COMPASS addressed several physical processes that may explain the behaviour of important phenomena. This paper presents results related to the main fields of COMPASS research obtained in the recent two years, including studies of turbulence, L-H transition, plasma material interaction, runaway electron, and disruption physics: Tomographic reconstruction of the edge/SOL turbulence observed by a fast visible camera allowed to visualize turbulent structures without perturbing the plasma. Dependence of the power threshold on the X-point height was studied and related role of radial electric field in the edge/SOL plasma was identified. The effect of high-field-side error fields on the L-H transition was investigated in order to assess the influence of the central solenoid misalignment and the possibility to compensate these error fields by low-field-side coils. Results of fast measurements of electron temperature during ELMs show the ELM peak values at the divertor are around 80% of the initial temperature at the pedestal. Liquid metals were used for the first time as plasma facing material in ELMy H-mode in the tokamak divertor. Good power handling capability was observed for heat fluxes up to 12 MW m(-2) and no direct droplet ejection was observed. Partial detachment regime was achieved by impurity seeding in the divertor. The evolution of the heat flux footprint at the outer target was studied. Runaway electrons were studied using new unique systems-impact calorimetry, carbon pellet injection technique, wide variety of magnetic perturbations. Radial feedback control was imposed on the beam. Forces during plasma disruptions were monitored by a number of new diagnostics for vacuum vessel (VV) motion in order to contribute to the scaling laws of sideways disruption forces for ITER. Current flows towards the divertor tiles, incl. possible short-circuiting through PFCs, were investigated during the VDE experiments. The results support ATEC model and improve understanding of disruption loads.

Runaway electrons (RE) are typically generated during plasma disruptions in tokamaks and they represent a serious issue for the integrity of the device [1]. Several strategies have been proposed to control and suppress RE beams in tokamak plasmas, such as the use of massive gas injection, pellet injection or use of resonant magnetic perturbations [2]. An alternative strategy to control the RE beam energy was proposed, which relies on the resonant interaction between high energy electrons and whistler waves [3]. The mechanism of destabilization of electromagnetic waves by a RE beam was proposed in 2006 [4], but a first direct observation of RE-driven whistler waves in a tokamak plasma was performed only in 2018 in DIII-D [5] and a detailed study of radiofrequency emissions in presence of a RE beam was carried out more recently in FTU [6]. Here a model for the description of plasma waves destabilization in presence of a RE beam in COMPASS is proposed. Two different situations are considered: the one in which the waves are spontaneously generated inside the RE beam and the one in which the waves are injected from the outside by an antenna. Wave propagation is calculated using the ray-tracing method [7]; the linear growth rate of the wave is calculated using analytical formulas [8] . Multiple reflections of the wave inside the plasma are followed and the positions of maximum wave amplification are identified. Considerations on the optimal wave injection strategy are made.

A large superconducting machine, JT-60SA has been constructed to provide major contributions to the ITER program and DEMO design. For the success of the ITER project and fusion reactor, understanding and development of plasma controllability in ITER and DEMO relevant higher beta regimes are essential. JT-60SA has focused the program on the plasma controllability for scenario development and risk mitigation in ITER as well as on investigating DEMO relevant regimes. This paper summarizes the high research priorities and strategy for the JT-60SA project. Recent works on simulation studies to prepare the plasma physics and control experiments are presented, such as plasma breakdown and equilibrium controls, hybrid and steady-state scenario development, and risk mitigation techniques. Contributions of JT-60SA to ITER and DEMO have been clarified through those studies.

