The operational domain for active control of type-I edge localized modes (ELMs) with an n = 1 external magnetic perturbation field induced by the ex-vessel error field correction coils on JET has been developed towards more ITER-relevant regimes with high plasma triangularity, up to 0.45, high normalized beta, up to 3.0, plasma current up to 2.0 MA and q 95 varied between 3.0 and 4.8. The results of ELM mitigation in high triangularity plasmas show that the frequency of type-I ELMs increased by a factor of 4 during the application of the n = 1 fields, while the energy loss per ELM, W/W , decreased from 6% to below the noise level of the diamagnetic measurement (<2%). No reduction of confinement quality (H 98Y ) during the ELM mitigation phase has been observed. The minimum n = 1 perturbation field amplitude above which the ELMs were mitigated increased with a lower q 95 but always remained below the n = 1 locked mode threshold. The first results of ELM mitigation with n = 2 magnetic perturbations on JET demonstrate that the frequency of ELMs increased from

An experiment has been performed at the Joint European Torus (JET) which has demonstratedclear self-heating of a deuterium-tritium plasma by alpha particles produced in fusion reactions. Since the alpha power was approximately 10% of the total power absorbed by the plasma, the heating was distinguished from other changes, due to isotopic effects, by scanning the plasma and neutral beam mixtures together from pure D to nearly pure T in a hot ion H-mode with 10.5MW neutral beam power. At an optimum mixture of 60±20% T, the fusion gain (=P fusion / P absorbed ) was 0.65 and the alpha heating showed clearly as a maximum in electron temperature.

The JET 2019-2020 scientific and technological programme exploited the results of years of concerted scientific and engineering work, including the ITER-like wall (ILW: Be wall and W divertor) installed in 2010, improved diagnostic capabilities now fully available, a major Neutral Beam Injection (NBI) upgrade providing record power in 2019-2020, and tested the technical & procedural preparation for safe operation with tritium. Research along three complementary axes yielded a wealth of new results. Firstly, the JET plasma programme delivered scenarios suitable for high fusion power and alpha particle physics in the coming D-T campaign (DTE2), with record sustained neutron rates, as well as plasmas for clarifying the impact of isotope mass on plasma core, edge and plasma-wall interactions, and for ITER pre-fusion power operation. The efficacy of the newly installed Shattered Pellet Injector for mitigating disruption forces and runaway electrons was demonstrated. Secondly, research on the consequences of long-term exposure to JET-ILW plasma was completed, with emphasis on wall damage and fuel retention, and with analyses of wall materials and dust particles that will help validate assumptions and codes for design & operation of ITER and DEMO. Thirdly, the nuclear technology programme aiming to deliver maximum technological return from operations in D, T and D-T benefited from the highest D-D neutron yield in years, securing results for validating radiation transport and activation codes, and nuclear data for ITER.

Results of experiments investigating the performance of the JET Mark IIA divertor are reported and compared to the performance of its Mark I predecessor. The principal e ect of reducing the divertor width increasing closure was to increase pumping for both deuterium and impurities while reducing upstream neutral pressure. Neither the orientation of the divertor target relative to the divertor plasma nor the width of the divertor had a major in uence on core plasma performance in ELMy H-modes. Changing the core triangularity and thus the edge magnetic shear modi es the ELM frequency in ELMy H-mode plasmas thereby c hanging the peak divertor power loading. The integrated performance of the core and divertor plasmas is reviewed with a view to extrapolation to the requirements of ITER. The con nement of JET ELMy H-modes with hot, medium density edges is good and follows gyro-Bohm scaling. The impurity content of these discharges is low and within the ITER requirements. When one attempts to raise the density with deuterium gas fuelling the ELM frequency increases and the con nement, especially in the edge, decreases. Good con nement can be achieved in JET either by producing a large edge pedestal, typically in discharges with neutral beam heating or by centrally peaked heating with ICRH schemes. Large amplitude, Type I ELMs, which are present in all discharges with a large edge pedestal, would result in unacceptable divertor plate erosion when scaled to ITER. Since the power deposition pro le due to heating in ITER is calculated to be intermediate between JET NB and RF heating pro les, it is likely that operation in ITER with small ELMs in order to reduce rst wall loading will result in degraded con nement compared to present d a y scaling laws.

During the JET Preliminary Tritium Experiment (PTE), an estimated 2 X IO'* Bq (1.1 X 10'' atoms) of tritium were injected into the JET vacuum vessel. A series of experiments was performed whose purpose was to deplete the torus of tritium, to compare the effectiveness of different methods of tritium removal and to obtain a quantitative understanding of the processes involved. The effectiveness of the cleaning procedures was such that the normal tokamak programme was resumed one week after the PTE and the routing of exhaust gases to the atmosphere after two weeks. The release of tritium from the vessel was found to scale with the deuterium release from the vessel, suggesting that dilution and mixing of the hydrogen isotopes in the vessel walls is important. High density, disruptive tokamak discharges were found to be the most successful plasma pulses for tritium removal. Purges with deuterium gas were also effective and have the advantage of operational simplicity. Helium discharges, on the other hand, resulted in low tritium release from the vessel walls. It was demonstrated that the tritium release rate could be predicted using data from hydrogen to deuterium changeover experiments. Using the superior quality of data available from the tritium cleanup experiment, the physical mechanisms necessary to describe the hydrogenic uptake and release from the JET torus were identified. The release of tritium is reproduced using a model that incorporates implantation into a thin surface layer as well as diffusion of tritium into and out of the bulk material.