Helium-cooled ceramic breeder blanket designs are commnoly based on the use of beryllium as both multiplying and moderating material. The possibility of unig lead as multiplier and graphite as moderator, instead of beryllium, is investigated by means of a proper optimization code, aiming at maximizing the tritium breeding ratio (TBR) vs the material composition under the constraints of the thermal-hydraulic dimensioning. For a fixed value of the volume percentage of voids (25%, including helium coolant), and for upper and lower limits of ceramic breeder (<13%) and structure (>10%), respectively, an optimal blanket configuration is obtained with a 1-D TBR of 1.40 (lithium silicate) and 1.30 (lithium alluminate), with a 75% Li6 enrichment of the breeder material. Results of the optimization code together with hteir design implications are discussed in the paper. A conceptual design of the optimal blanket configuration is developed, starting from a poloidal breeder-in-tube scheme based on the ongoing ENEA helium-cooled blanket design for DEMO. The main operating conditions and features of the DEMO-relevant ceramic breeder lead/graphite blanket design are max. neutron wall load = 2.6 MW/m2; helium coolant inlet/outlet temperature = 250/500 C; max. structure temperature = 550 C; cooling pumpping power percentage = 4%; and helium purge circuit entirely separated from the coolant circuit.
A helium-cooled ceramic breeder blanket concept with lead/graphite as multiplier/moderator / L., Anzidei; M., Gallina; L., Petrizzi; Gatto, Renato. - STAMPA. - 1:(1991), pp. 27-30. ((Intervento presentato al convegno 16th Symphosium on Fusion Technology tenutosi a London nel 3-7 Settembre 1990.