Control of heat exhaust is essential for the operation of power producing fusion reactors. Here, we present results of heat exhaust feedback control experiments in JET and AUG. In JET, we demonstrate the first X-point radiator (XPR) control in DD and DT discharges using argon seeding. In AUG, we improve the XPR control with nitrogen seeding, resulting in achieving the first detached L-H and H-L transition (in a single discharge). The controllers are designed using a model-based design procedure. The required models are obtained experimentally using perturbative (system identification) experiments. We study the dynamic response of the XPR to various seeding species and varying operating conditions. We find that the sensitivity (relative gain) of the XPR varies as function the height of the radiator inside the confined region but that the relative phase is consistent for all operating points. In AUG, the XPR is also less sensitive to impurity seeding changes for higher heating powers. In JET, we show that the XPR dynamics are the same for DD and DT plasmas. However, we observe that XPR control is only possible with argon and not with neon. The results show that a controller might well be designed in earlier stages of operation of a future device, but remains applicable and can be further tuned for full power operation.

In ITER, to mitigate the deleterious effects of plasma disruptions, massive quantities of radiating impurities will be injected into the disrupting plasma by shattered pellet injectors (SPI) to pre-emptively radiate away the stored thermal and magnetic energy (Lehnen et al Proc. 27th IAEA Fusion Energy Conf. (FEC 2018) (Gandhinagar, India) EX/P7-12). However, asymmetries in the radiation pattern could result in intense photon flashes during the thermal quench that could locally damage or erode the stainless steel plasma-facing surface of the diagnostic port plugs (Pitts et al 2015 J. Nucl. Mater. 463 748-75). Experiments have been undertaken at JET to assess the potential dependence of the radiated power asymmetry on plasma energy during SPI mitigated disruptions. Calculations of the toroidal asymmetry in the radiated power indicate that the toroidal peaking factor is largest near the SPI position and decreases with the plasma stored energy, which is a promising result in view of radiation heat loads during mitigated disruptions in ITER.

JET, the world’s largest operating tokamak with unique Be/W wall and tritium handling capability, completed a Deuterium-Tritium (D-T) campaign in 2021 (Maggi et al 29th Fusion Energy Conf. ) following a decade of preparatory experiments, dedicated enhancements, technical rehearsals and training (Horton et al 2016 Fusion Eng. Des. 109–111 925). Operation with tritium raises significant technical, safety and scientific challenges not encountered in standard protium or deuterium operation. This contribution describes the tritium operational requirements, pulses and technical preparations, new operating procedures, lessons learned and details on the achieved operational availability and performance. The preparation and execution of the recent JET tritium experiments benefitted from the previous experience in 1991 (Preliminary Tritium Experiment), 1997 (DTE1 campaign) and 2003 (Trace Tritium Campaigns) and consisted of the following five phases: technical rehearsals and scenario preparation, tritium commissioning, 100% tritium campaign, D-T campaign (DTE2), tritium clean-up. Following the clean-up JET resumed normal operation and is currently undertaking a further D-T campaign (DTE3).

This work describes the usage of Error Field Correction Coil (EFCC) system [Barlow I. et al. 2001 Fusion Eng. Des. 58-59 189], which is a set of 4 coils located external to the vessel of the JET device, with the aim of introducing non-axisymmetric n =1 magnetic field perturbations in various targeted plasma experiments. Besides being used to characterize and correct the intrinsic error field, the EFCCs system has been exploited to i) probe plasma stability in high-ON regimes through the MHD spectroscopy technique, allowing the identification of resistivewall mode stable operational scenarios, ii) study radiation asymmetry when testing the shattered pellet injector as the prime candidate for ITER disruption mitigation system, and iii) control edge localized modes, permitting a reduction in heat fluxes and carbon erosion of the divertor target plates.

The achievement of a steady ELMy H-mode phase with high ion temperature, but without a gradual rise in plasma radiation, has been a crucial point to establish high plasma performance scenarios in JET ITER-like-wall plasmas. Indeed, radiation events, due to the release of high Z impurities, such as Nickel and Copper, and W sputtered from the divertor, can strongly reduce the power crossing the plasma separatrix and slow the ELMs dynamics, thus inducing H to L transition. In particular, in JET baseline plasmas, because of the outward neo-classical transport [A.R. Field et al 2021 Plasma Phys. Control. Fusion 63 095013], plasma impurities are mainly localized in the mantle region, as detected by a real-time surrogate model for bolometer tomography based on machine learning [D.R. Ferreira et al 2021 Fusion Engineering and Design 164], and the consequent excessive radiation in this region is the main cause of plasma termination in recent Deuterium, Tritium and Deuterium -Tritium operations. To guarantee impurity accumulation being flushed by the ELMs, ELM control schemes, which ensure a throughput of particles, either via gas fueling and via pellets, have been exploited. In this work, the staged approach strategy towards radiation control, which allowed to sustain for more than 10 s Tritium and Deuterium-Tritium baseline discharges without radiation issues, is presented.

