**1. Introduction**

The development of high-performance materials is the major concern for realizing magnetic confinement fusion reactors. The plasma-facing materials (PFMs) work under extreme conditions including high-energy neutron (14.1 MeV) bombardments, severe thermal loads (up to 20 MWm<sup>−</sup><sup>2</sup> ), and high fluences of high-flux (>1021 m−<sup>2</sup> s<sup>−</sup><sup>1</sup> ) and low-energy (<100 eV) hydrogen and helium plasma irradiations [1–3]. The tolerable peak power of the plasma-facing components (PFCs), like the divertor targets, imposes significant restrictions on the design of future fusion reactors such as demonstration (DEMO) reactor or China fusion engineering experiment reactor (CFETR) [4]. Developing advanced materials with improved properties could substantially improve the performance of such PFCs.

Tungsten (W) is a kind of refractory metals that has supreme properties such as high melting temperature, high thermal conductivity, good erosion resistance, low vapor pressure, low swelling, and low tritium retention [1, 5, 6]. These properties are appealing for applications as PFCs in future fusion facilities. However, high brittleness in several regimes including low-temperature embrittlement (relatively high ductile-brittle transition temperature (DBTT)), irradiation embrittlement, and recrystallization embrittlement [7, 8] is one of the main disadvantages, which

limits the engineering application of W alloys, particularly in the nuclear energy fields [3, 9, 10]. The thermal load resistance of material is also intimately linked to the strength and DBTT, because the cracks would initiate when the thermal stress is larger than the ultimate strength of material at temperatures above DBTT or than the yield strength at temperatures below DBTT [11–14]. Currently, the performance of pure W can constrainedly satisfy the ITER servicing condition, but for the CFETR and DEMO with higher working parameters, it is not enough. Therefore, tungsten materials with more excellent mechanical properties, high-temperature stability, and irradiation resistance would be highly desirable for PFCs in future fusion reactors.

For tungsten alloys, decreasing DBTT has been a major goal over recent decades. It is generally understood that the low ductility of tungsten is closely related to their weak grain boundary (GB) cohesion, due to the segregation of interstitial impurities, such as O at GBs [15, 16]. In this sense, methods that purify impurity at GBs should be effective to improve the strength and ductility of tungsten. However at high temperatures (above 1000°C), the strength of pure W decreases significantly [17, 18]. Recent research results indicated that trace active elements, such as Zr, Ti, and Y, in tungsten can react with oxygen and diminish the influence of free oxygen on GBs via forming thermally stable nano-oxide particles and thus purify and strengthen the GBs, raising the stability at high temperature [19–21].

Therefore, several approaches are developed to improve the mechanical properties such as increasing ductility and fracture toughness and decreasing the DBTT. First, tungsten materials with high strength and high thermal stability can be obtained by the dispersing second-phase particles, such as oxides or carbides forming the oxides or carbides and dispersion-strengthened (ODS or CDS) tungsten-based materials [22–30]. Recently, various ODS-W or CDS-W materials with enhanced strength and thermal stability were fabricated and investigated for fusion applications [3, 30]. For example, W-La2O3 and W-Y2O3 showed enhanced strength, high recrystallization temperatures, and high thermal shock resistance [3, 24, 31]. Carbides such as TiC, ZrC, and HfC have much higher melting temperatures than that of the abovementioned oxides and better compatibility with tungsten, which may lead to excellent comprehensive performances in CDS-W. For example, an ultrafine-grained (UFG) W-1.1%TiC fabricated by severe plastic deformation exhibits a very high bending strength up to ~4.4 GPa and appreciable ductility at room temperature (RT) [23]. Zhang et al. [14, 28, 31] developed a bulk W-ZrC alloy with a flexural strength of 2.5 GPa and a strain of 3% at room temperature (RT) and a DBTT of less than 100°C. In addition, the W-ZrC alloy plate can sustain 4.4 MJ/m2 thermal load without any cracks at RT. The low-energy and high-flux plasma irradiation resistance of this W-ZrC alloy is better than that of W-La2O3, ITER grade pure W, and commercial pure W; the hydrogen retention is also lower than that of ITER grade pure W [31]. W fiberreinforced Wf/W composites also provide a good candidate for PFCs. In this chapter, the comprehensive introduction of the abovementioned materials will be reviewed.
