UNDERSTANDING RETENTION AND TRANSPORT OF HYDROGEN IN DISPLACEMENT-DAMAGED TUNGSTEN

In order to achieve tritium self-sufficiency in a future fusion reactor, the limitations in terms of tritium retention and loss to the coolant need to be very stringent. Among many other favorable properties, intrinsically low fuel retention makes tungsten, therefore, one material of choice as plasma-facing material. However, during plasma operation defects in the tungsten lattice will evolve that will trap hydrogen isotopes. While for present day experiments this increased retention is only limited to the near surface as a cause of the interaction with eV to keV energy ions and neutrals the material will be modified throughout the whole bulk in future nuclear devices because of the neutron irradiation.

In this presentation, I will first motivate the need for well-defined laboratory experiments to study transport and retention of hydrogen isotopes in plasma-facing materials. The two possible experimental strategies to study the influence of displacement damage will be presented: fission neutron irradiation and MeV ion irradiation. Recent experiments showing sub-threshold and super-threshold creation of defects will be presented to show the different approaches found in literature. Finally, hydrogen isotope exchange experiments will be presented that underline the present understanding of hydrogen retention by trapping in defects and transport by thermal de-trapping.