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

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 Lind polycrystalline tungsten samples were implanted with 1.3 MeV He-3 in doses ranging from 10(19) m(-2) to 10(22) m(-2). Implanted samples were analyzed by He-3(d,P)(4) He nuclear reaction analysis and He-3(n,p)T neutron depth profiling techniques. Surface blistering was observed for doses greater than 10(21) He/m(2). For He fluences of 5 x 10(20) He/m(2), 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. (c) 2005 Elsevier B.V. All rights reserved.