Neutron damage and activation of the first wall of inertial confinement fusion reactors: Recycling and waste disposal

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Abstract
Even though general conclusions cannot be derived for all the protection schemes in inertial confinement fusion (ICF) reactors, the feasibility of the ferritic alloy HT-9 as the main component of the first structural wall (FSW) in ICF facilities using thin-film Li17Pb83 liquid protection, flowing through porous tubes (INPORT), can be demonstrated as a solution in terms of radiation damage. Swelling and shift in the ductile-brittle transition temperature (DBTT) can be analyzed using the results of experimental fast-fission reactors, which are demonstrated to be good experimental tools in that ICF range. The good performance of HT-9 is remarkable. The generation of new solid transmutants and the depletion of initial constituents need also be considered. Further, a reduced-activation HT-9 (niobium-free) has been studied using recycling and shallow land burial (SLB) criteria. The recycling using that HT-9 is shown to be not feasible, as is SLB waste disposal. The unexpected critical role of some short-lived isotopes is remarkable, and more research on their nuclear data must be performed.