Tritium Removal from Carbon Plasma Facing Components

Skinner, C.H. ; Coad, J.P. ; Federici, G.

Princeton University. Plasma Physics Laboratory.

关键词: Removal;    Heating;    Tritium Plasma-Wall Interaction;    70 Plasma Physics And Fusion Technology;    Tritium;   

DOI  :  10.2172/820208
RP-ID  :  PPPL-3906
RP-ID  :  AC02-76CH03073
RP-ID  :  820208

美国|英语

来源: UNT Digital Library

【 摘 要 】

Tritium removal is a major unsolved development task for next-step devices with carbon plasma-facing components. The 2-3 order of magnitude increase in duty cycle and associated tritium accumulation rate in a next-step tokamak will place unprecedented demands on tritium removal technology. The associated technical risk can be mitigated only if suitable removal techniques are demonstrated on tokamaks before the construction of a next-step device. This article reviews the history of codeposition, the tritium experience of TFTR (Tokamak Fusion Test Reactor) and JET (Joint European Torus) and the tritium removal rate required to support ITER's planned operational schedule. The merits and shortcomings of various tritium removal techniques are discussed with particular emphasis on oxidation and laser surface heating.