Understanding of fuel retention and release processes from materials for ITER (International Thermonuclear Experimental Reactor) is important from fundamental and technological aspects. Detailed information regarding fuel retention and release characteristics will allow global fuel inventories to be estimated in fusion devices as well as indicate the requirements for development detritiation methods and the re-use of tritium. Selected beryllium (Be) tiles were extracted from the JET vacuum vessel after each experimental campaign period with the ITER-like wall (ILW); so called ILW1 (2011-2012), ILW2 (2013-2014) and ILW3 (2015-2016). Desorption of hydrogen isotopes of samples taken from the inner wall guard limiter (IWGL), outer poloidal limiter (OPL) and dump plate (DP) tiles were analysed by means of thermal desorption spectrometry (TDS). The results presented compare data across ILW1, ILW2 and ILW3 and show the long term trends of fuel retention in Be limiter tiles. For all three campaigns the level of retention correlates with erosion and deposition that takes place during plasma operations Deuterium retention varies from 0.01-1 × 1018 atoms cm-2. Deuterium release takes place in several stages, related to different types of traps which can be within the co-deposit layer and/or below the surface in the Be bulk, with the main release stages around 700-760, 850-900 and 1020 K. The level of tritium in Be was found to be 104 times lower than deuterium for ILW campaigns 1-3.
|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
- Hydrogen isotopes
- Thermal desorption