The impact of surface conditions on tritium recycling, permeation and inventory in a self-cooled Pb-17Li fusion reactor blanketFunded under: FP1-FUSION 9C
Tritium behaviour in a self cooled liquid breeder blanket, using the permeation of tritium from the blanket through the first wall into the torus chamber as tritium refuelling method, has been investigated. Tritium recycling and desorption, permeation to the shield and inventory under different conditions at the inner first wall surface and the outer blanket surface have been calculated by a numerical code. The development of these properties as a function of time is discussed. The calculations have been performed for a self cooled blanket using V as a first wall material and Pb-17Li as a liquid breeder material.
Bibliographic Reference: Paper presented: Course and Workshop on Tritium Technology, Varenna (IT), June 6-16, 1989
Availability: Available from (1) as Paper EN 34927 ORA
Record Number: 199011558 / Last updated on: 1994-12-02
Original language: en
Available languages: en