The Divertor Tokamak Test facility (DTT) is a Controlled Thermonuclear Fusion experiments which is in the scope of the Eurofusion programme to investigate the effects of thermal loads on the divertor. DTT will be realized at the Enea Research Center in Frascati, Italy and in the first experimental phase it will exploit an Electron Cyclotron Resonant Heating system (ECRH) exploiting 16 radio frequency (RF) sources (Gyrotrons) rated for 1MW power. These are fed in couple by a total of 8 high voltage power supply set (HVPS) based on the ITER design which procurement will be launched in 2023. Each HVPS is composed of one Main High Voltage Power Supply (MPS), rated for 55kV 110A, and two Body power Supplies (BPSs), rated for 35kV 20mA and this work will describe the power supply requirements, the conceptual design of the DTT HVPSs system, the layout and interfaces with the main systems.

The Divertor Tokamak Test (DTT) facility, whose construction has started in Frascati (Italy), will be equipped with an ECRH (electron cyclotron resonance heating) system including 32 gyrotrons as microwave power sources. The procurement of the first batch of sources with 16 MW total power, based on 170 GHz/>= 1 MW/100 s vacuum tubes, is in progress and will be available for the first DTT plasma. The system is organized into four clusters of 8 gyrotrons each. The power is transmitted from the Gyrotron Hall to the Torus Hall Building (THB) by a quasioptical transmission line (TL), mainly composed of large mirrors shared by eight beams coming from eight different gyrotrons and designed for up to 1.5 MW power per single beam, similar to the TL installed at the stellarator W7-X. One of novelties introduced in the DTT system is that the mirrors of the TLs are embodied in a vacuum enclosure, using large metal seals, mainly to avoid air absorption and risk of arcs. The main reason is to reduce the risk of air breakdown, maintaining a pressure of 10-5 mbar far away from the Paschen minimum. The TL estimated volume is between ~70 and ~85 m 3. The direct connection of the TL to the tokamak vacuum vessel has been evaluated, and different solutions have been proposed in order to prevent a possible impact on DTT operations. The microwave power is injected into the tokamak using independent single-beam front-steering launchers, real-time controlled and located in the equatorial and upper ports of four DTT sectors. In-vessel piezoelectric walking drives are the most promising candidates for the launcher mirror movement considering their compactness and capability to operate in an environment with strong magnetic field under ultra-high vacuum. The DTT ECRH system design, presented here, is based mainly on existing and assessed solutions, although the challenging adaptations to the DTT case are considered.

GyM is a linear plasma device operating at Istituto per la Scienza e Tecnologia dei Plasmi, Consiglio Nazionale delle Ricerche, Milan, with the original aim of studying basic plasma physics, such as turbulent processes. Since 2014, GyM experimental program has been mainly focused on the issue of plasma-material interaction (PMI) for magnetic confinement nuclear fusion applications. GyM consists of a stainless steel vacuum chamber (radius and length of 0.125 m and 2.11 m), a pumping system, a gas injection system, 10 magnetic field coils and two magnetron sources at 2.45 GHz, capable of delivering a total microwave power up to 4.5 kW. Highly reproducible steady-state plasmas of different gas species, at a maximum working pressure of ~10-1 Pa, can be obtained by electron cyclotron resonance heating in the resonance layer at 87.5 mT. Plasmas of GyM have electron and ion temperature <=15 eV and ~0.1 eV, respectively. The electron density is in the range of 1015-1017 m-3 and the ion flux is <=5 × 1020 ions?m-2s-1. Main plasma diagnostics of GyM comprise Langmuir probes, an optical emission spectrometer, a mass spectrometer and a fast camera system equipped with an image intensifier unit. For the purpose of investigating the topic of PMI, GyM is provided with two sample exposure systems. Both are biasable at a negative bias voltage down to -400 V, to tune the energy of the impinging ions. One of them is also equipped with a heating lamp and can reach and sustain a temperature of 990 K for several hours, thus allowing to study the role of sample temperature during the plasma-material interaction. This contribution presents the layout of GyM, the diagnostics, the sample exposure systems and the typical plasma parameters. A brief overview of the main PMI activities carried out so far and a description of future machine upgrades are also given.