Construction of the JT-60SA tokamak was completed on schedule in March 2020. Manufacture and assembly of all the main tokamak components satisfied technical requirements, including dimensional accuracy and functional performances. Development of the plasma heating systems and diagnostics have also progressed, including the demonstration of the favourable electron cyclotron range of frequency (ECRF) transmission at multiple frequencies and the achievement of long sustainment of a high-energy intense negative ion beam. Development of all the tokamak operation control systems has been completed, together with an improved plasma equilibrium control scheme suitable for superconducting tokamaks including ITER. For preparation of the tokamak operation, plasma discharge scenarios have been established using this advanced equilibrium controller. Individual commissioning of the cryogenic system and the power supply system confirmed that these systems satisfy design requirements including operational schemes contributing directly to ITER, such as active control of heat load fluctuation of the cryoplant, which is essential for dynamic operation in superconducting tokamaks. The integrated commissioning (IC) is started by vacuum pumping of the vacuum vessel and cryostat, and then moved to cool-down of the tokamak and coil excitation tests. Transition to the super-conducting state was confirmed for all the TF, EF and CS coils. The TF coil current successfully reached 25.7 kA, which is the nominal operating current of the TF coil. For this nominal toroidal field of 2.25 T, ECRF was applied and an ECRF plasma was created. The IC was, however, suspended by an incident of over current of one of the superconducting equilibrium field coil and He leakage caused by insufficient voltage holding capability at a terminal joint of the coil. The unique importance of JT-60SA for H-mode and high-beta steady-state plasma research has been confirmed using advanced integrated modellings. These experiences of assembly, IC and plasma operation of JT-60SA contribute to ITER risk mitigation and efficient implementation of ITER operation.

The tokamak a configuration variable (TCV) continues to leverage its unique shaping capabilities, flexible heating systems and modern control system to address critical issues in preparation for ITER and a fusion power plant. For the 2019-20 campaign its configurational flexibility has been enhanced with the installation of removable divertor gas baffles, its diagnostic capabilities with an extensive set of upgrades and its heating systems with new dual frequency gyrotrons. The gas baffles reduce coupling between the divertor and the main chamber and allow for detailed investigations on the role of fuelling in general and, together with upgraded boundary diagnostics, test divertor and edge models in particular. The increased heating capabilities broaden the operational regime to include T (e)/T (i) similar to 1 and have stimulated refocussing studies from L-mode to H-mode across a range of research topics. ITER baseline parameters were reached in type-I ELMy H-modes and alternative regimes with 'small' (or no) ELMs explored. Most prominently, negative triangularity was investigated in detail and confirmed as an attractive scenario with H-mode level core confinement but an L-mode edge. Emphasis was also placed on control, where an increased number of observers, actuators and control solutions became available and are now integrated into a generic control framework as will be needed in future devices. The quantity and quality of results of the 2019-20 TCV campaign are a testament to its successful integration within the European research effort alongside a vibrant domestic programme and international collaborations.

Since the 2018 IAEA FEC Conference, FTU operations have been devoted to several experiments covering a large range of topics, from the investigation of the behaviour of a liquid tin limiter to the runaway electrons mitigation and control and to the stabilization of tearing modes by electron cyclotron heating and by pellet injection. Other experiments have involved the spectroscopy of heavy metal ions, the electron density peaking in helium doped plasmas, the electron cyclotron assisted start-up and the electron temperature measurements in high temperature plasmas. The effectiveness of the laser induced breakdown spectroscopy system has been demonstrated and the new capabilities of the runaway electron imaging spectrometry system for in-flight runaways studies have been explored. Finally, a high resolution saddle coil array for MHD analysis and UV and SXR diamond detectors have been successfully tested on different plasma scenarios.

In the last runaway electron campaigns that were performed in 2020, broad range of experi- mental results were obtained in several areas of runaway electron physics. Special diagnostic tools and methods were applied including HXR spectrometry, limiter calorimetry, estimates of energy based on runaway electron equilibrium, synchrotron radiation diagnostics (REIS2), matrix SXR detectors and passive and active high frequency antennas. The measurements of RE beam average energy based on the relativistic pressure and equilibrium properties seem to present a fast and efficient method that can be compared well with the methods based on the hard X-ray bremsstralung and synchrotron radiation diagnostics. The comparison of RE beams generated by injections of different gas species and amounts shows that the least dangerous terminations of the RE beam can be achieved via Ne injection or impurity injection followed by a D2 secondary injection. The D2 secondary injection beneficial effects tend to be consistent with DIII-D and JET results. The RMP experiments have also provided many interesting results including significant reduction of HXRs and impact energy of the RE beam with application of RMPs during the RE beam generation. Measurements with high frequency antennas, which will be briefly introduced, also showed presence of wide range of high frequency kinetic instabilities including the whistler-like and various chirping phenomena. Last but not least the room temperature solid state pellet injector has been implemented and used for RE beam generation as well as for enhancing the RE dissipation.

