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
Tritium permeation barrier is required in fusion blanket for reduction of loss of fuel and health hazard. In this study, deuterium permeation experiments have been 2 performed on four kinds of steels and erbium oxide coatings fabricated by a filtered arc deposition method. The permeation flux of uncoated samples shows diffusion-limited regime in the temperature range 573−723 K and the permeability is corresponding to literature data. The coated samples deposited at room temperature have been tested at 773 K. It is found that the coatings suppress the deuterium permeation to a close level in spite of different types of steel substrates. In addition, the exponent of the driving pressure slightly changes compared to the uncoated sample. However, the permeation regime is still near diffusion limited.
Suppression of tritium permeation through structural materials is essential in order to mitigate fuel loss and radioactivity concerns. Ceramic coatings have been investigated for over three decades as tritium permeation barriers (TPBs); however, a very limited number of investigations on the mechanism of hydrogen-isotope permeation through the coatings have been reported. In this study, deuterium permeation behaviour of erbium oxide coatings fabricated by filtered arc deposition on reduced activation ferritic/martensitic steels has been investigated. The samples coated on both sides of the substrates showed remarkably lower permeability than those coated on one side, and the maximum reduction efficiency indicated a factor of 105 compared with the substrate. The different permeation behaviour between the coatings facing the high and low deuterium pressure sides has been found by the crystal structure analysis and the evaluation of the energy barriers. It is suggested that the permeation processes on the front and back surfaces are independent, and the TPB efficiency of the samples coated on both sides can be expressed by a multiplication of that of each side.
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