Study on Microstructure and High Temperature Stability of WTaVTiZrx Refractory High Entropy Alloy Prepared by Laser Cladding.

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.

Microstructure and performance of rare earth element-strengthened plasma-facing tungsten material.

Pure W and W-(2%, 5%, 10%) Lu alloys were manufactured via mechanical alloying for 20 h and a spark plasma sintering process at 1,873 K for 2 min. The effects of Lu doping on the microstructure and performance of W were investigated using various techniques. For irradiation performance analysis, thermal desorption spectroscopy (TDS) measurements were performed from room temperature to 1,000 K via infrared irradiation with a heating rate of 1 K/s after implantations of He(+) and D(+) ions. TDS measurements were conducted to investigate D retention behavior. Microhardness was dramatically enhanced, and the density initially increased and then decreased with Lu content. The D retention performance followed the same trend as the density. Second-phase particles identified as Lu2O3 particles were completely distributed over the W grain boundaries and generated an effective grain refinement. Transgranular and intergranular fracture modes were observed on the fracture surface of the sintered W-Lu samples, indicating some improvement of strength and toughness. The amount and distribution of Lu substantially affected the properties of W. Among the investigated alloy compositions, W-5%Lu exhibited the best overall performance.

Effects of zirconium element on the microstructure and deuterium retention of W-Zr/Sc2O3 composites.

Dense W and W-Zr composites reinforced with Sc2O3 particles were produced through powder metallurgy and subsequent spark plasma sintering (SPS) at 1700 °C and 58 MPa. Results showed that the W-1vol.%Zr/2vol.%Sc2O3 composites exhibited optimal performance with the best relative density of up to 98.93% and high Vickers microhardness of approximately 583 Hv. The thermal conductivity of W-Zr/Sc2O3 composites decreased initially and then increased as the Zr content increased. The moderate Zr alloying element could combine well with Sc2O3 particles and W grains and form a solid solution. However, excess Zr element leads to agglomeration in the grain boundaries. W-1vol.%Zr/2vol.%Sc2O3 composite had a good deuterium irradiation resistance very closing to pure tungsten compared with the other Zr element contents of composites. Under 500 K, D2 retention and release of them were similar to those of commercial tungsten, even lower between 400 K to 450 K. Pre-irradiation with 5 keV-He(+) ions to a fluence of 1 × 10(21) He(+)/m(2) resulted in an increase in deuterium retention (deuterium was implanted after He(+) irradiation), thereby shifting the desorption peak to a high temperature from 550 K to 650 K for the W-1vol.%Zr/2vol.%Sc2O3 composite.

Chemical Synthesis and Oxide Dispersion Properties of Strengthened Tungsten via Spark Plasma Sintering.

Highly uniform oxide dispersion-strengthened materials W-1 wt % Nd2O3 and W-1 wt % CeO2 were successfully fabricated via a novel wet chemical method followed by hydrogen reduction. The powders were consolidated by spark plasma sintering at 1700 °C to suppress grain growth. The samples were characterized by performing field emission scanning electron microscopy and transmission electron microscopy analyses, Vickers microhardness measurements, thermal conductivity, and tensile testing. The oxide particles were dispersed at the tungsten grain boundaries and within the grains. The thermal conductivity of the samples at room temperature exceeded 140 W/m·K. The tensile tests indicated that W-1 wt % CeO2 exhibited a ductile-brittle transition temperature between 500 °C and 550 °C, which was a lower range than that for W-1 wt % Nd2O3. Surface topography and Vickers microhardness analyses were conducted before and after irradiations with 50 eV He ions at a fluence of 1 × 1022 m-2 for 1 h in the large-powder material irradiation experiment system. The grain boundaries of the irradiated area became more evident than that of the unirradiated area for both samples. Irradiation hardening was recognized for the W-1 wt % Nd2O3 and W-1 wt % CeO2 samples.

Mechanical properties and microstructural change of W-Y2O3 alloy under helium irradiation.

A wet-chemical method combined with spark plasma sintering was used to prepare a W-Y2O3 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.