A build-up of co-deposits in remote areas of the divertor can contribute significantly to the overall fuel retention. The control of plasma-material interactions via the study and understanding of erosion-deposition of PFCs provides vital information for the efficient future operation of ITER. The major aim of this work is to reveal details of beryllium deposition and fuel (deuterium) retention on divertor plasma-facing componentsremoved from the JET ITER-Like Wall divertor after cumulative exposure during the first two (ILW-1 + 2) and all three (ILW-1 + 2 + 3) campaigns. Ion beam analysis techniques such as Rutherford backscattering spectrometry, nuclear reaction analysis and proton induced X-ray emission have been extensively used for post-mortem analyses of selected tiles from JET following each campaign and can provide relevant information on plasma-surface interactions like tungsten erosion, beryllium deposition and plasma fuel retention with divertor tiles via implantation or co-deposition. The studied divertor tiles represent a unique set of samples, which have been exposed to plasmas since the beginning of the JET-ILW operation for three successive plasma campaigns. This is a comprehensive comparison of divertor components after these operation periods. The results presented summarise deposition and fuel retention on Tiles 4 (inner base) and 6 (outer base). Although the deposition pattern is similar to that determined after individual campaigns, D retention is not a cumulative process and is determined mainly by the last campaign, and the total Be deposit after the 3 campaigns (i.e. data 1 + 2 + 3 = tile exposed 2011-2016) is less than the sum of the deposits after each individual campaign (sum 1 + 2 + 3) for Tile 4 but greater for Tile 6.
|State||Published - 1 Jan 2020|
|Event||17th International Conference on Plasma-Facing Materials and Components for Fusion Applications, PFMC 2019 - Eindhoven, Netherlands|
Duration: 20 May 2019 → 24 May 2019
- Ion beam analysis
- Plasma-material interaction