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

doi:10.1016/j.jnucmat.2004.11.005

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

Engineering Science for Advanced energy system

Research output: Contribution to journal › Article › peer-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 language	English
Pages (from-to)	83-92
Number of pages	10
Journal	Journal of Nuclear Materials
Volume	340
Issue number	1
DOIs	https://doi.org/10.1016/j.jnucmat.2004.11.005
Publication status	Published - Apr 1 2005

All Science Journal Classification (ASJC) codes

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

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

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.",

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T1 - Tritium release behavior from the graphite tiles used at the dome unit of the W-shaped divertor region in JT-60U

AU - Katayama, K.

AU - Takeishi, T.

AU - Manabe, Y.

AU - Nagase, H.

AU - Nishikawa, M.

AU - Miya, N.

PY - 2005/4/1

Y1 - 2005/4/1

N2 - 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.

AB - 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.

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

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