The European DEMO is a pulsed device with pulse length of 2 hours. The functions devoted to the heating and current drive system are: plasma breakdown, plasma ramp-up to the flat-top where fusion reactions occur, the control of the plasma during the flat-top phase, and finally the plasma ramp-down. The EU-DEMO project was in a Pre-Concept Design Phase during 2014-2020, meaning that in some cases, the design values of the device and the precise requirements from the physics point of view were not yet frozen. A total of 130 MW was considered for the all phases of the plasma: in the flat top, 30 MW is required for neoclassical tearing modes (NTM) control, 30 MW for burn control, and 70 MW for the control of thermal instability (TI), without any specific functions requested from each system, Electron Cyclotron (EC), Ion Cyclotron (IC), or Neutral Beam (NB) Injection. At the beginning of 2020, a strategic decision was taken, to consider EC as the baseline for the next phase (in 2021 and beyond). R&D on IC and NB will be risk mitigation measures. In parallel with progresses in Physics modelling, a decision point on the heating strategy will be taken by 2024. This paper describes the status of the R&D development during the period 2014-2020. It assumes that the 3 systems EC, IC and NB will be needed. For integration studies, they are assumed to be implemented at a power level of at least 50 MW. This paper describes in detail the status reached by the EC, IC and NB at the end of 2020. It will be used in the future for further development of the baseline heating method EC, and serves as starting point to further develop IC and NB in areas needed for these systems to be considered for DEMO.

Disruptions represent one of the highest concerns for next-step fusion devices based on the tokamak principle. Active disruption avoidance and off-normal event handling strategies need to be envisaged and carefully designed in modern Plasma Control Systems (PCS) to monitor and predict when the plasma approaches operational boundaries. In the recent years, in the context of TCV internal and EUROfusion WPTE framework programs, several real-time control algorithms for disruption avoidance and prevention, such as for NTM stabilization and control as well as for H-mode density limit (HDL) active avoidance, have been successfully developed and embedded in the TCV real-time plasma supervision system (SAMONE). An NBI-heated scenario for the HDL has been newly developed and tested in a large number of experiments carried out for different plasma currents and divertor baffles configurations, allowing to reproduce the same physics characteristic phenomenology observed also in other devices. A first demonstration of the concept of portability across different devices has been achieved by transferring from AUG to TCV the same control algorithm based on the distance with respect to an empirically defined disruption boundary in the space of the H-Mode confinement factor (H98y,2) and a normalized line integrated edge electron density. Such a distance metrics, combined with a deep learning model for energy confinement-state detection, allows to react in real-time activating different control tasks regulating NBI power and gas flux with the objective of recovering from the strong confinement degradation observed when approaching the density limit. This contribution will present an overview of the experimental results, modelling activities supporting density limit scalings, advances in algorithms for detection of proximity to operational limits as well as the advances in the development of a generic control architecture enabling the integration of active disruption avoidance strategies and exception handling.

