Radiation Damage and Activation in the HT-9 Structure of ICF Reactors Using Inport Protection

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Abstract
The feasibility of the ferritic alloy HT-9 as the main component of the first structural wall (FSW) of inertial confinement fusion (ICF) reactors, such as HIBALL-II or LIBRA, which use thin film liquid protection through porous tubes (INPORT) has been studied in terms of radiation damage and activation. Swelling and shift in the ductile brittle transition temperature (DBTT) have been analyzed in the light of the results of experimental fast breeder reactors, which are demonstrated to be good experimental tools in our ICF range. The good performance of HT-9 is remarkable. An analysis of the generation of new solid transmutants and the depletion of initial constituents is given. Activation has been studied using recycling and shallow land burial (SLB) criteria. The interest has been focussed in a reduced activation HT-9 (Niobium-free). Recycling using HT-9 is shown to be not feasible. SLB waste disposal is also not feasible. The critical role of some short lived isotopes as Pt193, Nb93m, Re186 is analyzed, together with that of the more conventional Re186m, Nb94, Bi210m.