Helium retention and surface blistering characteristics of tungsten with regard to first wall conditions in an inertial fusion energy reactor

摘要

The first wall of an inertial fusion energy reactor may suffer from surface blistering and exfoliation due to helium ion fluxes 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 and tungsten microstructure was conducted to learn how the damaging effects of helium may be diminished. Single crystal and polycrystalline tungsten samples were implanted with 1.3 MeV ~3He in doses ranging from 10~(19)/m~2 to 10~(22)/m~2. Implanted samples were analyzed by ~3He(d,p)~4He nuclear reaction analysis and neutron depth profiling techniques. Surface blistering occurred for doses greater than 10~(21) He/m~2 and was analyzed by scanning electron microscopy. Repeated cycles of implantation and flash annealing indicated that helium retention was reduced with decreasing implant dose per cycle. A carbon foil energy degrader, currently in development, will allow a continuous spectrum of helium implantation energy matching the theoretical models of He ion fluxes within the IFE reactor.