One of the main goals of the Divertor Tokamak Test (DTT) facility is to reach a ratio of power crossing the separatrix over the major radius of about 15 MW m - 1, as the one expected in DEMO. For this purpose, up to 45 MW of external additional heating power shall be coupled to the plasma, provided by Electron Cyclotron Resonance Heating (ECRH), Ion Cyclotron Resonance Heating and Neutral Beam Injection. The foreseen total ECRH power installed at full power shall be 32 MW, generated using 1 MW/170 GHz gyrotrons, for 100 s. The present ECRH system foresees two launching antennas per DTT sector, based on the front-steering concept. The equatorial antenna is dedicated to plasma heating and current drive and the upper antenna to the stabilization of MHD instabilities. This paper focuses on the latest design concept for these two antennas and on the definition of the ex-vessel matching optics unit of the last section of the evacuated Transmission Line (TL). The design has been updated to be compatible with the insertion of CVD diamond windows, to separate the vacuum environment of DTT from the one of the TL. This choice requires adding corrugated waveguide sections between the last mirror of the TL and the first mirror inside the port, requiring some adaptation of the in-vessel optics and of the supporting structure. The possibility to modify the steering range for the launching mirror has been also investigated to be compatible with the new design of the first wall and for the upper antenna, to reach the q = 2 surface in the new plasma scenario.

In this work the Electron Cyclotron (EC) physics performances of the EC system foreseen for the new Divertor Tokamak Test facility (DTT) are investigated using the beam tracing code GRAY on the flat top phase of the most recent DTT full power scenario. The whole core plasma region can be reached by EC beams with complete absorption, assuring bulk heating and core current drive (CD) for profile tailoring, and NTM mitigation in correspondence of the rational surfaces. A detailed analysis regarding modifications of the EC propagation, absorption and CD location due to density fluctuations caused by pellet injection is performed. The compatibility between the EC system and the pellet injection system is verified: the density variations due to pellet injection are foreseen to negligibly influence the EC performances, allowing the EC beams to reach the plasma central region for bulk heating and to drive current on the rational surfaces for NTM mitigation. Finally, the polarization variations originated by the angle steering foreseen for the operational and physics tasks accomplishment during the flat top phase of the discharge are assessed. Negligible power losses have been found keeping fixed polarization during the needed steering.

Deuterium plasma discharges of the Divertor Tokamak Test facility (DTT) in different operational scenarios have been predicted by a comprehensive first-principle based integrated modelling activity using state-of-art quasi-linear transport models. The results of this work refer to the updated DTT configuration, which includes a device size optimisation (enlargement to R0 = 2.19 m and a = 0.70 m) and upgrades in the heating systems. The focus of this paper is on the core modelling, but special attention was paid to the consistency with the scrape-off layer parameters required to achieve divertor plasma detachment. The compatibility of these physics-based predicted scenarios with the electromagnetic coil system capabilities was then verified. In addition, first estimates of DTT sawteeth and of DTT edge localised modes were achieved.

Designing a new tokamak requires concerted efforts of engineers and physicists. In order to reduce costs and minimise risks, a first-principle based integrated modelling as comprehensive as possible of plasma discharges in different operational scenarios is an essential tool. Therefore, main baseline scenarios of the future Divertor Tokamak Test facility (DTT) [1] (R0 = 2:19m, a = 0:70m,Wfirst wall and divertor, pulse length 100s, plasma current Ipl 5:5MA, vacuum toroidal field Btor 5:85T, total power by auxiliary heating systems Ptot 45MW) have been simulated extensively. This modelling work led to the optimisation of the device size and of the reference heating mix, as widely described in [2], and provided reference profiles for diagnostic system design, estimates of neutron yields, calculations of fast particle losses, gas puffing and/or pellet feature requirements for fuelling, MHD evaluations, and other tasks. The latest simulation results of the DTT scenarios with the Single Null magnetic configuration are presented here. These runs, carried out with the JINTRAC [3] suite or the ASTRA [4] transport solver, make use of theory based quasi-linear transport models (QLK [5] and TGLF SAT2 [6]), ensuring the highest fidelity presently achievable in integrated modelling. A specific attention to the consistency between the control coil system capabilities and plasma profiles has been paid and the edge requirements to have plasma scenarios compatible with divertor and first wall power handling capability and tungsten influx have been taken into account.

The scenario integrated modelling is a top priority work during the design of a new tokamak, as the Divertor Tokamak Test facility (DTT) under construction at the ENEA Research Center in Frascati. The first simulations of the main baseline scenarios contributed to the optimization of the DTT project, particularly with regard to the machine size and heating systems, besides serving as reference for diagnostics design. In this paper we report the first simulations of the full power baseline scenario in the final configuration of the machine and heating mix.

