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.

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.

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.

for supporting scenario preparation and plasma operation through a standardised data model designed to support both simulated and experimental data 1. The IMAS platform provides a high degree of modularity between physics models and physics workflows. One of the most sophisticated physics workflows developed so far in IMAS is for Heating and Current Drive (H&CD) modelling. The H&CD workflow developed by the ITER Organization is based upon an earlier development carried out within the European Integrated Modelling activities 2. It has been written in Python for its availability and portability, and its extensive use within the modelling community worldwide.

The inductive goal of ITER is to produce 500s long burning plasmas with Q = Pfus/Paux >=10[1]. This requires the development of operationally robust scenarios that span the whole plasma discharge from startup to termination not only in Deuterium Tritium (DT) but also in the Pre Fusion-Plasma Operation (PFPO) phase in Hydrogen (H) and Helium (He). In the PFPO phase, subsystems, such as the ELM mitigation system, will be commissioned and important lessons will be learnt about how to optimise and operate ITER plasmas within machine protection limits. As ITER's plasma facing surfaces (PFCs) are made of Beryllium (Be) and Tungsten (W), ITER operation will require applying the ITER heating and fuelling and impurity seeding systems in an optimum way to achieve the best plasma performance while ensuring low power fluxes and low erosion of the PFCs. In particular, the optimisation will include: i) minimising the release of tungsten by plasma-wall interactions; ii) controlling tungsten transport into the core plasma to avoid accumulation; iii) acceptable divertor power loads (<10MWm-2); iv) tolerable Neutral Beam (NB) shine-though loads; and in the Fusion-Plasma Operation (PFO) phase also v) the control of the DT mix in the core plasma. JINTRAC[2], developed by EUROfusion, is in a prime position to tackle this scenario development challenge with its suite of core (JETTO/SANCO/EDWM) and SOL/divertor (EDGE2D/EIRENE) transport codes that concurrently can simulate all these aspects.

Among the additional heating systems foreseen for the first planned reactor DEMO, which has the aim of demonstrating the production of electricity through controlled thermonuclear fusion reactions, the Electron Cyclotron Resonance Heating and Current Drive (ECRH&CD) system is considered essential for granting reliable plasma operation. In this work the quantitative study of EC system performances in the main DEMO operational scenario is presented, focusing on the physics tasks which such system is foreseen to accomplish: bulk heating, current drive and mitigation of MHD instabilities in the core of the plasma, and control of thermal instability in the plasma edge region. To this aim, the EC system must provide proper different radial localizations and combination of heating and non-inductive CD. In order to find suitable launcher configurations, the beam tracing code GRAY has been used to perform scans in the launcher parametric space defined by the injection angles, the wave frequency and the antenna position. Main optimization results are shown, taking into account the peculiar physics requirements and issues, and the engineering constraints of a fusion power plant reactor.

To evaluate the capability of electron cyclotron emission (ECE) diagnostic in the Chinese Fusion Engineering Testing Reactor (CFETR), ECE spectra have been simulated using the code SPECE. The results indicate that measurements of the 2nd harmonic X-mode and the 1st harmonic O-mode ECE spectra combined are capable of providing the information of local electron temperature with fairly good spatial resolution and adequate radial range covering from the plasma core to the edge. Followed with the simulation results, the conceptual design of the front-end, transmission line, instrumentation, and the calibration strategy are discussed briefly. Base on the state-of-the-art techniques, no foreseen technical obstacles exist. However, it is worthy to study the relativistic down-shift effect in the high temperature plasmas in order to obtain the position of the emission layer with required precision in a timely manner.

The optimum conditions for access to and sustainment of H-mode plasmas and their expected plasma parameters in the pre-fusion power operation 1 (PFPO-1) phase of the ITER research plan, where the additional plasma heating will be provided by 20 MW of electron cyclotron heating, are assessed in order to identify key open R&D issues. The assessment is performed on the basis of empirical and physics-based scalings derived from present experiments and integrated modelling of these plasmas including a range of first-principle transport models for the core plasma. The predictions of the integrated modelling of ITER H-mode plasmas are compared with ITER-relevant experiments carried out at JET (low-collisionality high-current H modes) and ASDEX Upgrade (significant electron heating) for both global H-mode properties and scale lengths of density and temperature profiles finding reasonable agreement. Specific integration issues of the PFPO-1 H-mode plasma scenarios are discussed taking into account the impact of the specificities of the ITER tokamak design (level of ripple, etc).

The ITER Research Plan envision operation around half of the nominal magnetic field (i.e. around B = 2.65 T) as a path to baseline operation. This work discusses constraints on the optimal range of magnetic field, which is bounded in the lower limit by the presence of the third-harmonic electron cyclotron resonance at half field, and on the upper limit by the loss of core heating and current drive. It will be shown that increasing the magnetic field by only 3%, i.e. to 2.75 T, eliminates the third harmonic parasitic absorption without compromising demonstration of access to H-mode, while operating at a magnetic field of 3.0 T - previously proposed for optimal use of the ion cyclotron system - would impair the use of the electron cyclotron system for core-heating and current drive. Operation at 2.65 T would still be possible if the polarization of the equatorial launcher is changed from X-mode to O-mode in the current flattop phase.

First plasma operation in ITER will start after completing the assembly of the tokamak vessel and the installation of the main sub-systems, but prior to the installation of the blanket modules and the divertor cassettes. Utilization of temporary limiters and divertor replacement structures will provide a poloidal and toroidal reference side position to the plasma edge to protect the vacuum vessel and other components already installed during operations. An additional set of mirrors is required to reflect the power injected from the upper launcher towards the plasma resonance for EC-assisted breakdown of the plasma up to a beam dump needed to trap and absorb the power of the beams after crossing the plasma in order to reduce the stray radiation escaping back into the vacuum chamber down to less than 10% of the total power. The quasi-optical system has been designed, with shape and size of the mirrors compliant with the requirements provided by ITER for their installation, realization and plasma performances, resulting in two standard focussing mirrors and one grating mirror. The beam dump consists of a box with five metal plates, the first providing a spreading of the high incident power and the others coated with absorbing material with thickness distribution studied to gradually reduce the power during the multiple reflections inside the box, avoiding damages to the coating itself. This work focuses on the validation of the quasi-optical design of the mirrors and the assessment of the dump performances, based on a multi-bounces model developed ad-hoc for this purpose. The study includes a tolerance analysis for the beam dump to include the effect of uncertainties in the thickness of the absorbing coating and misalignments of the mirrors, to verify the performances of the dump also when operating in different conditions with respect to the nominal ones.

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.