As a licensed nuclear facility, ITER must limit the in-vessel tritium (T) retention to reduce the risks of potential release during accidents, the inventory limit being set at 1. kg. Simulations and extrapolations from existing experiments indicate that T-retention in ITER will mainly be driven by co-deposition with beryllium (Be) eroded from the first wall, with co-deposits forming mainly in the divertor region but also possibly on the first wall itself. A pulsed Laser-Induced Desorption (LID) system, called Tritium Monitor, is being designed to locally measure the T-retention in co-deposits forming on the inner divertor baffle of ITER. Regarding tritium removal, the baseline strategy is to perform baking of the plasma-facing components, at 513. K for the FW and 623. K for the divertor. Both baking and laser desorption rely on the thermal desorption of tritium from the surface, the efficiency of which remains unclear for thick (and possibly impure) co-deposits. This contribution reports on the results of TMAP7 studies of this efficiency for ITER-relevant deposits.
- Tritium retention
De Temmerman, G., Baldwin, M. J., Anthoine, D., Heinola, K., Jan, A., Jepu, I., Likonen, J., Lungu, C. P., Porosnicu, C., & Pitts, R. A. (2017). Efficiency of thermal outgassing for tritium retention measurement and removal in ITER. Nuclear Materials and Energy, 12, 267-272. https://doi.org/10.1016/j.nme.2016.10.016