The electron cyclotron heating (ECH) launchers for the Divertor Tokamak Test (DTT) facility are currently being designed. The aim is to inject power in different plasma locations with high flexibility to allow plasma current ramp-up and ramp-down, central heating with tailored deposition profiles, and neoclassical tearing modes (NTMs) stabilization. Two antennas are developed, one for the equatorial port with six beam lines and one for the upper port with two beam lines, with similar front-steerable mirror modules. Each beam-line module is composed of a cooled corrugated waveguide launching a beam toward a fixed shaping mirror (M1) and then to the plasma-facing plane mirror (M2), movable around two axes, both cooled to sustain the high heat load due to the ohmic losses of the incident 1 MW beam and the plasma radiation. Thermal loads also include plasma and microwave stray radiation. Main structural loads on the mirrors and their supports, including the launcher supporting baseplate, are forces and torques due to the induced electro-magnetic (EM) currents, during normal operation and disruptions, interacting with the high magnetic field in which the mirrors and supports are immersed. Thermal and structural challenges, in particular on the M1 and M2 mirrors, have opposite solutions that require a delicate trade-off: optimal cooling requires high-thermal-conductivity materials such as copper alloys, while low-electrically-conductive materials are needed to have lower induced currents during disruptions. Solutions under study for the mirror cooling for minimal deformations and stresses, and for drive mechanisms, are presented. The possible reduction of induced currents during disruptions using different materials, supporting structures, and layouts is discussed.

European DEMOnstration (EU-DEMO) fusion reactor is currently progressing through its conceptual design phase. Reliability assessments play a key role in demonstrating that the system architecture can fulfill its intended functions over a defined lifespan. In the current approach, availability assessments begin with a failure mode and effects analysis (FMEA) to systematically identify potential failure events affecting system performance. Fault tree (FT) models are developed for subsystems exhibiting multiple failure modes, while the overall system configuration is modeled using a reliability block diagram (RBD). This study focuses on the electron cyclotron (EC) heating system, which supports plasma heating during ramp-up and ramp-down phases, stabilizes neoclassical tearing modes, and mitigates radiative instabilities. The analysis targets the current design of the transmission line (TL) and launcher subsystems, excluding gyrotrons from the current scope. Availability simulations have been performed over a 20-year operational period. Parametric analyses are also conducted to quantify the sensitivity of availability to maintenance duration and restoration parameters, providing key insights into design robustness and operational planning.

Abstract This paper reports the main design keys and the challenging issues of the Heating and Current Drive System (HCD) of the Divert Tokamak Test (DTT) facility that is under construction at ENEA site in Frascati with the aim to perform studies on the power exhaust in a flexible and easily modifiable environment. The selected HCD systems for DTT are those with the most consolidated technology and expected to be relevant for the future reactor. The status of each system is reported, both in terms of design and procurement, which are well advanced for the system required in the first phase of the DTT exploitation: electron and ion cyclotron resonant heating. The third system is neutral beam injector, based on negative ion acceleration, which will be installed in DTT in a second phase, after the first five years of operation. The full heating power will be reached with the doubling of the radiofrequency power to reach the 45 MW at plasma needed to test the divertor with a power density at reactor level.

