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Conceptual Design of the Blanket Tritium Recovery System for the Prototype Fusion Reactor

T. Kakuta, S. Hirata, S. Mori, S. Konishi, Y. Kawamura, M. Nishi, Y. Ohara

Fusion Science and Technology / Volume 41 / Number 3P2 / May 2002 / Pages 1069-1073

Blanket Material and Process / Proceedings of the Sixth International Conference on Tritium Science and Technology Tsukuba, Japan November 12-16, 2001

In this design work, a combination of the hydrogen pumps in charge of the function of hydrogen isotopes recovery and the oxygen pump was adopted to the blanket tritium recovery system for the prototype fusion reactor designed by Japan Atomic Energy Research Institute (JAERI). The main functions of this system are described below. 1) Transport of tritium with helium purge gas: tritium released from ceramic breeding material in the blanket is transported to the tritium recovery system by the helium purge gas which contains a small amount of hydrogen gas. 2) Steam electrolysis and removal of oxygen gas: the oxygen pump with electrolyte of oxygen ionic conductors electrolyzes the steam (H2O and HTO) contained in the puige gas into hydrogen isotopes and oxygen, and simultaneously removes impurity of oxygen by electrical membrane permeation. 3) Recovery of hydrogen isotopes: the hydrogen pump with electrolyte of protonic conductors electrically recovers the pure hydrogen isotopes (HT and H2) from the purge gas. Based on the experimental data obtained by feasibility study and the present design effort, it was revealed that the simple and continuous tritium recovery system for gaseous stream is possible and attractive for fusion power reactors.

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