The success of magnetic confinement fusion as an energy source relies crucially on reaching the necessary high temperatures for the fuel D and T ions.In a fusion reactor, plasma heating with waves in the ion cyclotron range of frequencies (ICRF) is the only system capable of providing a large fraction of bulk ion heating.Furthermore, in view of a better understanding of the non-linear physics of alpha heating in ITER and future reactors, generating MeV-range ions and studying the impact of fast ions on plasma stability and confinement becomes progressively more important.This paper summarizes recent theoretical progress and experimental demonstrations of the so-called three-ion ICRF scenario on the tokamaks Alcator C-Mod, ASDEX Upgrade and JET.In particular, the paper highlights several key results relevant for ITER and future fusion reactors.

The JET hybrid scenario has been developed from low plasma current carbon wall discharges to the record-breaking Deuterium-Tritium plasmas obtained in 2021 with the ITER-like Be/W wall. The development started in pure Deuterium with refinement of the plasma current, and toroidal magnetic field choices and succeeded in solving the heat load challenges arising from 37 MW of injected power in the ITER like wall environment, keeping the radiation in the edge and core controlled, avoiding MHD instabilities and reaching high neutron rates. The Deuterium hybrid plasmas have been re-run in Tritium and methods have been found to keep the radiation controlled but not at high fusion performance probably due to time constraints. For the first time this scenario has been run in Deuterium-Tritium (50:50). These plasmas were re-optimised to have a radiation-stable H-mode entry phase, good impurity control through edge Ti gradient screening and optimised performance with fusion power exceeding 10 MW for longer than three alpha particle slow down times, 8.3 MW averaged over 5 s and fusion energy of 45.8 MJ.

The RAPTOR suite of codes combines real-time model-based predictions of the plasma state with the available diagnostic measurements. Following the work on TCV and ASDEX-Upgrade, this paper presents the implementation of the RAPTOR suite for JET. This suite embeds: the upgraded equilibrium reconstruction EQUINOX code, the new FLUXMAP algorithm, which maps the diagnostic measurements from geometric to normalized magnetic flux coordinates; the RABBIT code for the NBI reconstruction and eventually RAPTOR state observer, which combines the output from all these codes with the predictions of 1D control-oriented transport code. The suite is both implemented in MATLAB/Simulink (R) and it is being integrated in the C++ real-time MARTe2 framework. Thanks to its user-friendly interfaces, which are based on the MDSplus I/O and visualization tools, the RAPTOR suite can be used both offline, for a fast reconstruction of the plasma state, and in integrated control algorithms once it will be deployed in the JET real-time data network.

The termination of high performance plasmas in tokamak devices with high Z metal plasma facing components presents challenges related to the influx of heavy impurities which, if not kept under control, cause an increase of the radiative losses, radiative cooling and high probability of disruption. A number of key players in these dynamics have been identified by intensive research performed after the first years of operation in machines as AUG and JET in preparation of ITER operation. Inward neoclassical convection related to the peaking of the density profile, poloidal asymmetries, plasma rotation and centrifugal effects, temperature screening, pedestal temperature, pedestal density and ELMs control are among them. The objective of D-T fuelled plasmas with high neutron yield in stationary conditions, foreseen in the near future at JET, focuses the operations towards high performance in terms of thermal energy content and plasma current and consequently with higher disruption risk. The reduction of such risks is being pursued for the specific features of the two plasma scenarios being developed, baseline (βN ~1.8, q95 ~ 3) and hybrid (βN ~2–3, q95~4). The analysis of the previous experimental campaign and the data so far collected in the present campaign indicate that the combination of edge and core W control is needed to obtain a safe plasma termination, with the optimized use of the available actuators: gas and pellet for ELMs control, ramp-down waveform of the NBI heating power while maintaining a relevant ICRH additional power, sweeping of the separatrix hitting point on the divertor to reduce the heat load and to decrease the W source.

