Tritium release behavior from the graphite tiles used at the dome unit of the W-shaped divertor region in JT-60U

K. Katayama, T. Takeishi, Y. Manabe, H. Nagase, M. Nishikawa, N. Miya

Research output: Contribution to journalArticlepeer-review

9 Citations (Scopus)

Abstract

Release behavior of tritium from the graphite tiles used at dome top and inner dome wing in JT-60U was investigated by the thermal desorption method in dry argon, argon with oxygen and water vapor, or argon with hydrogen. It was found that approximately 20-40% of total tritium is left in graphite even after heating to the high temperature above 1000 °C in dry argon. The residual tritium could be removed by exposing the graphite tile to oxygen with water vapor or hydrogen at the high temperature above 1000 °C. The tritium retention of the dome top tile was quantified as 84-30 kBq/cm2. The inner dome wing tile had a steep tritium distribution from 8 to 0.1 kBq/cm 2. It is observed that a measurable amount of tritium existed in the deep site of the graphite tile.

Original languageEnglish
Pages (from-to)83-92
Number of pages10
JournalJournal of Nuclear Materials
Volume340
Issue number1
DOIs
Publication statusPublished - Apr 1 2005

All Science Journal Classification (ASJC) codes

  • Nuclear and High Energy Physics
  • Materials Science(all)
  • Nuclear Energy and Engineering

Fingerprint Dive into the research topics of 'Tritium release behavior from the graphite tiles used at the dome unit of the W-shaped divertor region in JT-60U'. Together they form a unique fingerprint.

Cite this