The TCV tokamak is augmenting its unique historical capabilities (strong sh aping, strong electron heating) with ion heating, additional electron heating compatible with high densities, and variable divertor geometry, in a multifaceted upgrade program desig ned to broaden its operational range without sacrificing its fundamental flexibility. The TCV program is rooted in a three-pronged approach aimed at ITER support, explorations towards DEMO, and fundamental research. A 1-MW, tangential neutral beam injector (NBI) was recently installed and promptly extended the TCV parameter range, with record ion temperatures and toroidal rotation velocities and measurable neutral-beam current drive. ITER-relevant scenario development has received particular attention, with strategies aimed at maximizing performance through optimized discharge tr ajectories to avoid MHD instabilities, such as peeling-ballooning and neoclassical tearing modes. Experiments on exhaust physics have focused particularly on detachment, a necessary step to a DEMO reactor, in a comprehensive set of conventional and advanced divertor concepts. The specific theoretical prediction of an enhanced radiation region between the two X -points in the low-field-side snowflake-minus configuration was experimentally confirmed. Fundamental investigations of the power decay length in the scrape -off layer (SOL) are progressing rapidly, again in widely varying configurations and in both D and He plasmas; in par ticular, the double decay length in L-mode limited plasmas was found to be replaced by a single length at high SOL resistivity. Ex periments on disruption mitigation by massive gas injection and electron-cyclotron resonance heating (ECRH) have begun in earnest, in parallel with studies of runaway electron generation and control, in both stable and disruptive conditions; a quiescent runaway beam carrying the entire electrical current appears to develop in some cases. Developments in plasma control have benefited from progress in individual controller design and have evolved steadily towards controller integration, mostly within an environment supervised by a tokamak profile control simulator. TCV has demonstrated effective wall conditioning with ECRH in He in support of the preparations for JT-60SA operation.

Radiofrequency emission in the 0.4-3 GHz range from FTU tokamak in presence of runaway electrons (REs) has been measured in various plasma regimes. Rapid emission bursts associated with enhanced RE pitch-angle scattering reveal kinetic instabilities affecting evolution of the RE population from the buildup phase. Such measurements also provide a sensitive monitor for instabilities during early RE formation. The leading edge of radio bursts is much shorter than interleaving periods of low emission; spectral broadening during growth indicates nonlinear wave coupling as an explanation for the observed intermittency. Both broadband and coherent spectra have been observed. Radio emission disappears at the beginning of post-disruption RE plateaus, and subsequently reappears in the shape of very intense bursts accompanied by macroscopic magnetic perturbations.

Results from the last FTU campaigns on RE mitigation strategies for quiescent and post-disruption RE beams are presented. We provide experimental evidence that for some RE quiescent scenarios D2 solid pellets achieve complete RE suppression capability, mainly due to the induced burst MHD activity expelling RE seed, whereas in other cases we report clear indications of avalanche multiplication of RE. Results on the assimilation of solid deuterium pellets on RE quiescent scenarios are provided. Quantitative indications of dissipative effects of anomalous Doppler instabilities (ADI) and MHD activity, in terms of critical electric field increase, is introduced and supported by experimental evidence. Multiple analysis are provided to show the significant energy conversion/dissipation of large ADI on post-disruption RE beams suggesting new strategies for RE energy suppression. We also demonstrate experimentally that modulated ECRH could be used for ADI pacing.