Tokamak a configuration variable (TCV), recently celebrating 30 years of near-continual operation, continues in its missions to advance outstanding key physics and operational scenario issues for ITER and the design of future power plants such as DEMO. The main machine heating systems and operational changes are first described. Then follow five sections: plasma scenarios. ITER Base-Line (IBL) discharges, triangularity studies together with X3 heating and N2 seeding. Edge localised mode suppression, with a high radiation region near the X-point is reported with N-2 injection with and without divertor baffles in a snowflake configuration. Negative triangularity (NT) discharges attained record, albeit transient, beta(N) similar to 3 with lower turbulence, higher low-Z impurity transport, vertical stability and density limits and core transport better than the IBL. Positive triangularity L-Mode linear and saturated ohmic confinement confinement saturation, often-correlated with intrinsic toroidal rotation reversals, was probed for D, H and He working gases. H-mode confinement and pedestal studies were extended to low collisionality with electron cyclotron heating obtaining steady state electron iternal transport barrier with neutral beam heating (NBH), and NBH driven H-mode configurations with off-axis co-electron cyclotron current drive. Fast particle physics. The physics of disruptions, runaway electrons and fast ions (FIs) was developed using near-full current conversion at disruption with recombination thresholds characterised for impurity species (Ne, Ar, Kr). Different flushing gases (D2, H2) and pathways to trigger a benign disruption were explored. The 55 kV NBH II generated a rich Alfvenic spectrum modulating the FI fas ion loss detector signal. NT configurations showed less toroidal Alfven excitation activity preferentially affecting higher FI pitch angles. Scrape-off layer and edge physics. gas puff imaging systems characterised turbulent plasma ejection for several advanced divertor configurations, including NT. Combined diagnostic array divertor state analysis in detachment conditions was compared to modelling revealing an importance for molecular processes. Divertor physics. Internal gas baffles diversified to include shorter/longer structures on the high and/or low field side to probe compressive efficiency. Divertor studies concentrated upon mitigating target power, facilitating detachment and increasing the radiated power fraction employing alternative divertor geometries, optimised X-point radiator regimes and long-legged configurations. Smaller-than-expected improvements with total flux expansion were better modelled when including parallel flows. Peak outer target heat flux reduction was achieved (>50%) for high flux-expansion geometries, maintaining core performance (H-98 > 1). A reduction in target heat loads and facilitated detachment access at lower core densities is reported. Real-time control. TCV's real-time control upgrades employed MIMO gas injector control of stable, robust, partial detachment and plasma beta feedback control avoiding neoclassical tearing modes with plasma confinement changes. Machine-learning enhancements include trajectory tracking disruption proximity and avoidance as well as a first-of-its-kind reinforcement learning-based controller for the plasma equilibrium trained entirely on a free-boundary simulator. Finally, a short description of TCV's immediate future plans will be given.

An overview is presented of the progress since 2021 in the construction and scientific programme preparation of the Divertor Tokamak Test (DTT) facility. Licensing for building construction has been granted at the end of 2021. Licensing for Cat. A radiologic source has been also granted in 2022. The construction of the toroidal field magnet system is progressing. The prototype of the 170 GHz gyrotron has been produced and it is now under test on the FALCON facility. The design of the vacuum vessel, the poloidal field coils and the civil infrastructures has been completed. The shape of the first DTT divertor has been agreed with EUROfusion to test different plasma and exhaust scenarios: single null, double null, X-divertor and negative triangularity plasmas. A detailed research plan is being elaborated with the involvement of the EUROfusion laboratories.

The European DEMO (EU-DEMO) reactor studies within EUROfusion aim to develop a fusion power plant concept. The large tokamak device needs an auxiliary heating power which, at the present stage, is provided by the Electron Cyclotron (EC) heating system with up to 130 MW foreseen to reach different regions of plasma for heating, suppression of instabilities and the possibility to support ramp-up and ramp-down phases. The present conceptual design of the system is based on 2 MW coaxial-cavity gyrotron sources, a transmission line (TL) using both circular corrugated waveguides and quasi-optical evacuated multi-beam TLs, and mirror antennas located in the Equatorial Port. In order to create a modular system, the sources are grouped in ‘clusters’, whose powers are combined in the quasi-optical TL, up to the tokamak building, where they are split and routed as single waveguides. In the launcher, they are combined together again on the launching mirrors, to save space for the apertures in the Breeding Blanket. The present EC heating system has a certain flexibility to adapt to changing design guidelines. The development status of the system is presented.

The Divertor Tokamak Test (DTT) facility is a controlled thermonuclear fusion device, whose construction has started at the ENEA Research Center in Frascati, Italy, with the aim to study solutions to mitigate the issue of power exhaust in an integrated environment and operating conditions relevant for future fusion reactors. A power of 45 MW to the plasma is foreseen and provided by a mix of heating and current drive (HCD) systems comprising electron cyclotron resonance heating (ECRH), ion cyclotron resonance heating (ICRH), and neutral beam injector (NBI). DTT, in the first experimental phase, will exploit the ECRH system, including 16 radio frequency (RF) sources (gyrotrons) with potential collector depressed, each rated for 1 MW power, 170 GHz frequency, and 100 s pulse length. The RF sources are fed in pairs by a high voltage power supply (HVPS) set, composed of one main power supply (MPS) and two body power supplies (BPSs), one for each gyrotron, and by low-voltage (LV) power supplies dedicated to the operation of the auxiliaries of the gyrotrons. This work describes the conceptual design of the HVPS system, the evaluation that led to the definition of its main requirements and to the selection of a reference design. In addition, the novel concept is presented for a dummy load (DL) able to mimic the real gyrotrons operation in normal and abnormal conditions for testing an HVPS set.

