AbstractAmong the low-Z materials, carbon and beryllium are seriously considered as plasma facing materials (PFC) for the next step fusion experimental reactor, because of their wide experiences as first wall and divertor plate protection in present tokamaks. In addition, their excellent plasma performance has been demonstrated. Carbon based materials have been chosen for protection of high heat flux components, in particular to off-normal operation, e.g. disruption, plasma excursion, whilst beryllium has been proposed as a PFC material for getting oxygen impurity in plasma. However, as next generation D/T plasma devices, i.e. International thermonuclear experimental reactor (ITER) producing 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 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 ∼3.5–9.0×1014 cm−2 s−1 for the first wall, whilst the neutron flux for the divertor is ∼1–3×1014 cm−2 s−1, which leads to a damage of ∼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 of PFC materials is being undertaken. This paper presents the first results of experiments performed to investigate the effects of thermal annealing on neutron damage of thermal conductivity and tritium inventory of various carbon based materials. The consequences are discussed.
在低原子序数材料中，碳和铍被认真考虑作为下一代聚变实验装置（如国际热核实验反应堆ITER）的面向等离子体材料（PFM），因为它们在现有托卡马克装置中作为第一壁和偏滤器保护材料已有广泛经验。此外，它们优异的等离子体性能已得到证实。碳基材料已被选择用于保护高热通量部件，特别是在非正常运行工况（如破裂、等离子体位移）下，而铍则被提议作为面向等离子体材料以降低等离子体中的氧杂质。然而，对于下一代D/T等离子体装置（即产生强中子通量的ITER），需要进行大量的研发工作，以阐明中子辐照损伤对这些材料微观结构和关键性能（如热导率、肿胀和氚滞留）的影响，因为这些影响可能限制这些材料在下一代聚变装置中的使用。中子引起的热导率变化、尺寸稳定性、力学性能以及氚相互作用行为是需要更好理解的关键问题。ITER第一壁的评估中子通量约为3.5–9.0×10¹⁴ cm⁻² s⁻¹，而偏滤器的中子通量约为1–3×10¹⁴ cm⁻² s⁻¹，这导致在满功率运行一年后，第一壁的损伤约为10–20 dpa，偏滤器的损伤约为3–6 dpa。在欧洲聚变研发计划的框架下，正在大力开展关于中子对PFM材料影响的研究。本文介绍了为研究热退火对多种碳基材料中子损伤后热导率和氚滞留的影响而进行的实验的首批结果，并讨论了其后果。