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The silicon carbide fiber-reinforced silicon carbide matrix (SiC/SiC) composite system for fusion applications has seen a continual evolution from development a fundamental understanding of the material system and its behavior in a hostile irradiation environment to the current effort which is directed at a broad-based program of technology maturation program. In essence, over the past few decades this material system has steadily moved from a laboratory curiosity to an engineering material, both for fusion structural applications and other high performance application such as aerospace. This paper outlines the recent international scientific and technological achievements towards the development of SiC/SiC composite material technologies for fusion application and discusses future research directions. It also reviews the materials system in the larger context of progress to maturity as an engineering material for both the larger nuclear community and broader engineering applications.

The fusion plasma, with a typical temperature of 10 keV, has to be brought into contact with a physical wall in order to remove the helium produced and drain the excess energy in the fusion plasma. The fusion plasma is far too hot to be brought into direct contact with a physical wall. It would degrade the wall and the debris from the wall would extinguish the plasma. Therefore, schemes are developed to cool down the plasma locally before it impacts on a physical surface. The resulting plasma wall-interaction in ITER is facing several challenges including surface erosion, material redeposition and tritium retention. In JET facility both beryllium and carbon are used together as plasma facing materials, and the resultant surfaces show considerable mixing between the two elements. In this work a first XPS characterisation has been performed on the surface of several PFC (plasma facing components) samples obtained from various JET tiles from the Mk II GB divertor in JET. The substrates of the tiles are CFC, and the deposited films have various D/C ratios depending on the position in the tokamak. Ion beam analysis shows that the D/C ratio is ∼1 in the shadowed areas and Be/C ∼1–2 in the plasma-exposed regions of the inner divertor. These values could give rise to the formation of beryllium carbides, and indications of a carbide component are found in the XPS spectra. The processes for material migration and the morphology of the deposits are discussed.

KS-4V quartz glass (SiO2) is a candidate material to be used in ITER diagnostic systems where it will play an important role for optical components and possibly as electrical insulator. In addition to neutron and gamma radiation, the material will be subjected to bombardment by low energy ions and neutral particles. Previous work has shown severe electrical and optical degradation with a marked temperature dependence when the material is subjected to proton or alpha particle bombardment, due to loss of oxygen from the surface caused by preferential sputtering by energetic ions. In this work, a systematic study of ion bombardment induced surface degradation as a function of the ion mass, energy and dose rate has been carried out. A clear dependence on ion mass has been observed, with degradation occurring much more rapidly for heavier ions. Also the surface electrical degradation depends strongly on the implanted particle energy, degradation is higher for higher energy. No dependence on dose rate was observed.

Ceramic insulators and windows in ITER will be subjected to bombardment by energetic hydrogen isotopes and helium as a consequence of ionization of the residual gas by gamma radiation and acceleration of the ions by local electric fields. Most of the energy carried by these particles will be deposited at or very near the surface giving rise to possible electrical and optical degradation. In this work, radiation induced electrical conductivity for air, helium, and hydrogen for pressures from 1000 to 10 −3 mbar, at 3 Gy/s, and applied voltages up to 1500 V (10 5 V/m) has been measured. The ion currents at the low pressure ((0.2–2) × 10 −6 A/(Gy s −1 m 3 )) are much higher than those expected from both theoretical extrapolations and experimental measurements near to atmospheric pressure (10 −8 A/(Gy s −1 m 3 )). If such high ion currents bombard the insulator surfaces, severe degradation will occur.

CVD diamond is the prime candidate material for use as windows in the ITER ECRH system, and can also be used as transmission component in ECE diagnostics. Hence electrical conductivity, radioluminescence, and optical absorption are important properties for the use of this material. In case of use for ECE diagnostics, the main concerns are absorption and radioluminescence. To address these issues measurements of surface electrical degradation, optical absorption, and radiolumniscence have been made. Intense radioluminescence bands related to nitrogen aggregates and carbon vacancies are observed within the bulk of CVD diamond during electron irradiation. Severe surface degradation occurs in CVD diamond when subjected to ion bombardment, increasing the surface electrical conductivity by many orders of magnitude and drastically reducing the optical transmission.

The long-term objective of the European Fusion Development Agreement (EFDA) fusion materials programme is to develop structural and armor materials in combination with the necessary production and fabrication technologies for reactor concepts beyond the International Thermonuclear Experimental Reactor. The programmatic roadmap is structured into four engineering research lines which comprise fabrication process development, structural material development, armor material optimization, and irradiation performance testing, which are complemented by a fundamental research programme on "Materials Science and Modeling." This paper presents the current research status of the EFDA experimental and testing investigations, and gives a detailed overview of the latest results on materials research, fabrication, joining, high heat flux testing, plasticity studies, modeling, and validation experiments. (C) 2013 Elsevier B.V. All rights reserved.

The current magnetic confinement nuclear fusion power reactor concepts going beyond ITER are based on assumptions about the availability of materials with extreme mechanical, heat, and neutron load capacity. In Europe, the development of such structural and armour materials together with the necessary production, machining, and fabrication technologies is pursued within the EFDA long-term fusion materials programme. This paper reviews the progress of work within the programme in the area of tungsten and tungsten alloys. Results, conclusions, and future projections are summarized for each of the programme's main subtopics, which are: (1) fabrication, (2) structural W materials, (3) W armour materials, and (4) materials science and modelling. It gives a detailed overview of the latest results on materials research, fabrication processes, joining options, high heat flux testing, plasticity studies, modelling, and validation experiments. (C) 2012 Elsevier B.V. All rights reserved.