In this work, the deuterium (D) retention in plasma facing components of the divertor of ITER is estimated. Three scenarios are simulated with 3 different surface temperatures, 1456 K, 870 K and 435 K. They represent the exposure of different parts of the divertor during an attached plasma. Our 1D rate equation code MHIMS (migration of hydrogen in materials) is used to model the retention in the super-saturated layer formed in the first 10 nm: the D retention integrated in this 10 nm-layer is ≈1019 D m−2 for the coldest scenarios. It is also used to differentiate the evolution of deuterium retention during pulsed and continuous plasma exposures which shows that: (i) there is a retention during the ramp-down in the first 10 μm which is released during the ramp up and (ii) the bulk retention is not affected by the cycling of plasma exposure. The concentration of mobile deuterium in the implantation zone is used as an input of our finite element code FESTIM (finite element simulation of tritium in materials) which is used to assess the deuterium retention and migration in the 2D complex geometry of the actively cooled plasma facing components. In the end, this work enable to determine the three following macroscopic quantities: the total deuterium retention, the permeation flux to the cooling pipe and the desorption flux from the toroidal edges of the components. It is shown that (i) the coldest scenario leads to the highest retention despite the lowest exposure flux which has already been observed in past retention studies, (ii) the permeation to the cooling pipes happens after few thousands of seconds only for the hottest scenario, (iii) the release of deuterium from the toroidal edges is a small fuel recycling source.
Influence of plasma impurities on the fuel retention in tungsten