Tritium Removal from Carbon Plasma Facing Components
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.
- Research Organization:
- Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)
- Sponsoring Organization:
- USDOE Office of Science (SC) (US)
- DOE Contract Number:
- AC02-76CH03073
- OSTI ID:
- 820208
- Report Number(s):
- PPPL-3906; TRN: US0305721
- Resource Relation:
- Other Information: PBD: 24 Nov 2003
- Country of Publication:
- United States
- Language:
- English
Similar Records
Tritium Removal from JET and TFTR Tiles by a Scanning Laser
In-vessel tritium retention and removal in ITER