To have reliable predictions of the thermal loads on the divertor plates and of the core plasma purity in the proposed Fusion Advanced Studies Torus (FAST) tokamak, numerical self-consistent simulations have been made for the H-mode and steady-state scenario by using the 2D multi-fluid code COREDIV. In the simulations full W plasma facing components, foreseen for basic operation, as well as liquid lithium divertor targets have been considered. Impurity seeding, for reducing divertor heat loads, was allowed. The overall picture shows that, marginally in the intermediate and, necessarily in the high density H-mode scenarios (average density 〈ne〉 = 2 × 1020 m−3 and 5 × 1020 m−3, respectively), impurity seeding should be foreseen with W as target material; however, only a small amount of Ar (0.03% atomic concentration), not affecting the core purity, is sufficient to maintain the divertor peak loads below 18 MW m−2, which represents the safety limit for the W monoblock technology, presently accepted for the ITER divertor tiles. Li always needs additional impurities for decreasing divertor heat loads. At low plasma densities (but ⩾1.3 × 1020 m−3), typical of steady-state regimes, W alone is effective in dissipating the input power by radiative losses, without excessive core contamination. Impurity seeding would lead to excessive W sputtering by Ar and too high Zeff. The impact of unmitigated giant (1.5 MJ) type I edge localized modes on the W divertor targets was also analysed: the resulting maximum energy load of 1 MJ m−2, larger than the tolerable one by a factor of 3, seems not difficult to recover by foreseen mitigation tools.
A new quasi-stationary, very high density plasma regime on the W7-AS stellarator