The Divertor Tokamak Test (DTT) facility [1], whose construction is started in Frascati (Italy), will be equipped with an ECRH (Electron Cyclotron Resonance Heating) system of 32 gyrotrons. The procurement of the first batch of 16 MW, based on 170 GHz/>= 1 MW/100s vacuum tubes, is in progress and will be available for the first DTT plasma. The system is organized in 4 clusters of 8 gyrotrons each. The power is transmitted from the Gyrotron Hall to the Torus Hall Building by a Quasi-Optical transmission line (TL), mainly composed of large mirrors shared by 8 different beams and designed for up to 1.5 MW power per single beam, similar to the one installed at W7-X Stellarator. One of novelties introduced in DTT system is that the mirrors of the TLs are embodied in a vacuum enclosure, using large metal seals, to avoid air losses and microwave leaks. The TL estimated volume is between ~70 and ~85 m3 to be maintained at 10-5 mbar. The direct connection of the TL to the tokamak vessel vacuum has been evaluated and solutions proposed in order to prevent a possible impact on DTT operations. The microwave power is injected into the tokamak using independent single beam front-steering launchers, real-time controlled and located in the equatorial and upper ports of 4 DTT sectors. In-vessel piezoelectric walking drives are the most promising candidates for the launcher mirror movement because of their compactness in an environment with strong magnetic field under ultra-high vacuum. The DTT ECH system design, presented here, is based mainly on existing and assessed solutions, although the challenging adaptations to the DTT case are considered.

An initial concept for the plasma diagnostic and control (D&C) system has been developed as part of European studies towards the development of a demonstration tokamak fusion reactor (DEMO). The main objective is to develop a feasible, integrated concept design of the DEMO D&C system that can provide reliable plasma control and high performance (electricity output) over extended periods of operation. While the fusion power is maximized when operating near to the operational limits of the tokamak, the reliability of operation typically improves when choosing parameters significantly distant from these limits. In addition to these conflicting requirements, the D&C development has to cope with strong adverse effects acting on all in vessel components on DEMO (harsh neutron environment, particle fluxes, temperatures, electromagnetic forces, etc.). Moreover, space allocation and plasma access are constrained by the needs for first wall integrity and optimization of tritium breeding. Taking into account these boundary conditions, the main DEMO plasma control issues have been formulated, and a list of diagnostic systems and channels needed for plasma control has been developed, which were selected for their robustness and the required coverage of control issues. For a validation and refinement of this concept, simulation tools are being refined and applied for equilibrium, kinetic and mode control studies.

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.

Ionization current measured at the spark plug during combustion in spark ignition engines has often been proposed to determine the crank-angle at combustion pressure peak, namely the peak pressure angle, for the purpose of regulating spark timing to attain maximum brake torque (MBT). The proposal is based on the assumption that agreement exists between peak pressure angle and the angular position of the ionization current second peak, although no one has ever proved it by an appropriate statistical analysis. The aim of this work, for the first time and by rigorous statistical methods, is to prove the agreement between Peak Pressure Angle and Ionization Current Second Peak Angle (ICSPA), without which a MBT control via ICSPA would be ineffective. Our experimental database consisted of about 9000 pairs of Peak Pressure Angle and Ionization Current Second Peak Angle values corresponding to 90 different operating conditions of a spark ignition engine. A two-sample comparison was first carried out between mean values of Peak Pressure Angle and Ionization Current Second Peak Angle, which showed a statistically significant difference between them. Then Bland-Altman analysis (Lancet, 1986), widely known and used for checking agreement between two different measurement methods, was conducted. It demonstrated that under almost all the experimental operating conditions, there was no agreement between the Ionization Current Second Peak Angle and the Peak Pressure Angle.

For long-pulse tokamaks, one of the main challenges in the control strategy is to simultaneously reach multiple control objectives and to robustly handle in real-time (RT) unexpected events [off-normal-events (ONEs)] with a limited set of actuators. We have developed in our previous work a generic architecture of the plasma control system to deal with these issues. Due to this generic feature, we are able to extend it with an advanced supervisor: Supervisory control and Actuator Management with ONEs (SAMONE) to deal with multiple ONEs and multiple control scenarios in this work. We first standardize the evaluation of ONEs and, thereby, simplify significantly the supervisor decision logic, as well as facilitate the modifications and extensions of ONE states in the future. Then, we present the recent developments of real-time decision-making by the supervisor to switch between different control scenarios (normal, backup, shutdown, disruption mitigation, and so on) during the discharge based on ONE states. The developed SAMONE has been implemented on the TCV tokamak, applied to disruption avoidance with density limit experiments, demonstrating the excellent capabilities of the new RT integrated strategy.