This paper summarizes the physical principles behind the novel three-ion scenarios using radio frequency waves in the ion cyclotron range of frequencies (ICRF). We discuss how to transform mode conversion electron heating into a new flexible ICRF technique for ion cyclotron heating and fast-ion generation in multi-ion species plasmas. The theoretical section provides practical recipes for selecting the plasma composition to realize three-ion ICRF scenarios, including two equivalent possibilities for the choice of resonant absorbers that have been identified. The theoretical findings have been convincingly confirmed by the proof-of-principle experiments in mixed H–D plasmas on the Alcator C-Mod and JET tokamaks, using thermal 3He and fast D ions from neutral beam injection as resonant absorbers. Since 2018, significant progress has been made on the ASDEX Upgrade and JET tokamaks in H–4He and H–D plasmas, guided by the ITER needs. Furthermore, the scenario was also successfully applied in JET D–3He plasmas as a technique to generate fusion-born alpha particles and study effects of fast ions on plasma confinement under ITER-relevant plasma heating conditions. Tuned for the central deposition of ICRF power in a small region in the plasma core of large devices such as JET, three-ion ICRF scenarios are efficient in generating large populations of passing fast ions and modifying the q-profile. Recent experimental and modeling developments have expanded the use of three-ion scenarios from dedicated ICRF studies to a flexible tool with a broad range of different applications in fusion research.

Operation of a magnetic fusion experiment, such as JET, relies on the availability of real-time (RT) control schemes, which supervise the plasma as it approaches the expected target performance while maintaining the integrity of the machine and its subsystems. At JET, there have been a series of recent efforts since (Lennholm M. et al 2017 Fusion Engineering and Design 123 535?540) to develop and test RT control schemes in preparation for the upcoming Deuterium-Tritium (DT) campaign. When operating JET in DT, each plasma discharge will in fact be a precious resource, being both T and neutron budget limited. Among the developed control schemes, this paper deals with the isotope ratio controller, which will maintain the required 50:50 DT ratio needed to favor nuclear fusion processes; the dud detector (L. Piron et al 2019 Fusion Engineering and Design 146 1364?1368), which will terminate a discharge moving toward controllers for detecting excessive radiation. Moreover, brandnew detectors, also based on machine learning approaches, have been implemented for detecting off-normal events or pre-disruptive states and have been included in the Plasma Event TRiggering and Alarms system (C. I. Stuart et al 2020 SOFT conference). Work is also ongoing to deploy into JET the RAPTOR suite, a RT observer for plasma state monitoring (C. Piron et al 2020 SOFT conference), and to identify control schemes within RAPTOR capabilities, which could contribute to support the development of high performance plasma scenarios.

A study of mixed hydrogen-deuterium H-mode plasmas has been carried out in JET-ILW to strengthen the physics basis for extrapolations to JET D-T operation and to support the development of strategies for isotope ratio control in future experiments. Variations of input power, gas fuelling and isotopic mixture were performed in H-mode plasmas of the same magnetic field, plasma current and divertor configuration. The analysis of the energy confinement as a function of isotope mixture reveals that the biggest change is seen in plasmas with small fractions of H or D, in particular when including pure isotope plasmas. To interpret the results correctly, the dependence of the power threshold for access to type-I ELMing H-modes on the isotope mixture must be taken into account. For plasmas with effective mass between 1.2 and 1.8 the plasma thermal stored energy (W-th) scales as m(eff)(0.1), which is weaker than that in the ITER physics basis, IPB98 scaling. At fixed stored energy, deuterium-rich plasmas feature higher density pedestals, while the temperature at the pedestal top is lower, showing that at the same gas fuelling rate and power level, the pedestal pressure remains constant with an exchange of density and temperature as the isotope ratio is varied. Isotope control was successfully tested in JET-ILW by changing the isotope ratio throughout a discharge, switching from D to H gas puffing. Several energy confinement times (300 ms) are needed to fully change the isotope ratio during a discharge.

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 the MGI prior to disruption. Long lasting LM (>= 100 ms) do exist prior to disruption in 75% of cases. However, 10% of non-disruptive pulses have a LM which eventually vanished without disruption. The plasma current quench (CQ) may result in 3D configurations, termed as asymmetrical disruptions, which are accompanied by sideways forces. Unmitigated vertical displacement events (VDEs) generally have significant plasma current toroidal asymmetries. Unmitigated non-VDE disruptions also have large plasma current asymmetries presumably because there is no plasma vertical position control during the CQ and so they too are subject to large vertical displacements. MGI is a reliable tool to mitigate 3D effects and correspondingly sideways forces during the CQ. The vessel structure loads depend on the force impulse and force time behaviour, including their rotation. The toroidal rotation of 3D configuration may cause resonance with the natural frequencies of the vessel components in large tokamaks such as ITER. The JET-ILW amplitude-frequency interdependence of toroidal rotation of 3D configurations is presented.