The COMPASS tokamak provided unique conditions to study Runaway Electrons (RE) physics in the ITER-like geometry. Twelve experimental campaigns were dedicated to identify limiting thresholds for safe termination or efficient suppression of RE beam [1] and to thoroughly explore the possible advantages of alternative mitigation strategies. High flexibility and reproducibility of COMPASS RE scenarios allowed the execution of extensive scans devoted, for example, to decoupling effects of individual factors acting on RE dynamics. The studied factors include resonant magnetic field perturbations (RMP), radiation drag, instabilities, and the newly developed unique RE radial position feedback [2]. The main principle of the RE radial feedback was successfully used as a fast and robust method to obtain information about the RE energy in parallel to HXR radiation and unique calorimetry measurements. These methods are used to show the impact of different mitigation materials, externally triggered RMPs and the feedback itself on the RE beam energy and its current decay rate. The experimentally observed effects of RMPs are supported by a full orbit RE tracer developed at COMPASS, MARS-F [3] and ORBIT [4] codes. First tests with an active launch of low power 0.5 GHz waves into a RE populated low density plasma were performed. Preliminary results from measurements of electromagnetic waves in the 0.1-1.5 GHz range during this scenario are presented.

Analyses of experimental data collected in the last FTU campaign provide interesting results on Runaway Electrons (REs) suppression by means of large (wrt FTU plasma volume) deuterium pellets on RE quiescent discharges, mainly inducing bursts of MHD activity that expel the RE seeds. This phenomenon was found to be very reproducible on discharges at 0.5 MA and 5.3 T. Pellets injected during current ramp-ups are also shown to lead to a complete suppression of the RE seed; however, they triggered disruptions as well. These promising results on pellet cleaning effects for different RE quiescent scenarios are discussed. Avalanche multiplication of REs after single pellet injection on 0.36 MA RE quiescent discharges is reported. We provide quantitative indications of RE mitigation effects in terms of increased Connor-Hastie (critical) electrical field due to Anomalous Doppler Instabilities (ADI) for RE quiescent scenarios and post-disruption RE beams. Analysis of large fan-like instabilities on post-disruption RE beams, that seem correlated to the applied negative electrical field and the background density drops, revealed their strong capability to dissipate the RE energy increasing their pitch angle, expelling some of them, and transferring a large fraction of their magnetic energy to the background plasma. Then, ADI can lead to efficient RE energy dissipation (on FTU) as well as full conversion from runaway into thermal electrons (on TCV), indicating a new possible strategy for RE mitigation by controlling the onset of large fan-like instabilities. We also show that short pulses of ECRH, combined with D2 gas puffing, generate a density increase that gives way to a density decrease, when ECRH is switched off, triggering ADI and opening the path to a possible alternative RE mitigation strategy.

Latest results on quiescent and post-disruption runaway electrons mitigation experiments at Frascati Tokamak Upgrade.

The high field, high density tokamak FTU closed its 30-years of operation at the end of 2019. FTU is a circular machine (R0=0.93 m, a=0.29 m) with an Inconel Vacuum Vessel, Ni and Fe being its dominant elements, and Mo poloidal and toroidal limiters. The relatively high plasma densities, in combination with baking and boronization conditioning techniques, have ensured the possibility of producing plasmas characterized by an extremely low level of impurities of any kind, thus making FTU especially well-suited for investigating non-intrinsic impurities and the performances of liquid metal limiters under high thermal loads (up to 18 MW/m2). Initial tests were performed with a Lithium Liquid Limiter, while the more recent experiments have explored the plasma behavior with a Tin Liquid Limiter (TLL). Both are based on the innovative Capillary Porous System [1]. Lithium contamination was considerable, and traces can occasionally still be seen on various spectroscopic diagnostics. Oxygen is hardly present, and C is also low; N is detected at times, while He, Ne and Argon are detected when injected for diagnostic purposes.

