Abstract
The main features of a tritium breeding blanket for a Demonstration Power Reactor involving the eutectic Pb-17Li as liquid breeder and water as coolant are presented.The configuration of the blanket segments and breeder modules as well as their arrangement inside the reactor vacuum vessel are outlined. The main design aspects and the corresponding design limits are reviewed, namely those related to thermomechanics, neutronics, magneto-hydrodynamics, tritium permeation and recovery.First results of safety analyses, in particular those connected with the rupture of a coolant tube in the breeder module are presented and discussed.As a conclusion, the feasibility of the concept looks attractive. A problem which requires further investigation is that of the tritium self-sufficiency. It is shown that a net tritium production near to one can be obtained if beryllium tiles are placed in front of the plasma, provided that they are cooled by heavy water.
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