The EU DEMO Tokamak is foreseen to be equipped with an electron cyclotron (EC) system for plasma heating, magnetohydrodynamic (MHD) control, and thermal instability suppression. Up to six launchers will be installed into equatorial ports with the aim to inject a maximum of 130 MW millimeter wave power at frequencies of 136/170/204 GHz toward dedicated positions into the plasma. This article presents the current layout of the typical EC system launcher components within the available space reservation areas in the equatorial level of the DEMO baseline model. Beside the general arrangement of the EC launcher components, preliminary design features of the launcher port plug modules, an active cooling system approach, integration concepts for mirrors and waveguides (WGs) of the optical system, the shielding component formation, and safety important elements of the first containment barrier of the launcher are presented. Furthermore, relevant aspects of assembly and potential remote maintenance (RM) procedures are discussed.

Within the 9th European Framework programme, since 2021 EUROfusion is operating five tokamaks under the auspices of a single Task Force called ‘Tokamak Exploitation’. The goal is to benefit from the complementary capabilities of each machine in a coordinated way and help in developing a scientific output scalable to future largre machines. The programme of this Task Force ensures that ASDEX Upgrade, MAST-U, TCV, WEST and JET (since 2022) work together to achieve the objectives of Missions 1 and 2 of the EUROfusion Roadmap: i) demonstrate plasma scenarios that increase the success margin of ITER and satisfy the requirements of DEMO and, ii) demonstrate an integrated approach that can handle the large power leaving ITER and DEMO plasmas. The Tokamak Exploitation task force has therefore organized experiments on these two missions with the goal to strengthen the physics and operational basis for the ITER baseline scenario and for exploiting the recent plasma exhaust enhancements in all four devices (PEX: Plasma EXhaust) for exploring the solution for handling heat and particle exhaust in ITER and develop the conceptual solutions for DEMO. The ITER Baseline scenario has been developed in a similar way in ASDEX Upgrade, TCV and JET. Key risks for ITER such as disruptions and run-aways have been also investigated in TCV, ASDEX Upgrade and JET. Experiments have explored successfully different divertor configurations (standard, super-X, snowflakes) in MAST-U and TCV and studied tungsten melting in WEST and ASDEX Upgrade. The input from the smaller devices to JET has also been proven successful to set-up novel control schemes on disruption avoidance and detachment.

The Divertor Tokamak Test (DTT) facility is equipped with auxiliary heating systems in order to be able to load the divertor with a power flux relevant to study the power exhaust issue in a reactor relevant range of parameter. The powerful system is the Electron Cyclotron Heating (ECH) with an installed power of 32 MW in its largest extension. Together with the bulk heating of the DTT plasma, the ECH system will cover several tasks for the plasma operation. This paper summarizes the main characteristics and design choices of the DTT ECH system and the related physics studies, based on the reference DTT plasma, to develop and control the plasma, fulfilling the functional tasks, with the support of simulation activities. Dedicated studies have been carried out to investigate the capability of EC power to assist plasma start-up, stabilize MHD activity and support current ramp up/down. In addition, it has been studied how changes of the ECH power distribution can have an impact on the plasma profiles, affecting the fueling pellet effectiveness and MHD modes.

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 10 15 –10 17 m −3 and the ion flux is ≤5 × 10 20 ions⋅m −2 s −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 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 m3. 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.

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.

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.

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.

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.

The tokamak à 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 ∼ 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.

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.

Within the European development of a future tokamak demonstration fusion power plant (DEMO) [1] the pre-conceptual studies on the plasma diagnostic and control (D&C) system are progressing to prepare the basis for reliable plasma operation at high overall performance [2]. A variety of plasma diagnostics will be employed on DEMO together with advanced control techniques in order to provide an accurate knowledge of the plasma state, which is needed to maintain plasma operation within the allowed physical and technical limits. The integration of diagnostic front-end components has to cope with strong adverse effects arising from neutron and gamma irradiation, heat loads, impinging particles and forces. In this environment, the quality of measurements can only be ensured for longer periods by using robust diagnostic components, mounting them in sufficiently protected (retracted) locations, and any maintenance can only be performed via remote handling. Major open issues are the durability of magnetic measurements in the presence of irradiation induced effects and the feasibility of detachment control under DEMO conditions. In parallel to diagnostic developments, the details of the main control issues are being formulated and investigated by quantitative plasma control simulations. To obtain the envisaged performance DEMO operates close to some physics limits where even small disturbances, if not properly controlled, can trigger major variations of the plasma parameters. Equilibrium control requires high control power and can drive the poloidal field coil system to its operational limits. Within this paper, we will provide an overview on the current status of the ongoing D&C developments for the European DEMO concept.

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.