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

Abstract

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.

Enhanced mechanical properties in oxide-dispersion-strengthened alloys achieved via interface segregation of cation dopants

“…Especially, for fusion application, the demanding work environment, including high-energy particle irradiation, high temperature field, high stress field and hydrogen retention, put forward a higher quality requirement for material selection [5]. Recently, ODS tungsten (ODS-W) alloys are considered as the most promising candidate for plasma facing materials (PFMs) due to their outstanding microstructural and mechanical stability, especially at high temperatures [3,6]. Generally, the intergranular oxide second phase particles can effectively pin grain boundary of W matrix, which is beneficial to the hightemperature microstructural stability of ODS-W alloys and then raises their recrystallization temperature [7].…”

Grain size effect on radiation tolerance of nanocrystalline Mo

“…Body-centered cubic (bcc) metals and alloys have attracted much attention in the past decade due to their reduced-activation under irradiation environment [14][15][16][17][18]. The studies on the oxide-dispersion strengthened (ODS) ferrite steels [19][20][21][22] have showed great radiation tolerance since dispersed nanoparticles in the matrix increase the volume fraction of the interfaces which can act as sinks for irradiation-induced defects, especially for He bubbles.…”

“…Formation of voids will lead to swelling and embrittlement, which are the main cause of material failure under irradiation environment [12]. Therefore, how to control the generation of irradiation-induced defects and mitigate the negative effects of He bubbles is the key to design advanced radiation tolerant materials with a balance of mechanical and thermal properties [2,13].Body-centered cubic (bcc) metals and alloys have attracted much attention in the past decade due to their reduced-activation under irradiation environment [14][15][16][17][18]. The studies on the oxide-dispersion strengthened (ODS) ferrite steels [19][20][21][22] have showed great radiation tolerance since dispersed nanoparticles in the matrix increase the volume fraction of the interfaces which can act as sinks for irradiation-induced defects, especially for He bubbles.…”

On the thermal stability of the nanostructured tungsten coatings

“…i.e. previous works suggest that mechanical response of W could be further improved when it is doped with nanoparticles or nanocomposites like K, Y 2 O 3 or Re [3,4,14,15]. Other studies suggest that nanostructured tungsten (NW) exhibit a larger radiation-resistance than coarse-grained tungsten since grain boundaries contribute to (i) annihilation of Frenkel-pairs (self-healing behaviour), and (ii) pinning centres for light species [16][17][18][19][20][21][22].…”