Blister formation and D retention in W have been investigated for low energy (∼55 ± 15 eV), high flux (∼1022 m−2 s−1), high fluence (⩽4.5 × 1026 m−2) ion bombardment at moderate temperature (∼573 K) in mixed species D+He plasmas in the linear divertor plasma simulator PISCES-A. The amount of D retained in W is found to decrease significantly when compared with that in W exposed to pure D plasmas, as measured with high resolution thermal desorption spectroscopy. Scanning electron microscopy observations reveal the suppression of the blisters, a surface feature known to drive up retention, in the D + He mixture plasma exposed W samples. Reduced D retention is accompanied by the formation of nano-sized high density He bubbles in the near surface, observed with a transmission electron microscope (TEM). It is believed that the nano-bubbles act as a diffusion barrier to implanted D atoms and consequently reduce the amount of uptake in the W material. This newly observed effect implies that current predictions of D retention in W, in actual fusion devices, may be overestimated, since there will be He ash in fusion plasma. Toughness enhanced, fine-grained (grain size of ∼1 µm) W–TiC samples, exposed to pure D plasma conditions, also show little or no evidence of blistering. The measured D retention in the W–TiC samples was approximately 1 × 1019 D m−2 corresponding to about 2 × 10−7 of the implanted D fluence, and is very low compared with the retention in pure stress relieved W, which exhibited surface blisters and had a D retention of about 1 × 1021 D m−2.

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Plasma-facing materials and components in a fusion reactor are the interface between the plasma and the material part. The operational conditions in this environment are probably the most challenging parameters for any material: high power loads and large particle and neutron fluxes are simultaneously impinging at their surfaces. To realize fusion in a tokamak or stellarator reactor, given the proven geometries and technological solutions, requires an improvement of the thermo-mechanical capabilities of currently available materials. In its first part this article describes the requirements and needs for new, advanced materials for the plasma-facing components. Starting points are capabilities and limitations of tungsten-based alloys and structurally stabilized materials. Furthermore, material requirements from the fusion-specific loading scenarios of a divertor in a water-cooled configuration are described, defining directions for the material development. Finally, safety requirements for a fusion reactor with its specific accident scenarios and their potential environmental impact lead to the definition of inherently passive materials, avoiding release of radioactive material through intrinsic material properties. The second part of this article demonstrates current material development lines answering the fusion-specific requirements for high heat flux materials. New composite materials, in particular fiber-reinforced and laminated structures, as well as mechanically alloyed tungsten materials, allow the extension of the thermo-mechanical operation space towards regions of extreme steady-state and transient loads. Self-passivating

Tungsten materials are candidates for plasma-facing components for the International Thermonuclear Experimental Reactor and the DEMOnstration power plant because of their superior thermophysical properties. Because these materials are not common structural materials like steels, knowledge and strategies to improve the properties are still under development. These strategies discussed here, include new alloying approaches and microstructural stabilization by oxide dispersion strengthened as well as TiC stabilized tungsten based materials. The fracture behavior is improved by using tungsten laminated and tungsten wire reinforced materials. Material development is accompanied by neutron irradiation campaigns. Self-passivation, which is essential in case of loss-of-coolant accidents for plasma facing materials, can be achieved by certain amounts of chromium and titanium. Furthermore, modeling and computer simulation on the influence of alloying elements and heat loading and helium bombardment will be presented.

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Retention of D in neutron-irradiated W and desorption were examined after plasma exposure at 773 K. Deuterium was accumulated at a relatively high concentration up to a large depth of 50-100 m due to the trapping effects of defects uniformly induced in the bulk. A

The effects of neutron and ion irradiations on deuterium (D) retention in tungsten (W) were investigated. Specimens of pure W were irradiated with neutrons to 0.3 dpa at around 323 K and then exposed to high-flux D plasma at 473 and 773 K. The concentration of D significantly increased by neutron irradiation and reached 0.8 at% at 473K and 0.4 at% at 773 K. Annealing tests for the specimens irradiated with 20 MeV W ions showed that the defects which play a dominant role in the trapping at high temperature were stable at least up to 973 K, while the density decreased at temperatures equal to or above 1123 K. These observations of the thermal stability of traps and the activation energy for D detrapping examined in a previous study (≈1.8 eV) indicated that the defects which contribute predominantly to trapping at 773 K were small voids. The higher concentration of trapped D at 473 K was explained by additional contributions of weaker traps. The release of trapped D was clearly enhanced by the exposure to atomic hydrogen at 473 K, though higher temperatures are more effective for using this effect for tritium removal in fusion reactors.2

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