The evaluation of hydrogenic retention in present tokamaks is of crucial importance to estimate the expected tritium (T) vessel inventory in ITER, limited from safety considerations to 350 g. In the framework of the European Task Force on Plasma Wall Interaction (EU TF on PWI) efforts are underway to investigate gas balance and fuel retention during discharges, and to compare the data obtained with those from post-mortem analysis of in-vessel components exposed over whole experimental campaigns. This paper summarizes the principal findings from coordinated studies on gas balance and fuel retention from a number of European tokamaks, namely, ASDEX-Upgrade (AUG), JET, TEXTOR and Tore Supra (TS). For most devices, the long-term retention fraction deduced from integrated particle balance is ~10–20%. This is larger than the ~3–4% deduced from post-mortem analysis of plasma facing components (PFCs). However, from the database available for tokamaks with their main PFCs made of carbon, the important conclusion is that the T inventory limit (set by the working guideline for operations) could be reached in ITER within fewer than 100 discharges. This, therefore, would seriously impact on operation of the device unless efficient T removal processes are developed.
- fusion energy
- fusion reactors
- plasma-wall interactions
- tritium retention
Loarer, T., Brosset, C., Bucalossi, J., Coad, P., Esser, G., Hogan, J., Likonen, J., Mayer, M., Morgan, P., Philipps, V., Rohde, V., Roth, J., Rubel, M., Tsitrone, E., & Widdowson, A. (2007). Gas balance and fuel retention in fusion devices. Nuclear Fusion, 47(9), 1112-1120. https://doi.org/10.1088/0029-5515/47/9/007