Tritium Retention and Removal in Tokamaks
- Princeton Plasma Physics Laboratory, POB 451, Princeton, NJ 08543 (United States)
Management of tritium inventory remains one of the grand challenges in the development of fusion energy. Tritium is an important source term in safety assessments, it is expensive and in short supply. Tritium can be continuously retained in a tokamak by codeposition with eroded carbon or beryllium and JET and TFTR with carbon plasma facing components showed a tritium retention level that would be unacceptable in ITER or future fusion reactors. Asdex-U and Alcator C-mod have shown reduced hydrogenic retention with tungsten clad and molybdenum plasma facing components. Once the tritium inventory approaches the administrative limit, tritium must be removed to permit continued D-T plasma operations. Several candidate techniques are being considered and need to be proven at a relevant speed and efficiency in contemporary tokamaks. Projections for ITER are discussed.
- OSTI ID:
- 21260252
- Journal Information:
- AIP Conference Proceedings, Vol. 1095, Issue 1; Conference: 2. ITER international summer school; 47. summer school of JSPF for young plasma scientists: Confinement, Kasuga, Fukuoka (Japan), 22-25 Jul 2008; Other Information: DOI: 10.1063/1.3097310; (c) 2009 American Institute of Physics; Country of input: International Atomic Energy Agency (IAEA); ISSN 0094-243X
- Country of Publication:
- United States
- Language:
- English
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