Tritium trapping on the plasma irradiated tungsten surface
Tungsten (W) is a candidate material for plasma-facing high heat-flux structures in future fusion reactors. The aim of this study is to assess how reasonably one can predict the tritium inventory in actual fusion machines using data on the hydrogen isotope inventory obtained in laboratory experiments. W specimens previously exposed to deuterium (D) plasmas both in the TEXTOR tokamak and high flux linear plasma generator (LPG) were subsequently loaded with tritium at 573 K for 3 h. The retention of tritium in the near-surface W layer was examined by imaging plate technique. The study shows that on the TEXTOR-plasma-exposed W surface, tritium was mainly trapped in carbon deposits, and for LPG-plasma-exposed W specimens, tritium was trapped in defects created in the near-surface layer during the course of D plasma exposure.
- Hydrogen Isotope Research Center, University of Toyama, Toyama (Japan)
- Tritium Technology Group, Japan Atomic Energy Agency - JAEA, Tokai-mura, Naka-gun, Ibaraki (Japan)
- Graduate School of Engineering, Osaka University, Suita, Osaka (Japan)
- International Research Center for Nuclear Materials Science, Institute for Materials Research - IMR, Tohoku University, Oarai, Ibaraki (Japan)
- Institute of Energy and Climate Research - Plasma Physics, Forschungszentrum Juelich, Association EURATOM-FZJ, Juelich (Germany)
- Publication Date:
- OSTI Identifier:
- Resource Type:
- Journal Article
- Resource Relation:
- Journal Name: Fusion Science and Technology; Journal Volume: 67; Journal Issue: 3; Conference: TRITIUM 2013: 10. International Conference on Tritium Science and Technology, Nice Acropolis (France), 21-25 Oct 2013; Other Information: Country of input: France; 11 refs.
- Country of Publication:
- United States
- 70 PLASMA PHYSICS AND FUSION TECHNOLOGY; CARBON; DEUTERIUM; FIRST WALL; PLASMA; SURFACES; TEMPERATURE RANGE 0400-1000 K; TEXTOR TOKAMAK; THERMONUCLEAR REACTORS; TRITIUM; TUNGSTEN