Tritium retention in plasma facing materials of JET ITER-Like-Wall retrieved from the vacuum vessel in 2012 (ILW1), 2014 (ILW2) and 2016 (ILW3). (June 2021)

• Record Type: Journal Article
• Authors: Pajuste, E.; Kizane, G.; Avotina, L.; Vitins, A.; Teimane, A.S.
• Is Part Of: Nuclear materials and energy. Volume 27(2021)
• Publication Date: 2021-06
• Subjects: Tritium -- Beryllium -- ITER-Like Wall -- Joint European Torus -- Fuel retention; Nuclear energy -- Periodicals; Nuclear fuels -- Periodicals; Nuclear reactors -- Materials -- Periodicals; Radioactive substances -- Periodicals; 621.4833
• DOI: 10.1016/j.nme.2021.101001
• Languages: English
• ISSNs: 2352-1791

Abstract

Highlights: Tritium concentration in the plasma facing surfaces has changed over the campaigns. Highest tritium concentration was observed in the samples from the outer wall. Tritium accumulation patterns differ from those of deuterium. Abstract: ITER-Like-Wall (ILW) project has been carried out at Joint European Torus (JET) to test plasma facing materials relevant to International Thermonuclear Experimental Reactor – ITER [1]. Limiters and an upper dump plate of the vacuum vessel are made of bulk beryllium tiles, whereas for the divertor bulk tungsten and tungsten-coated carbon fibre (CFC) composite tiles are used. During the shutdowns in ILW1 (2012), ILW2 (2014) and ILW3 (2016), selected beryllium tiles were removed from the vacuum vessel. In this study, tiles from three positions were analysed, and analysis results were compared regarding both the tile position in the vacuum vessel and differences in the exploitation conditions during the respective three ILW campaigns: ILW1, ILW2, ILW3. Tritium results have been compared to deuterium data published by other authors [2, 3]. Tritium measurements were performed by two methods - thermal desorption spectroscopy and beryllium chemical etching. Prior to tritium measurements, scanning electron microscopy was used to study structure of the plasma-facing surfaces. Experimental results revealed that tritium content in beryllium samples over all three campaigns is in range of 1.0·10 10 to 9.7·10 13 tritium atoms per square centimetre of the plasma-facing surface area (atoms/cm 2 ). Highest tritium content was found in the samples from outer wall of the vacuum vessel - up to 4.0·10 13 atoms/cm 2 in ILW1, 4.2·10 13 atoms/cm 2 in ILW2 and 9.7·10 13 atoms/cm 2 in ILW3. Whereas, the lowest - in the upper part of the vacuum vessel: 5.1·10 12, 2.8·10 12 and 1.0·10 10 atoms/cm 2 in ILW1, ILW2 and ILW3, respectively. In contrary to the deuterium, in the outer wall tile higher tritium concentrations were found in the central part of the tiles where plasma induced erosion had occurred according to the SEM analysis data. Difference between tritium content in the central part and side part of tile could reach a magnitude of an order - for example, 9.7·10 13 and 6.9·10 12 atoms/cm 2 in the outer wall tile from the ILW3 campaign. Results obtained within this study give possibility to assess tritium retention mechanism and make estimates of its possible inventory in larger machines such as ITER.