A method is proposed to effectively obtain plasma parameters in low-density, low-temperature plasmas that are easily perturbed by Langmuir probes. The methodology is based on the perimeter sheath expansion method because it provides a simple, but effective interpretation of the Langmuir probe characteristics and, in particular, of the non-saturating trend of the ion saturation due to the expansion of the probe sheath. However, the perimeter sheath expansion method is proposed here in a modified version with a dual purpose: to permit the verification of the Child-Langmuir dependence of the sheath thickness from the potential and, most importantly, to properly assess the plasma potential. When probe perturbation is significant the proposed method provides an effective correction term for the plasma potential. The resulting sheath voltage is consistent in all analysed cases and in close agreement with the expected value for cylindrical probes in Ar. The method is applied to Ar plasmas here, but because no specific assumption was made on the atomic species, it may be applicable to other situations of interest.

The pre-conceptual layout for an electron cyclotron system (ECS) in DEMO is described. The present DEMO ECS considers only equatorial ports for both plasma heating and neoclassical tearing mode (NTM) control. This differs from ITER, where four launchers in upper oblique ports are dedicated to NTM control and one equatorial EC port for heating and current drive (H&CD) purposes as basic configuration. Rather than upper oblique ports, DEMO has upper vertical ports to allow the vertical removal of the large breeding blanket segments. While ITER is using front steering antennas for NTM control, in DEMO the antennas are recessed behind the breeding blanket and called mid-steering antennas, referred to the radially recessed position to the breeding blanket. In the DEMO pre-conceptual design phase two variants are studied to integrate the ECS in equatorial ports. The first option integrates waveguide bundles at four vertical levels inside EC port plugs with antennas with fixed and movable mid-steering mirrors that are powered by gyrotrons, operating at minimum two different multiples of the fundamental resonance frequency of the microwave output window. Alternatively, the second option integrates fixed antenna launchers connected to frequency step-tunable gyrotrons. The first variant is described in this paper, introducing the design and functional requirements, presenting the equatorial port allocation, the port plug design including its maintenance concept, the basic port cell layout, the transmission line system with diamond windows from the tokamak up to the RF building and the gyrotron sources. The ECS design studies are supported by neutronic and tokamak integration studies, quasi-optical and plasma physics studies, which will be summarized. Physics and technological gaps will be discussed and an outlook to future work will be given.

The Divertor Tokamak Test (DTT) facility [1], whose construction is starting, will study a suitable solution for the power exhaust in conditions relevant for the future fusion device DEMO. DTT can achieve the value of 15 MW/m for the divertor figure of merit PSEP/R by employing 45 MW of auxiliary heating power to the plasma. To achieve this goal, the selected heating systems are Electron Cyclotron Resonance Heating (ECRH), Ion Cyclotron Resonance Heating (ICRH) and Negative (ion based) Neutral Beam Injector (NNBI). The ECRH system relies on up to 32 gyrotrons (operating each at 170 GHz to supply from a minimum of 1MW to a maximum of 1.2 MW for 100 s), a Quasi Optical (QO) transmission line (TL), consisting of multi-beam mirrors installed under vacuum to reduce the overall transmission losses below the target of 10% and independent (single-beam) front-steering mirrors capable to direct the beams individually in real-time for assisted plasma breakdown, control of neoclassical tearing modes and sawtooth, ECCD and main electron heating. Although the ECRH system design presented here will be based mainly on existing and assessed technologies, like the 170 GHz gyrotron type developed for ITER and the QO TL installed at W7-X, challenging adaptations to the DTT case have to be made. In particular, the design of a QO TL under vacuum is novel and needs detailed analysis of the stray radiation along the line in order to set the requirements for the mirror dimensions and/or the cooling of the vacuum chamber that encloses the mirrors. A further relevant question is the reliability of the ECRH system: the development of automatic algorithms to control such a large number of gyrotrons is foreseen to provide the required amount and distribution of power into the plasma.

