current work, an analysis of the effects of Y on the radiation hardening and microstructure response of a V-4Cr-4Ti alloy has been conducted after 30 keV D ion irradiation at room temperature using transmission electron microscopy (TEM) and nanoindentation. The results show that the formation of large Y2O3 and small Y2V2O7 nanoparticles was confirmed, indicating that the addition of Y reduces the amount of dissolved oxygen. The addition of Y has been shown to affect the radiation-induced dislocation loops, radiation hardening, and Ti-rich segregation of the V-4Cr-4Ti alloy. With the addition of Y, the mean size of the radiation-induced dislocation loop decreased, which may result from the strong sink strength of the nanoparticle/matrix interface, interactions between Y atoms and SIA clusters, and the strong binding energy of vacancy–oxygen pairs. Some particles with core–shell structures were observed after ion irradiation, where Ti-rich segregations at the nanoparticle/matrix interface were confirmed. These results indicate that Y might promote abnormal segregation. Possible causes for this include the lower interface energy at the particle/matrix interface and the interaction between oxygen and solute atoms.

Combining the advantages of a wet chemical method and spark plasma sintering, carbide-doped materials W-1wt%TiC and W-1wt%ZrC were prepared. Microstructural evolution in W-1wt%TiC and W-1wt%ZrC under irradiation of 5 keV He+ at 600 °C to fluences up to 5.0 × 1021 ions/m2 with ion flux of about 8.8 × 1017 ions/m2s was investigated by transmission electron microscopy (TEM). The dislocation loop number density of W-1wt%TiC was higher than that of W-1wt%ZrC, but the average loop size of the W-1wt%TiC was in average smaller. There were no observable helium bubbles in W-1wt%TiC and W-1wt%ZrC, exhibiting higher radiation resistance to He+ compared to pure W. He+ pre-damaged and undamaged W-1wt%TiC and W-1wt%ZrC samples were irradiated by 5 keV D2+ to estimate the D retention in doped W materials. The irradiation damage impact of He+ on deuterium retention was examined by a method of thermal desorption spectroscopy (TDS). Compared with the undamaged samples, it was illustrated that D2 retention of W-1wt%TiC and W-1wt%ZrC increased after He+ pre-irradiation.

During the spark plasma sintering (SPS) consolidation process, the pressure affects the densification and microstructure evolution of the sintered body. In this paper, the W-Cr-Y-Zr alloy powder was heated to 1000 °C under different applied pressure conditions using spark plasma sintering process, and the effect of pressure on the densification process and microstructure was analyzed. Due to the low sintering temperature, the crystalline size of all the produced W-Cr-Y-Zr alloy is less than 10 nm, which is close to that of the original powders. Cr-rich phase can be detected in the sintered samples due to spinodal decomposition. It is found in this work that the external pressure will increase the contact area between the powder particles, resulting in a higher local pressure at the particle contact, which promotes densification by sliding between the particles under the condition of softening of the particle surface. Additionally, according to the viscous flow theory, the viscous flow activation energy decreases with the increase of pressure. This is because the pressure provides additional driving force to the powder viscous flow process and accelerates the powder shrinkage.

The microstructure of and damage to the upper divertor components in EAST were characterized by using metallography, EBSD, and SEM. Under the synergistic effect of heat load and plasma irradiation, cracking, recrystallization, and interface debonding were found in the components of the upper divertor target. The crack propagates downward from the heat loading surface along the heat flux direction, and the crack propagation mode is an intergranular fracture. The thermal loads deposited on the edge of monoblocks raise the temperature higher than the recrystallization temperature of pure tungsten, and the microstructure changes from being in a rolled state to being recrystallized. Additionally, cracks exist in both recrystallized and rolled areas. EBSD boundary maps show that the range of the recrystallization area is determined via the heat flux distribution. The Cu/CuCrZr interface of the cooling components near the thermal loading area is debonded, and the structural integrity is destroyed.

Tungsten and its alloys are considered to be the most nominated plasma-facing materials in fusion reactors, which will be exposed to enormously rigorous conditions such as thermal load, plasma exposure, and neutron radiation. At present, the research on the behavior of oxide particle-reinforced tungsten-based materials under long-term steady-state heat load and transient thermal shock is insufficient. The purpose of this study is to investigate the performance of yttria particle-reinforced tungsten plates prepared by the wet chemical method under heat loads by means of indirect coupling experiments. An Nd:YAG laser device is used to perform thermal shock events. The surface damage and microstructure evolution of rolled and fully recrystallized samples exposed to laser thermal shock are observed and analyzed. The cracking threshold of the rolled and fully recrystallized samples is about 0.40~0.48 GW/m2; the degree of surface damage of them aggravates with the increased laser power density. What is more, cracks or even melting damage could be observed on the surface and be accelerated by the process of recrystallization, resulting in the degradation of the ability to withstand the thermal shock of the material.