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Tritium profile in plasma-facing components following D-D operation
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2004 (English)In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 329-33, 874-879 p.Article in journal (Refereed) Published
Abstract [en]

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 mum depth. In contrast, tritium density tended to increase beyond 1 mum 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 mum depth, and this could be a potential issue in a next D-T machine.

Place, publisher, year, edition, pages
2004. Vol. 329-33, 874-879 p.
Keyword [en]
imaging plate technique, co-deposited layers, w-shaped divertor, jt-60u, jet, tftr, deuterium, retention, hydrogen, surface
National Category
Physical Sciences
Identifiers
URN: urn:nbn:se:kth:diva-23673DOI: 10.1016/j.jnucmat.2004.04.345ISI: 000223505000168Scopus ID: 2-s2.0-3342962653OAI: oai:DiVA.org:kth-23673DiVA: diva2:342372
Note
QC 20100525 QC 20110926. 11th International Conference on Fusion Reactor Materials (ICFRM). Kyoto, JAPAN. DEC 07-12, 2003 Available from: 2010-08-10 Created: 2010-08-10 Last updated: 2017-12-12Bibliographically approved

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Rubel, Marek J.

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