Next-generation tokamaks will require integrated plasma control solutions beyond the present state-of-the art. Such solutions are being actively explored on the TCV and ASDEX Upgrade tokamaks. A summary of recent achievements is provided in this paper. We report on advances in plasma state reconstruction algorithms including improved density profile estimations on ASDEX Upgrade and a first implementation of kinetic equilibrium reconstruction constrained by real-time transport equation solutions on TCV. Recent progress in development of generic plasma supervisory control system on TCV, including a mix of continuous control and off-normal event handling is discussed, together with its application to high-density limit disruption avoidance.The RABBIT code for real-time NBI deposition calculations has been integrated in the ASDEX Upgrade control system and quantities computed by the code have been used for real-time feedback control of the power deposited to the ions.Finally, we report on a first demonstration of model-based shot-to-shot iterative improvement of actuator trajectories,applied to control of the central ion and electron temperature by an appropriate mix of ECH and NBI.

Ionic Chromatography (IC) has been applied to absolute quantification of ammonia contained in the gas exhaust collected during dedicated experiments executed to study the nitrogen conversion in ammonia. o During the experiments residual RGA and OES diagnostics were both used to monitoring the gas species and radicals resulting from the plasma. o Results indicate that the nitrogen conversion never exceeds 10% and is affected by isotopic and wall material effects

TCV has a flexible, digital, distributed control system for testing experimental control algorithms, acquiring data from hundreds of diagnostic channels and controlling all magnetic, heating and fueling actuators. We present the state of the system, focusing on the latest upgrades, and the key control capabilities enabled by the system. The control algorithm code is developed and maintained in MATLAB/Simulink and run-time code is generated automatically using code generation. The previously used practice of just-in-time code generation and compilation before every shot has been abandoned in favor of a more reliable and efficient method where the run-time code is able to load parameters and waveforms from plant databases. The ability to simulate the control code is guaranteed by an object-oriented simulation framework in MATLAB/Simulink that reads parameters and waveforms from the same databases w.r.t. the real-time environment. This approach still allows very rapid development and deployment cycles with new algorithms deployed on TCV usually within a few days from the completion of their testing in simulation. The control algorithm software is managed through a DevOps methodology with extensive unit and regression tests as well as Continuous Integration / Deployment practices. The real-time environment has been completely replaced by the F4E MARTe2 framework, greatly improving standardization, modularity, maintainability and extensibility. The intrinsic data-driven application runtime buildup of the MARTe2 framework has naturally yet rigorously allowed the integration of the inter-shot tunable parameters and waveforms in the control code. The framework has also greatly enhanced interfaces between the real-time computers and the rest of TCV IT infrastructure, notably with its databases for shot configuration and control data acquisition. From the point of view of the hardware, the systems responsible for primary plasma controls (magnetic control and density control) have been upgraded with new ADC/DAC modules connected to two real time computers operable in parallel on the same discharge. This arrangement allows to use one control computer for the primary (released) main plasma controller while the second one can be used as a live test stand for plasma algorithms in state of testing or development. Also, a new EtherCAT real time industrial network has been laid down to operate distributed low I/O count subsystems boosting system flexibility at low additional cost and high speed of commissioning. This overhauling process has already granted a number of experimental advances on the machine, the foremost ones being: SAMONE a comprehensive real-time plasma supervision, off-normal event handling and actuator management system, plasma event detectors based on neural networks, novel linear controllers for improved vertical control for the formation and stabilization of doublet. Finally a number of existing realtime codes have already been ported to this new approach allowing them to be run seamlessly on every TCV discharge in real-time; notably they comprise RT-LIUQE, the real-time magnetic equilibrium reconstruction of TCV, coupled with real-time transport calculations; RT-MHD, the comprehensive real-time MHD analysis algorithms set and real time divertor radiation front control with multispectral 2D imaging diagnostics (MANTIS). Other applications include runaway and profile control.

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.