**4.3. Oxidation resistance**

W-Si-C composites produced by pyrolysis at 1800°C exhibit a flexural strength level of approximately 400 MPa. When such a composite is heat-treated at 1700°C, the flexural strength is reduced to ~350 MPa. The hardness and indentation modulus of post-heat-treated W-Si-C were found to be 7.8 GPa and 250 GPa, respectively [11]. The fractural strength is a function of the porosity/density; thus, to develop high-strength W composites, researchers have focused

Yttria particles produce W-based composites with improved high-temperature strength. The Young's modulus of 2%Y2O3/W was found to be 400 GPa, which is higher than those of pure W, 2%Y/W and 1%Y2O3/W [19, 4, 22]. Larger yttria particles and a low porosity level resulted in improved mechanical behavior [22]. However, the Young's modulus of a W-based compo‐ site was found to decrease when a composite with 1wt%TiC was prepared by chemical reduction [13]. The strength may also be improved by producing W composites with 1wt

combustion synthesis followed by centrifugal infiltration show a 12.7% improvement in the

The conventional sintering of W-based composites containing V, Ti, Nb, Ta, Fe, and Ni caused hydrogen embrittlement because the dissolution of H2 dissolves into the composite due to the negative formation energy of the vacancy-hydrogen complexes. Hence, conventional sintering

Fine-grained W-based composites produced by doping with rare-earth oxides such as Y2O3 and La2O3 and carbides such as TiC and ZrC were subjected to transient heat flux tests. It was

nearly 100% higher than conventionally sintered pure W. This promising behavior may be the result of the processing route, i.e., a sol–gel method, heterogeneous precipitation, spray drying,

However, oxide- and carbide-doped W composite samples, when subjected to a thermal shock, showed cracks. However, pure W sintered at 2400°C withstands thermal shocks well [20]. The responses of W composites consisting of 20%-80% porous W and infiltrated by Cu, Al or Si and then exposed to a high-temperature environment have been thoroughly studied. These composites exhibit good strength, conductivity and good melt layer stability at high temper‐ atures. In contrast with pure W, some W-based composites can withstand plasma edge

The thermal conductivity of TiC/W composites and pure W produced by chemical reduction followed by SPS at 1800°C decreases when the temperature is increased from ambient to 827°C. However, the conductivity remains above 120 W/m-K at RT [13]. The effect of the temperature on the thermal conductivity of ODS-W composites prepared by adding 1wt% Pr2O3 was examined, and the behavior of pure W was compared with that of pure W. The conductivity of both materials decreased when the temperature was increased from 25 to 800°C, but the

conductivity of these materials exceeds 150 W/m-K at room temperature [26].

bending strength due the well-bonded W, Cu powder and W fibers [18].


can be sustained by these materials, which is

on low porosity and high density [21].

152 Nuclear Material Performance

%Pr2O3 [26] and 0.25–0.8wt% TiC [5]. Moreover, Wf

in a hydrogen environment is not recommended [21].

hydrogen reduction and ordinary sintering in sequence [5].

observed that a high heat flux of 200 MW/m2

temperatures in excess of 200 eV [82].

**4.2. Thermal properties**

In fusion reactors, W-based materials will be exposed to high-temperature and plasma environments; therefore, their high-temperature oxidation resistance is of great concern in the case of loss of vacuum accident. Attempts to develop W-based composites with enhanced hightemperature oxidative ablation resistance have been made by incorporating ZrNp into a W matrix. 10–30% ZrNp/W samples, when tested using an oxyacetylene torch, showed enhanced oxidative ablation resistance. The rate of ablation was found to decrease with an increase in the ZrNp content. The microstructure of the ablated samples lacked a uniform protective film due to the insufficient amount of ZrNp [83]. Improved oxidative ablation resistance and a reduction in the ablation rate were also found by adding HfC to W up to 30%. A protective layer of HfO2 was found on post-ablated samples [84].

An analysis of the ablation properties of W-based composites suggests that composites with high oxidative ablation resistance can be developed by the incorporation of a W matrix with secondary phase particles having low thermal conductivity, a high melting point, and good oxidation resistance [84].

## **4.4. Behavior of composites in radiation and plasma environments**

Pure W also faces transmutation damage, and in DEMO full-power operations, W may be transformed into W-3.8at%Re-1.4at%Os. These chemical variations alter the mechanical properties of pure W [14]. At the end of the lifetime of the blanket, 5–8 at% of W may be transmuted; moreover, transmutation below 1% has no significant effect on the physical properties of W alloys [2]. The transmutation of W is mainly a function of the neutron's spectrum and fluence [2]. Neutrons in a fusion environment may be absorbed by W with a consequent release of alpha and beta particles and the production of impurities such as Re, Ta, and Os. The change in the chemistry of the material degrades its mechanical (and other) properties [2].

ODS-W composites when prepared by adding Y2O3 also show irradiation hardening when subjected to 24 and 2 MeV Fe and He ions, respectively, up to a damage level of 5 dpa at 300 and 700°C. Irradiation resulted in radiation loops in W and voids in Y2O3 particles, which caused the hardness to increase [4]. No voids or cracking were found on the grain boundaries, demonstrating the capability of the grain boundaries of Y2O3/W composites to accommodate heavy ions [4]. In another study in which a W-yttria composite produced by a powder metallurgical method was subjected to 24 and 2 MeV Fe and He ions, similar changes in the microstructure and hardness were observed [22].

An improvement in the ion irradiation damage resistance of W by doping with TiC was observed by Kurishita et al. In their study, mechanical alloying and HIP were used to produce 0.25–0.8wt%TiC/W composites. He ions at 500°C were used to irradiate TiC/W. The irradiation vacancy defect density was only 1/3 to 1/4 of the density of commercially pure W [5].

The interaction between plasma-facing materials, such as W and its composites, and a single hydrogen beam, a single helium beam and an electron beam results in a rough surface and a fine scale. This modification of the surface can affect the mechanical properties along with the thermal conductivity and tritium retention capabilities of the material [22]. The sputtering threshold energy may be increased from 2.5 KeV to more than 10 KeV by applying an alkali monolayer onto a W composite surface. Alkali metals having a higher mass are more useful for reducing sputter-induced erosion [82].

#### **4.5. Hydrogen retention**

The addition of 0.1wt%TiC to W does not cause significant differences in the deuterium retention behavior [69]. The deuterium retention behavior of a 1.1wt.%TiC/W composite was also investigated by exposing 1.1%TiC/W composite samples to D2 gas at 527–690°C and 100 kPa or by irradiating the 1.1%TiC/W composite via 38 eV/D ions at 527°C. Thermal desorption spectroscopy revealed slightly higher deuterium retention in 1.1%TiC/W as compared to pure W after exposure to D2 gas. However, it was significantly higher after D ion irradiation [61]. The deuterium retention properties of TiC/W when examined by irradiating TiC/W composite samples with 200 eV/D ions at fluence levels ranging from 1×1022 to 1×1024 D/m2 at a temper‐ ature of 37°C or with 38 eV/D ions at fluence levels ranging from 6×1022 to 6×1024 D/m2 at temperatures of 27 and 327°C were examined. At 37 and 27°C, no significant differences in the deuterium retention of TiC/W and pure W were observed. However, at 327°C, deuterium retention was higher in TiC/W than in pure W [70].
