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.

Alpha particles with energies on the order of megaelectronvolts will be the main source of plasma heating in future magnetic confinement fusion reactors. Instead of heating fuel ions, most of the energy of alpha particles is transferred to electrons in the plasma. Furthermore, alpha particles can also excite Alfvénic instabilities, which were previously considered to be detrimental to the performance of the fusion device. Here we report improved thermal ion confinement in the presence of megaelectronvolts ions and strong fast ion-driven Alfvénic instabilities in recent experiments on the Joint European Torus. Detailed transport analysis of these experiments reveals turbulence suppression through a complex multi-scale mechanism that generates large-scale zonal flows. This holds promise for more economical operation of fusion reactors with dominant alpha particle heating and ultimately cheaper fusion electricity.

One of the main problems in tokamak fusion devices concerns the capability to operate at a high plasma density, which is observed to be limited by the appearance of catastrophic events causing loss of plasma confinement. The commonly used empirical scaling law for the density limit is the Greenwald limit, predicting that the maximum achievable line-averaged density along a central chord depends only on the average plasma current density. However, the Greenwald density limit has been exceeded in tokamak experiments in the case of peaked density profiles, indicating that the edge density is the real parameter responsible for the density limit. Recently, it has been shown on the Frascati Tokamak Upgrade (FTU) that the Greenwald density limit is exceeded in gas-fuelled discharges with a high value of the edge safety factor. In order to understand this behaviour, dedicated density limit experiments were performed on FTU, in which the high density domain was explored in a wide range of values of plasma current (Ip = 500–900 kA) and toroidal magnetic field (BT = 4–8 T). These experiments confirm the edge nature of the density limit, as a Greenwald-like scaling holds for the maximum achievable line-averaged density along a peripheral chord passing at r/a ≃ 4/5. On the other hand, the maximum achievable line-averaged density along a central chord does not depend on the average plasma current density and essentially depends on the toroidal magnetic field only. This behaviour is explained in terms of density profile peaking in the high density domain, with a peaking factor at the disruption depending on the edge safety factor. The possibility that the MARFE (multifaced asymmetric radiation from the edge) phenomenon is the cause of the peaking has been considered, with the MARFE believed to form a channel for the penetration of the neutral particles into deeper layers of the plasma. Finally, the magnetohydrodynamic (MHD) analysis has shown that also the central line-averaged density at the onset of the MHD activity depends only on the toroidal magnetic field.

Plasma Physics and Controlled Fusion

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The paper presents an analysis of disruptions occurring during JET-ILW plasma operations covering the period from the start of ILW (ITER-like wall) operation up to completion of JET operation in 2016. The total number of disruptions was 1951 including 466 with deliberately induced disruptions. The average rate of unintended disruptions was 16.1 %, which is significantly above the ITER target at 15 MA. The pre-disruptive plasma parameters are: plasma current I p = (0.82-3.38) MA, toroidal field B T = (0.98-3.4) T, safety factor q 95 = (1.52-9.05), plasma internal inductance l i = (0.58-1.86), Greenwald density limit fraction FGWL = (0.04-1.61), with 720 X-point plasma pulses from a subset of 1420 unintended disruption shots. Massive gas injection (MGI) has been routinely used in protection mode both to terminate pulses when the plasma is at risk of disruption and to mitigate against disruption effects. The MGI was mainly triggered by the n = 1 locked mode (LM) amplitude exceeding a threshold or by the disruption itself, namely, either dI p /dt (specifically, a fast drop in I p ) or the toroidal loop voltage exceeding threshold values. For mitigation purposes, only the LM was used as a physics precursor and threshold on the LM signal was used to trigger a See Joffrin et al 2019 (

The intrinsic limited toroidal electric field (0.3 V m −1 ) in devices with superconducting poloidal coils (ITER, JT-60SA) requires additional heating, like electron cyclotron (EC) waves, to initiate plasma and to sustain it during the burn-through phase. The FTU tokamak has contributed to studying the perspective of EC assisted plasma breakdown. Afterward, a new experimental and modeling activity addressing the study of assisted plasma start-up in a configuration close to the ITER one (magnetic field, oblique injection, and polarization) has been performed and is presented here. These experiments have been supported by a 0D code, BKD0, developed to model the plasma start-up and linked to a beam tracing code computing, in a consistent way, EC absorption. The FTU results demonstrate the role of polarization conversion at the inner wall reflection. Dedicated experiments also showed the capability of EC power to sustain plasma start-up in the presence of strong error field (12 mT), with a null outside the vacuum vessel. The BKD0 code, applied to FTU data, has been used to determine the operational window of sustained breakdown as a function of toroidal electric field and neutral pressure. Experimental results in agreement with the BKD0 simulations support the use of the code to predict start-up in future tokamaks, like ITER and JT60SA.