Retention and surface blistering of helium irradiated tungsten as a first wall material

Abstract The first wall of an inertial fusion energy reactor may suffer from surface blistering and exfoliation due to helium ion irradiation and extreme temperatures. Tungsten is a candidate for the first wall material. A study of helium retention and surface blistering with regard to helium dose, temperature, pulsed implantation, and tungsten microstructure was conducted to better understand what may occur at the first wall of the reactor. Single crystal and polycrystalline tungsten samples were implanted with 1.3 MeV 3He in doses ranging from 1019 m−2 to 1022 m−2. Implanted samples were analyzed by 3He(d,p)4He nuclear reaction analysis and 3He(n,p)T neutron depth profiling techniques. Surface blistering was observed for doses greater than 1021 He/m2. For He fluences of 5 × 1020 He/m2, similar retention levels in both microstructures resulted without blistering. Implantation and flash heating in cycles indicated that helium retention was mitigated with decreasing He dose per cycle.