AbstractAmong the low-Z materials, carbon and beryllium are primary candidates for use as plasma facing materials for the International Thermonuclear Experimental Reactor (ITER), because of extensive experience in their application for first wall and divertor plate protection in existing tokamaks. In addition, their excellent plasma performance has been demonstrated. Carbon based materials have been chosen for protection of high heat flux components, whilst beryllium has been proposed as the first wall material for ITER. However, as next generation D/T plasma devices, i.e. ITER, will produce intense neutron fluxes, substantial R&D is needed to elucidate the effects of neutron-induced damage on the microstructure and critical properties of these materials, e.g. thermal conductivity, swelling, and tritium trapping, because they could limit the use of these materials in the next generation fusion devices. Neutron induced changes in thermal conductivity, dimensional stability, mechanical properties as well as behaviour of tritium interaction are crucial problems which need to be better understood. The assessed neutron flux of ITER will be around 3.5–9.0×1014 cm−2 s−1 for the first wall, whilst the neutron flux for the divertor is around 1–3×1014 cm−2 s−1, for which leads to a damage of around 10–20 dpa for the first wall and 3–6 dpa for the divertor for 1 full power year of operation. In the framework of European fusion R&D programs, an extensive effort on neutron effects on plasma facing component (PFC) materials is being undertaken. This paper presents the recent results of experiments performed to investigate the effects of neutron doses and irradiation temperature on the thermal conductivity, mechanical properties, dimensional stability and tritium inventory of various carbon based materials as well as beryllium. The consequences are discussed.
High dose neutron irradiation damage in beryllium as blanket material