Hong-Yu Chen

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.

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 × 1021 He+/m2 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.