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 1kg. 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 513K for the FW and 623K 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

After the second Deuterium–Tritium Campaign (DTE2) in the JET tokamak with the ITER-Like Wall (ILW) and full tritium campaigns that preceded and followed after the DTE2, a sequence of fuel recovery methods was applied to promote tritium removal from wall components. The sequence started with several days of baking of the main chamber walls at 240 °C and at 320 °C. Subsequently, baking was superimposed with Ion-Cyclotron Wall Conditioning (ICWC) and Glow Discharge Conditioning (GDC) cleaning cycles in deuterium. Diverted plasma operation in deuterium with different strike point configurations, including a Raised Inner Strike Point (RISP) configuration, and with different plasma heating—Ion Cyclotron Resonance Frequency (ICRF) and Neutral Beam Injection (NBI)—concluded the cleaning sequence. Tritium content in plasma and in the pumped gas was monitored throughout the experiment. The applied fuel recovery methods allowed reducing the residual tritium content in deuterium NBI-heated plasmas to about 0.1% as deduced from neutron rate measurements. This value is well below the requirement of 1% set by the maximum 14 MeV fusion neutron budget allocated in the ensuing deuterium plasma campaign. The quantified tritium removal over the course of the experiment was 13.4 ± 0.7 × 10 22 atoms or 0.67 ± 0.03 g with ∼58% attributed to baking, ∼12.5% to ICWC, ∼26% to GDC, and ∼3.5% to first low power RISP plasmas. The experimentally estimated amount of removed tritium is in good agreement with long-term tritium accounting by the JET tritium reprocessing plant, in which the unaccounted amount was reduced by 0.71 g after the cleaning experiment.

JET returned to deuterium-tritium operations in 2023 (DTE3 campaign), approximately two years after DTE2. DTE3 was designed as an extension of JET’s 2022-2023 deuterium campaigns, which focused on developing scenarios for ITER and DEMO, integrating in-depth physics understanding and control schemes. These scenarios were evaluated with mixed D-T fuel, using the only remaining tritium-capable tokamak until its closure in 2023. A core-edge-SOL integrated H-mode scenario was developed and tested in D-T, showing good confinement and partial divertor detachment with Ne-seeding. Stationary pulses with good performance, no tungsten accumulation, and even without ELMs were achieved in D-T. Plasmas with pedestals limited by peeling modes were studied with D, T-rich, and D-T fuel, revealing a positive correlation between pedestal electron pressure and pedestal electron density. The Quasi-Continuous Exhaust regime was successfully achieved with D-T fuel, with access criteria similar to those in D plasmas. A scenario with full detachment, the X-point radiator regime, was established in D-T, aided by the real-time control of the radiator’s position. The crucial characterisation of tritium retention continued in DTE3, using gas balance measurements and the new LID-QMS diagnostic. Nuclear technology studies were advanced during the DTE3 campaign, addressing issues such as the activation of water in cooling loops and single event effects on electronics. Building on the previous D, T and DTE2 campaigns and the lessons learned from them, DTE3 extended our understanding of D-T plasmas, particularly in scenarios relevant to next-generation devices such as ITER and DEMO.