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.

Designing a new tokamak requires concerted efforts of engineers and physicists. In order to reduce costs and minimise risks, a first-principle based integrated modelling as comprehensive as possible of plasma discharges in different operational scenarios is an essential tool. Therefore, main baseline scenarios of the future Divertor Tokamak Test facility (DTT) [1] (R0 = 2:19m, a = 0:70m,Wfirst wall and divertor, pulse length 100s, plasma current Ipl 5:5MA, vacuum toroidal field Btor 5:85T, total power by auxiliary heating systems Ptot 45MW) have been simulated extensively. This modelling work led to the optimisation of the device size and of the reference heating mix, as widely described in [2], and provided reference profiles for diagnostic system design, estimates of neutron yields, calculations of fast particle losses, gas puffing and/or pellet feature requirements for fuelling, MHD evaluations, and other tasks. The latest simulation results of the DTT scenarios with the Single Null magnetic configuration are presented here. These runs, carried out with the JINTRAC [3] suite or the ASTRA [4] transport solver, make use of theory based quasi-linear transport models (QLK [5] and TGLF SAT2 [6]), ensuring the highest fidelity presently achievable in integrated modelling. A specific attention to the consistency between the control coil system capabilities and plasma profiles has been paid and the edge requirements to have plasma scenarios compatible with divertor and first wall power handling capability and tungsten influx have been taken into account.

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 Divertor Tokamak Test facility (DTT) [1,2] is a new tokamak under construction in Frascati, Italy, with metallic wall and focus on power and particle exhaust. Its main parameters are R=2.19 m, a=0.70 m, BT<=6T, Ip<=5.5 MA, Ptot<=45 MW, pulse length <= 100 s. An intensive integrated modelling work using the JINTRAC [3] and ASTRA [4] 1.5 D frameworks has been performed for the baseline day-0, day-1 and full power DTT scenarios with the Single Null divertor configuration. This has provided reference scenarios to support the design of the device, e.g. for the optimization of the heating mix and of the fuelling systems, for the diagnostics, neutron shield and control coil design, for neutron licensing, for MHD stability studies, for the evaluation of fast particle losses, as well as to help the elaboration of a DTT scientific research plan. The flat-top phase has been simulated using the quasi-linear transport models TGLF [5] and QuaLiKiz [6], with validation against gyrokinetics for the specific parameter regimes of DTT [7]. The pedestal has been predicted by EUROped [8]. The separatrix density and temperature were chosen to be consistent with scrape-off layer parameters in detached conditions. The chosen heating mix includes power on plasma ECH (170 GHz) 29 MW, negative NBI (510 keV) 10 MW, ICH (60-90 MHz) 6 MW. For ne/nGW~0.5, central Te~ 15 keV, Ti ~ 9 keV, ne0~2.5 1020 m-3 are predicted, with ?N~ 1.4 , ?E (P=Psep) ~ 0.5 s , H98Y~0.9, total neutron emission~1 1017 neutron/s, Zeff~1.6 with mildly peaked profile. Ar or Ne are envisaged as seeding gas. Rotation is rather small and does not provide significant ExB turbulence stabilization. Edge neutrals penetrate up to ?tor~0.8, with a required gas puff level which is marginal with respect to achievable pumping capacity, therefore the additional use of pellets for fuelling is recommended. A pellet system with injection from Oblique High Field side, pellets size r ~ 1 mm, velocity ~ 500 m/s is under design. Simulations with the QuaLiKiz model indicate that an injection frequency ~ 16 Hz is required to maintain the density pedestal for zero gas puff, with edge density perturbations < 20%. First attempts to simulate the complete time evolution of a full power discharge have been made using the 0.5 D METIS code [9], tuned to the quasi-linear simulations. The simulations make use of the plasma boundaries generated by the CREATE-NL free-boundary solver [10] and optimize the heating location and timing to be consistent with the prescribed time evolution of ? and internal inductance li , with a current ramp lasting 27 s and L-H transition at t=32.8 s. The use of off-axis ECH during the current ramp is envisaged to keep li low. Time dependent simulations with ASTRA using quasi-linear models for the core and the IMEP [11] edge model are ongoing to provide a physics based prediction of the discharge evolution.

An intensive integrated modelling work of the main scenarios of the new Divertor Tokamak Test (DTT) facility with a single null divertor configuration has been performed using first principle quasi-linear transport models, in support of the design of the device and of the definition of its scientific work programme. First results of this integrated modelling work on DTT (R0 = 2.14 m, a = 0.65 m) are presented here along with outcome of the gyrokinetic simulations used to validate the reduced models in the DTT range of parameters. As a result of this work, the heating mix has been defined, the size of device has been increased to R0 = 2.19 m and a = 0.70 m, the use of pellets for fuelling has been recommended and reference profiles for diagnostic design, estimates of neutron yields and fast particle losses have been made available.