Operations using deuterium-tritium (DT) mixtures are envisaged in JET device in the forthcoming 2020 experimental campaign. These experiments will offer a unique possibility to study several open issues in support to ITER and DEMO, such as alpha particle heating, and to improve the high plasma performance obtained in the previous DT campaign Keilhacker et al. [1]. During DT operation, each plasma discharge will be a precious resource, being both T and neutron budget limited. It will be thus mandatory to promptly detect and safely terminate those plasma discharges which do not achieve the expected target parameters. A real-time detector of underperforming discharges has been developed for this purpose and it is named the dud detector. The dud detector calculates and monitors the time evolution of plasma performance indicators, which can be used to trigger an alarm and a proper plasma termination. The experience gained on the algorithm's behavior during DD operation will allow the use of this tool during DT operations and can be used as a guide to design dud detectors for ITER and future fusion reactors.

Post-disruption runaway electron (RE) beams in tokamaks with large current can cause deep melting of the vessel and are one of the major concerns for ITER operations. Consequently, a considerable effort is provided by the scientific community in order to test RE mitigation strategies. We present an overview of the results obtained at FTU and TCV controlling the current and position of RE beams to improve safety and repeatability of mitigation studies such as massive gas (MGI) and shattered pellet injections (SPI). We show that the proposed RE beam controller (REB-C) implemented at FTU and TCV is effective and that current reduction of the beam can be performed via the central solenoid reducing the energy of REs, providing an alternative/parallel mitigation strategy to MGI/SPI. Experimental results show that, meanwhile deuterium pellets injected on a fully formed RE beam are ablated but do not improve RE energy dissipation rate, heavy metals injected by a laser blow off system on low-density flat-top discharges with a high level of RE seeding seem to induce disruptions expelling REs. Instabilities during the RE beam plateau phase have shown to enhance losses of REs, expelled from the beam core. Then, with the aim of triggering instabilities to increase RE losses, an oscillating loop voltage has been tested on RE beam plateau phase at TCV revealing, for the first time, what seems to be a full conversion from runaway to ohmic current. We finally report progresses in the design of control strategies at JET in view of the incoming SPI mitigation experiments.

This paper presents results of JET ITER-like wall L-mode experiments in hydrogen and deuterium (D) plasmas, dedicated to the study of the isotope dependence of ion heat transport by determination of the ion critical gradient and stiffness by varying the ion cyclotron resonance heating power deposition. When no strong role of fast ions in the plasma core is expected, the main difference between the two isotope plasmas is determined by the plasma edge and the core behavior is consistent with a gyro-Bohm scaling. When the heating power (and the fast ion pressure) is increased, in addition to the difference in the edge region, also the plasma core shows substantial changes. The stabilization of ion heat transport by fast ions, clearly visible in D plasmas, appears to be weaker in H plasmas, resulting in a higher ion heat flux in H with apparent anti-gyro-Bohm mass scaling. The difference is found to be caused by the different fast ion pressure between H and D plasmas, related to the heating power settings and to the different fast ion slowing down time, and is completely accounted for in non-linear gyrokinetic simulations. The application of the TGLF quasi-linear model to this set of data is also discussed.

The present paper offers an overview of the potential of ion cyclotron resonance heating (ICRH) or radio frequency heating for the DEMO machine. It is found that various suitable heating schemes are available. Similar to ITER and in view of the limited bandwidth of about 10 MHz that can be achieved to ensure optimal functioning of the launcher, it is proposed to make core second harmonic tritium heating the key ion heating scheme, assisted by fundamental cyclotron heating He-3 in the early phase of the discharge; for the present design of DEMO-with a static magnetic field strength of B-o = 5.855 T-that places the T and 3He layers in the core for f = 60 MHz and suggests centering the bandwidth around that main operating frequency. In line with earlier studies for hot, dense plasmas in large-size magnetic confinement machines, it is shown that good single pass absorption is achieved but that the size as well as the operating density and temperature of the machine cause the electrons to absorb a non-negligible fraction of the power away from the core when core ion heating is aimed at. Current drive and alternative heating options are briefly discussed and a dedicated computation is done for the traveling wave antenna, proposed for DEMO in view of its compatibility with substantial antenna-plasma distances. The various tasks that ICRH can fulfill are briefly listed. Finally, the impact of transport and the sensitivity of the obtained results to changes in the machine parameters is commented on.