The Divertor Tokamak Test (DTT) facility will be built to study a solution to the issue of power exhaust in conditions relevant for DEMO. The Italian DTT tokamak, by coupling to plasma up to 45 MW of additional power, will reach the needed condition of power flow to the divertor of 15 MW/m. The selected Heating Systems to achieve this goal are Electron Cyclotron Heating (ECH), Ion Cyclotron Heating (ICH) and Negative Neutron Beam Injector (NNBI). The power will be installed in two stages: a day-1 configuration with a coupled power of 25 MW and a second step where the completion of the 45 MW will be realized in 4 years from the day-1. At first stage 16 MW of ECH power and 4 MW of ICH will be installed, making the DTT plasma dominated by RF heating. The EC system is based on 170 GHz, 1 MW gyrotron, while for the transmission line a Quasi Optical approach has been chosen, with the feature to install the multibeam mirrors (8 beams on each one) under vacuum. The goal is to reduce the overall losses at ~10% avoiding atmospheric absorption and selecting the proper polarization for the longest section. The power will be injected into the tokamak using front steering individual antennas and capable to real time steer all the beams for the tasks assigned to EC waves. The first module of the ICH systems will be based on transmitters, capable of a wide frequency range (60-90 MHz), connected, though standard coaxial cables and RF components, to two movable antennas inserted in the equatorial ports of DTT. The selected range is done to exploit different heating schemes. The choice of the antenna type will be based on reliability (i.e. power density) rather than on its performance in term of peak coupled power. This led to choose a two-strap antenna with a power density of 3.5 MW/m2, shaped to fit the DTT scrape-off plasma and with an adjustable radial position. An external matching system is envisaged to cope with fast variation of antenna loading, e.g. due to edge localized modes.

A research activity in the Institute for Plasma Science and Technology of National Research Council (ISTP-CNR, Italy, former IFP-CNR) and in L.T. Calcoli (LTC, Italy) is aimed at the design and construction of calorimetric loads for absorption and measurement of high power millimeter-waves in the electron cyclotron frequency range. Recently, two CW 170 GHz loads, one for the European ITER gyrotron test facility and the other for the FALCON test-bed, have been installed at the Swiss Plasma Center (SPC, Switzerland). One short pulse (2 s) load for 1 MW, designed and optimised to operate at two different frequencies (84 GHz and 126 GHz), was provided for testing and conditioning two new dual-frequency gyrotrons for the Tokamak à Configuration Variable (TCV, Switzerland). Two additional CW loads, designed for absorbing powers higher than 1 MW, have been delivered to the National Institutes for Quantum and Radiological Science and Technology (QST, Japan) and exploited for the acceptance tests and the conditioning of the prototype of the Japanese ITER gyrotron. The present status and the most recent experimental results achieved in the framework of this development activity are reported in the paper.

In the frame of the TCV Tokamak upgrade, two 84/126 GHz/2 s dual frequency gyrotrons designed by SPC and KIT and manufactured by THALES will be added to the existing EC-System. The first unit has been delivered to EPFL-SPC and tested. In the commissioning configuration, a matching optics unit (MOU) is connected to the gyrotron window. The RF is then coupled to the HE11 mode of a 63.5mm corrugated waveguide and dissipated in a load procured by CNR after 4m of waveguide and 2 miter bends. Owing to the flexible triode gun design giving the possibility to adjust the pitch angle parameter, the specifications were met at both frequencies. At 84 GHz (TE17,5 mode), a power of 0.930 MW was measured in the calorimeter, with a pulse duration of 1.1 s. At the high frequency (126 GHz, TE26,7 mode), a power of 1.04 MW was reached for a pulse length of 1.2 s. Accounting for the load reflection and the ohmic losses in the various subcomponents of the transmission line and the tube, it is estimated that the output power at the gyrotron window is in excess of 1 MW at both frequencies, with an electronic efficiency of 32% and 34% at 84 GHz and 126 GHz respectively. The gyrotron behavior is remarkably robust and reproducible, and the pulse length is limited by external systems that will be improved shortly.