Tritium retention in neutron-irradiated low- Z materials for use as plasma facing materials

Abstract

Among the presently available low- Z materials, graphite, carbon-based composites (CFC) and beryllium represent the primary candidate materials to be used as protection for both the first wall and the high heat flux components (e.g., the divertor) in the next-step fusion reactor. Research and evaluations are underway to study the characteristics of several graphite and CFC as well as beryllium grades associated with safety, tritium release, heat transfer, thermal-mechanical and irradiation stability. Several types of graphite and CFC samples were irradiated in the high flux reactor (HFR) at different temperatures but with the same irradiation damage, while beryllium samples were irradiated in the BR2 reactor at low temperature but with a very high neutron fluence. In this paper, the result of a series of out-of-pile annealing tests aiming at investigating the tritium release kinetics after both neutron irradiation and tritium-loading from three graphite grades (i.e., A05, RGTi(91), CL5890), two DunlopCFC grades (i.e., Concept 1, Concept 2) and four SepCarbCFC grades (i.e., N312C, NS11, N112, N312B) are presented. Furthermore, the tritium and helium release behavior in a beryllium grade (i.e., S-200E) produced by the Kawecki Berylco Industries is also presented.