Abstract
We have investigated the tritium depth profile near the surface of the limiter/divertor tiles used in the deuterium fueled machines, such as TEXTOR, TFTR and JT-60U by means of the imaging plate technique and a tritium survey monitor. Tritium depth profiles near the surface of the sample tiles were estimated by comparing the experimental results to a calculation using a 3-D Monte-Carlo code. In every sample tile, there was little tritium in the range from the surface to 1 μm depth. In contrast, tritium density tended to increase beyond 1 μm depth. These results indicate that the tritium retained near the surface was easily removed by isotope exchange with a deuterium plasma or various other tritium removal operations. On the other hand, such operations did not remove tritium retained beyond 1 μm depth, and this could be a potential issue in a next D-T machine.
| Original language | English |
|---|---|
| Pages (from-to) | 874-879 |
| Number of pages | 6 |
| Journal | Journal of Nuclear Materials |
| Volume | 329-333 |
| Issue number | 1-3 PART A |
| DOIs | |
| Publication status | Published - 2004 Aug 1 |
| Externally published | Yes |
| Event | Proceedings of the 11th Conference on Fusion Research - Kyoto, Japan Duration: 2003 Dec 7 → 2003 Dec 12 |
ASJC Scopus subject areas
- Nuclear and High Energy Physics
- Materials Science(all)
- Nuclear Energy and Engineering