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.

In the European Roadmap towards thermonuclear fusion power production, studying the controlled exhaust of energy and particles from a fusion reactor is a top priority research item. This is the main goal of the Divertor Tokamak Test (DTT) facility, a D-shaped superconducting tokamak (R = 2.19 m, a = 0.70 m, BT <= 6 T, Ip <= 5.5 MA, pulse length <= 100 s, auxiliary heating <= 45 MW, W first wall and divertor), whose construction is starting in Frascati. In order to support the device design and to help the elaboration of a DTT scientific work-programme, it is a key priority to achieve multi-channel integrated modelling of DTT scenarios based on state-of-art first-principle quasi-linear transport models. First modelling results of the main DTT scenarios are presented here. Steady-state profiles of ion and electron temperatures, densities, rotation, and current density were predicted with a calculated self-consistent equilibrium, with turbulent heat and particle transport calculated by the TGLF or QLK transport models, and with heating modelled self-consistently. As a result of this work, the heating mix was defined and reference profiles have been become available.

The construction of the Divertor Tokamak Test facility (DTT)[1,2], a new D-shaped superconducting tokamak (R=2.19m, a=0.70m, BT<= 6T, Ip <= 5.5MA, pulse length <= 100s, auxiliary heating <= 45MW, W first wall and divertor), is starting in Frascati, Italy. Its main mission is to study the controlled exhaust of energy and particle from a fusion reactor, which is a top priority research item in the European Roadmap[3] towards thermonuclear fusion power production. The characteristics of the machine will allow to address many ITER and DEMO relevant physics issues besides plasma wall interaction in a fusion relevant range of plasma parameters. In order to support the device design, and particularly the heating mix definition, the design of the neutron shields, of diagnostic systems, and of pellet injectors, and the assessment of fast particle losses, as well as to help the elaboration of a DTT scientific work-programme, it is a key priority to achieve multi-channel integrated modelling of DTT scenarios based on state-of-art first principle quasi-linear transport models. The integrated simulations of main DTT scenarios, carried out with the JINTRAC[4] suite or with the ASTRA[5] transport solver, cover the region inside the separatrix and predict steady-state profiles of ion and electron temperature, density, rotation, current density, impurity (Ar, W) density, with heating and current drive modelled self-consistently and with a calculated self-consistent equilibrium starting from a fixed boundary taken from [6]. Inside the top of the pedestal, the turbulent heat and particle transport is calculated by the Trapped-Gyro-Landau-Fluid (TGLF)[7] or QuaLiKiz (QLK)[8] quasi-linear transport models. First results of this modelling work are presented and for some cases these results are validated against GENE[9] gyrokinetic simulations with the specific DTT parameters, to corroborate the validity of the reduced models in the particular case of DTT. As a result of this work, the heating mix was defined. Particularly, in the full power scenario, an extrìernal power of about 45MW will be provided from the 3 heating auxiliary systems to the plasma: ~30MW from the 170 GHz ECRH system, ~10MW from the NBI system composed by one injector at 510keV, and ~6MW from the 60-90MHz ICRH system to the plasma. In the full power scenario, central temperatures of ~20 keV for electrons and ~10keV for ions with central densities ~2.5x1020m-3 are predicted in fair agreement by the two models used. Moreover, reference profiles for diagnostic design, estimates of neutron yields and fast particle losses have become available.

The use of the Electron Cyclotron Resonance Heating and Current Drive (ECRH&CD) system is considered essential for granting reliable plasma operation in future fusion reactors. In this work we present the quantitative study of the performances of the EC system in the main scenario foreseen for the first planned reactor DEMO (DEMO1 [1]), focusing on the following topics: (i) optimization of the CD efficiency and of the localization of the power and current deposition for the physics tasks regarding bulk heating (BH), CD and Neoclassical Tearing Modes (NTM) mitigation in the flat top phase; and (ii) first evaluations about the feasibility of the injection and absorption of significant amount of EC power in the plasma density pedestal region for Thermal Instability (TI) control. To accomplish these tasks, the EC system must provide proper different radial localizations and combination of heating and non-inductive CD, taking into account the peculiar physics requirements and the engineering constraints of a fusion power plant reactor. In order to find suitable launcher configurations, the beam tracing code GRAY [2] has been used to perform scans in the launcher parametric space defined by the injection angles, the wave frequency and the antenna position.