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Tritium and deuterium retention in graphite limiters in TEXTOR
KTH, Superseded Departments, Alfvén Laboratory.ORCID iD: 0000-0001-9901-6296
2002 (English)In: Fusion science and technology, ISSN 1536-1055, E-ISSN 1943-7641, Vol. 41, no 3, 924-928 p.Article in journal (Refereed) Published
Abstract [en]

In order to investigate tritium behavior in tokamak, we have measured surface distributions of deuterium and tritium on graphite limiter tiles used in TEXTOR under D-D operation by means of an ion beam analysis and tritium imaging plate technique, respectively. It was found that both distributions were quite different, i.e. deuterium retention was higher at the deposited area, whereas tritium retention was higher at the erosion dominated area. This is because tritium produced by the D-D reaction, initially having 1 MeV, did not fully lose its energy in the TEXTOR plasma and implanted into the plasma facing materials nearly homogeneously, whereas deuterium was codeposited with carbon and boron, the main impurities in the TEXTOR plasma. This is also confirmed by the finding that high level of tritium was detected beneath the deposited layer. Tritium distribution, however, was modified by the temperature increase due to plasma heat load. Thus the comparison of tritium profiles with the deuterium profile gives a large amount of important and new information on PMI, and may be used as a new diagnostic technique for PMI.

Place, publisher, year, edition, pages
2002. Vol. 41, no 3, 924-928 p.
Keyword [en]
plasma-facing component, imaging plate technique
Identifiers
URN: urn:nbn:se:kth:diva-21541ISI: 000175564800121OAI: oai:DiVA.org:kth-21541DiVA: diva2:340239
Note
QC 20100525Available 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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