[en] Among 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 and 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 ∼3.5-9.0x10¹⁴ cm^-2 s^-1 for the first wall, whilst the neutron flux for the divertor is ∼1-3x10¹⁴ 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 and 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