Deuterium retention in radiation damaged tungsten exposed to high-flux plasma

Nuclear fusion has the potential for large-scale sustainable energy production. Currently, the most promising fusion reactor concept is a tokamak. Scientists and engineers from all over the world are collaborating on building the next-generation fusion reactor: ITER. A critical component of the ITER design is the exhaust, the divertor. The material of choice for the divertor is tungsten in order to be able to withstand the extreme heat and particle fluxes that it experiences during operation. One of the main challenges for ITER operation is the tritium retention in the reactor wall. For safety and efficiency reasons, only a limited in vessel tritium inventory is allowed. High energy neutrons, produced in the fusion process, will create damage in the tungsten crystal lattice and enhance the tritium retention in tungsten. This thesis describes the investigation of deuterium retention in radiation damaged tungsten after high-flux plasma exposure. Our results suggest that tritium retention in the divertor of ITER will not be problematic, even when damage creation by neutrons is included.