(Invited) Plasma Facing Materials for Future Fusion Reactors Under Extreme Loading Conditions

Abstract

The plasma facing wall of future thermonuclear fusion reactors such as ITER or DEMO has to withstand harsh loading scenarios. This comprises various plasma-wall-interaction processes with high deuterium-, tritium- and helium-ion fluxes (up to ~1024 /m2/s) which will interact with the so-called first wall and the divertor by sputtering, erosion and redeposition processes. These processes are associated with quasi-stationary thermal loads up to 20 MW/m2 combined with short, extremely strong thermal transients up to the GW/m2-range during Edge Localized Modes (ELMs). In addition, irradiation effects resulting from the aforementioned plasma species (D, T, He) and the 14 MeV neutrons have strong impact on the integrity of the wall armour materials. Therefore, also synergistic effects resulting from simultaneous thermal, plasma and neutron wall loads have to be evaluated in complex experiments. Under reactor-relevant conditions, thermally induced defects such as cracking and melting of the plasma facing material (PFM), thermal fatigue damage of the joints between PFM and the heat sink, hydrogen-induced blistering, helium-generated formation of nano-sized tendrils (so-called ‘fuzz’), and neutron induced degradation of the wall armour via reduction of the thermal conductivity, embrittlement, transmutation and activation are the most serious damaging processes. Due to its high melting point (3410°C) and a thermal conductivity of approx. 160 Wm-1K-1 (at 20°C) tungsten is considered as the most reliable material for high heat flux components in future fusion reactors. However, a drawback is its brittle nature; tungsten is ductile and easily workable only above the ductile-brittle-transition-temperature of about 400°C. It also should be noted that tungsten has a strong tendency to recrystallize at high temperatures starting already at about 1200°C. The development of corrosion-resistant tungsten alloys is another important prerequisite for armour materials in future DT-burning fusion reactors. This is of significant importance since tungsten oxides which will form during loss of coolant accidents with air ingress are volatile. This would result in an unacceptable release of highly activated tungsten. For ITER or other large scale confinement experiments alternative solutions based on beryllium and carbon-fibre composites are promising material candidates. Due to their low-Z nature plasma contamination with higher impurity concentrations can be tolerated in the torus. However, erosion and redeposition processes are more serious compared to those with tungsten. A major drawback for the application of beryllium as a plasma facing material is the relatively low melting point (1287°C). In D-T-burning fusion reactors with carbon walls tritium-containing hydrocarbon deposits are formed on all in-vessel components. This will finally result in an inacceptable T-inventory in the fusion reactor. Depending on the selected fibre type and architecture, carbon-fibre reinforced graphites can be manufactured with thermal conductivities equal to or even better than copper (up to ≈ 400 Wm-1K-1). However, such an excellent thermal conductivity will degrade rapidly under the influence of energetic neutrons.