The need for predictive capabilities greater than 95% with very limited false alarms are demanding requirements for reliable disruption prediction systems in tokamaks such as JET or, in the near future, ITER.The prediction of an upcoming disruption must be provided sufficiently in advance in order to apply effective disruption avoidance or mitigation actions to prevent the machine from being damaged.In this paper, following the typical machine learning workflow, a generative topographic mapping (GTM) of the operational space of JET has been built using a set of disrupted and regularly terminated discharges.In order to build the predictive model, a suitable set of dimensionless, machine-independent, physics-based features have been synthesized, which make use of 1D plasma profile information, rather than simple zero-D time series.The use of such predicting features, together with the power of the GTM in fitting the model to the data, obtains, in an unsupervised way, a 2D map of the multi-dimensional parameter space of JET, where it is possible to identify a boundary separating the region free from disruption from the disruption region.In addition to helping in operational boundaries studies, the GTM map can also be used for disruption prediction exploiting the potential of the developed GTM toolbox to monitor the discharge dynamics.Following the trajectory of a discharge on the map throughout the different regions, an alarm is triggered depending on the disruption risk of these regions.The proposed approach to predict disruptions has been evaluated on a training and an independent test set and achieves very good performance with only one tardive detection and a limited number of false detections.The warning times are suitable for avoidance purposes and, more important, the detections are consistent with physical causes and mechanisms that destabilize the plasma leading to disruptions.

The First Mirror Test for ITER has been carried out in JET with mirrors exposed during: (i) the third ILW campaign (ILW-3, 2015-2016, 23.33 h plasma) and (ii) all three campaigns, i.e. ILW-1 to ILW-3: 2011-2016, 63,52 h in total. All mirrors from main chamber wall show no significant changes of the total reflectivity from the initial value and the diffuse reflectivity does not exceed 3% in the spectral range above 500 nm. The modified layer on surface has very small amount of impurities such as D, Be, C, N, O and Ni. All mirrors from the divertor (inner, outer, base under the bulk W tile) lost reflectivity by 20-80% due to the beryllium-rich deposition also containing D, C, N, O, Ni and W. In the inner divertor N reaches 5 x 10(17) cm(-2), W is up to 4.3 x 10(17) cm(-2), while the content of Ni is the greatest in the outer divertor: 3.8 x 10(17) cm(-2). Oxygen-18 used as the tracer in experiments at the end of ILW-3 has been detected at the level of 1.1 x 10(16) cm(-2). The thickness of deposited layer is in the range of 90 nm to 900 nm. The layer growth rate in the base (2.7 pm s(-1)) and inner divertor is proportional to the exposure time when a single campaign and all three are compared. In a few cases, on mirrors located at the cassette mouth, flaking of deposits and erosion occurred.

For the first time, the JET operation in deuterium-tritium (D-T) plasma, which is scheduled to take place on JET in 2020, will be performed in the ITER mix of plasma-facing component materials.In view of the preparation of the DT campaign (DTE2), several aspects of the plasma operation require significant improvements, such as a real-time protection of the first wall.The risk of damaging the metallic PFCs caused by beryllium melting or cracking of tungsten owing to thermal fatigue required a new reliable D-T compatible active protection system.Therefore, the future development of the JET real time first wall protection is focused on the D-T campaign and the ITER relevant conditions which may cause failure of camera electronics within the Torus hall.In addition to the technological aspect, the intensive preparation of the diverse software tools and real time algorithms for hot spot detection as well as alarm handling strategy required for the wall protection is in progress.This contribution describes the improved design, implementation, and operation of the near infrared (NIR) imaging diagnostic system of the JET-ILW plasma experiment and its integration into the existing JET protection architecture.To provide the reliable wall protection during the DTE2, two more sensitive logarithmic NIR camera systems equipped with new optical relays to take images and cameras outside of the biological shield have been installed on JET-ILW and calibrated with an in-vessel calibration light source (ICLS).Additionally, post-pulse data visualization and advanced analysis of all types of imaging data is provided by the new software framework JUVIL (JET users video imaging library).The formation of hot spots is recognized as a significant threat due to rapid surface temperature rise.Because it could trigger the protection system to stop a pulse, it is important to identify the mechanisms and conditions responsible for the formation of such hot spots.To address this issue the new software tool 'Hotspot Editor' has been developed.