Estimation of tritium accumulation in tungsten and its application for ITER divertor

Tungsten(W) is currently contemplated as plasma facing material because of its advantageous thermo physical properties and rather low solubility of tritium. Tritium solubility of W estimated in this study is 3 order higher than that reported by literature. Traps or oxide films may affect the retention capability of W and lead significantly modified release properties. It became clear that there were capture sites that had different thermal stability and capture intensity in W after polishing, or oxide films that were grown on the surface of W and had barrier effects. Detailed investigation of the impact of possibly rather diverse traps produced either during manufacturing-or via radiation-induced processes and oxide films after annealing on the uptake and retention properties of hydrogen isotopes retained by W used in first wall components of fusion machines is therefore necessary in order to assess correctly and minimize the tritium inventory during various phases of operation.