**1. Introduction**

Engineering materials for nuclear applications must be able to withstand extremely harsh service conditions. Among the various areas of nuclear engineering, the fusion reactor system presents a great challenge for materials engineering. The fusion environment, i.e., with high tempera‐ tures and high neutron and particle flux, steps up the degradation and alters the mechanical and thermal behavior of the fusion reactor materials, jeopardizing the dimensional stability and integrity of the materials [1]. The selection of proper materials for certain applications has always been somewhat difficult, and the fusion environment, involving high temperatures and likely radiation damage, makes this task more difficult [1].

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As a result of extensive research, tungsten (W) has emerged as a highly useful plasma-facing material (PFM) [1–3]. Its high melting point, high thermal conductivity, low coefficient of thermal expansion, high sputtering threshold energy, low tritium retention and low neutron activation make W a potential candidate for fusion applications [4–6]. Previously, applications of W were rare and were limited to experimental purposes only in tokamaks due to the formation of high-Z dust, which originates from materials eroding on the surfaces of the plasma-facing components. This dust has detrimental effects on the plasma parameters [7], but research has revealed the feasibility of plasma operations with W [1, 8]. The plasma impurity problem associated with W may be eliminated by ensuring that the energy of the plasma particles remains lower than the sputtering threshold (~700 eV for tritium) [9]. In fusion applications, an increase in the future utilization of W is foreseen [9, 10], as it is considered as a first wall [11] and a divertor surface [1] material for future fusion reactors, as illustrated in **Figure 1**.

**Figure 1.** (a) A schematic cross-section of tokamak and (b) a solid model illustration of divertor (commons.wikime‐ dia.com).

As compared to other high-temperature applications, which require good physical and mechanical properties such as high thermal conductivity, good thermal shock resistance and high-temperature strength, toughness and stiffness [12], fusion reactors impose very complex requirements on plasma-facing materials (PFMs) [1]. While PFMs are assumed to be able to sustain high mechanical, thermal and magnetic loads for prolonged periods of time [8], irradiation effects are particularly important in plasma-facing components [13]. Transmutation and ballistic damage, which alter the composition and microstructure of materials due to the interaction between the materials and the high-energy neutron flux (~14 MeV), reduce the mechanical properties of PFMs. The interaction of neutron flux with materials forms disloca‐ tion loops and clusters of transmutation products from non-equilibrium phases [14].

The energetic ions and neutral atoms in the service environments of fusion reactors cause sputtering erosion in plasma-facing components [1], which is a major concern associated with these materials [8]. The free atoms can be ionized and form contaminants. These contaminants can then be deposited at various locations on the chamber wall. The erosion and re-deposition create new layers which can shorten the service lifetime of the fusion reactor by enhancing tritium retention and morphology variations [8].

Appropriate mechanical, thermal and physical properties are commonly required in plasmafacing components [15]. The hydrogen isotope retention of the materials is also one of the essential considerations when selecting appropriate materials [9]. In the future, the challenges associated with PFMs are expected to increase due to the continuous increase in the thermal loads of upcoming fusion reactors. For example, PFMs may experience very high amounts of localized energy in a very short period of time during transient conditions. Plasma disruption, vertical displacement events (VDEs) and the edge-localized mode (ELM) cause these high localized thermal loads [16], and the inner and outer plates of the divertor may experience loads of ~7–40 MJ/m2 and ~4–25 MJ/m2 due to plasma disruption. Considering VDEs, the energy deposition on the outer wall blanket modules may increase to ~20–30 MJ/m2 in ~0.1/0.3 μs. The ELM, if it is controlled, may impose loads of 0.5 MJ/m2 and 0.3 MJ/m2 on the inner and outer plates, respectively, of the divertor, whereas the ELM if uncontrolled may be more severe as it can impart corresponding loads of 10 MJ/m2 and 6 MJ/m2 on the inner and outer plates of the divertor within ~0.25 to 0.5 μs [17].
