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Abstract

A conceptual study of a helium cooled solid breeder blanket for a tokamak reactor is presented. Tritium breeding capability together with system reliability are taken as the main design criteria. The blanket consists of tubular poloidal modules made of a central bundle of ceramic rods with coaxial distribution of the inlet/outlet coolant flow (He) surrounded by multiplier material (Be) in the form of bored bricks, in the ratio 1:4. The He inlet and outlet branches are cooling Be and ceramic respectively. A purge He flow running through small central holes of the ceramic rods is derived from the main flow. Under the typical conditions of a Tokamak Reactor (neutron wall load = 2MW/m2), a full coverage tritium breeding ratio of 1.47 is achieved for the following design and operating parameters: outlet He temperature = 570 C, inlet He temperature = 250 C, total extracted power = 2700 MW, He pumping power percentage = 2%, minimum/maximum ceramic temperature = 400/900 C, maximum structural temperature = 475 C, maximum Be temperature = 525 C.

Additional information

Authors: ENEA ENEA, CENTRO RICERCHE ENERGIA FRASCATI, ROMA (IT), ENEA, CENTRO RICERCHE ENERGIA FRASCATI, ROMA (IT)
Bibliographic Reference: REPORT: RT/FUS/88/1 (1988) 28 PP. AVAILABLE FROM SERVIZIO STUDI E DOCUMENTAZIONE, ENEA, CENTRO RICERCHE ENERGIA FRASCATI, C.P.65-00044 FRASCATI, ROMA (IT)
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