The Divertor Tokamak Test (DTT) is a new facility designed by Italian DTT Limited Liability Consortium (S.C.ar.l.) [1] aimed at validating an integrated solution for the power exhaust in support of DEMO [2]. The ICRH system, part of the DTT complex of additional heating systems, in its final configuration shall couple to the plasma up to 6 MW in the 60–90 MHz frequency range. All the main radiofrequency passive components of the system have been designed and optimized with the Ansys High Frequency Structures Simulator (HFSS) computer code. In particular 3 dB hybrid couplers, stubs, stretchers and vacuum feedthroughs have been examined. Results of the optimization of these components are presented and discussed in this paper.

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.

Since the last IAEA-FEC in 2021, significant progress on the development of long pulse steady state scenario and its related key physics and technologies have been achieved, including the reproducible 403 s long-pulse steady-state H-mode plasma with pure radio frequency (RF) power heating. A thousand-second time scale (similar to 1056 s) fully non-inductive plasma with high injected energy up to 1.73 GJ has also been achieved. The EAST operational regime of high beta(P) has been significantly extended (H-98y2 > 1.3, beta(P) similar to 4.0, beta(N) similar to 2.4 and n(e)/n(GW) similar to 1.0) using RF and neutral beam injection (NBI). The full edge localized mode suppression using the n = 4 resonant magnetic perturbations has been achieved in ITER-like standard type-I ELMy H-mode plasmas with q(95) approximate to 3.1 on EAST, extrapolating favorably to the ITER baseline scenario. The sustained large ELM control and stable partial detachment have been achieved with Ne seeding. The underlying physics of plasma-beta effect for error field penetration, where toroidal effect dominates, is disclosed by comparing the results in cylindrical theory and MARS-Q simulation in EAST. Breakdown and plasma initiation at low toroidal electric fields (<0.3 V m(-1)) with EC pre-ionization is developed. A beneficial role on the lower hybrid wave injection to control the tungsten concentration in the NBI discharge is observed for the first time in EAST suggesting a potential way toward steady-state H-mode NBI operation.

The intrinsic poor loading of Ion Cyclotron (IC) plasma-facing antennas makes the use of Tuning and Matching Systems (TMSs) a necessity. The antenna plus TMS is a resonant system; in the TMS and access lines high voltages (tens of kV) must be accounted for in the unavoidable unmatched part of the feeding lines. In this work, we propose and test an innovative type of IC launcher; it is based on achieving resonance of the self-standing antenna, i.e. without the TMS. A mechanical full-metal tuning mechanism is described and demonstrated to allow wide-band operation. A systematic analysis of possible antenna topologies has led to identifying a structure that can allow good impedance matching along with compliance with maximum electric field constraints. Most of the design is carried out using a simplified plasma and a commercial analysis tool and then validated with a realistic plasma using TOPICA code.

The paper reports the preliminary analysis of different typologies of ICRH launchers for chosing the most efficient solution for the ICRH system of the Divertor Tokamak Test facility (DTT), designed by the Italian DTT Limited Liability Consortium (S.C. a r.l.). In its final configuration this system will couple to the DTT plasma a nominal power of 6 MW in the 60 - 90 MHz frequency range by means of four launchers. This very preliminary analysis has been done with the ANSYS HFSS code.

An extensive linear analysis of the ICRH propagation and absorption in DTT full heating scenario has been performed by means of three advanced numerical tools: DISEMAG, FELICE and TORIC. The numerical codes solve respectively the full dispersion relation in the complex domain of the wavenumber (DISEMAG) and the integro-differential equation that accounts for the correct evaluation of the ICRH power absorption in a slab plasma (FELICE) and in the tokamak equilibrium configuration (TORIC). Moreover, by incorporating the antenna conceptual design, as released by the engineering design team, in FELICE and TORIC, the power spectrum, radiated by the antenna (3 straps) and coupled to the plasma, has been evaluated in the case of 60 and 90 MHz (3He and H minority heating respectively). By extensively using the abovementioned suite of codes, and after establishing the plasma target (fixing tokamak dimensions, density, temperatures, plasma current, magnetic field, isotopic composition, etc.), and the antenna characteristics (number, dimension, pitch, and radial position of the straps, antenna size, etc.), the power absorption on the various species (electrons, majority and minority ions) has been calculated as function of the minority concentration, parallel wavenumber, frequency, harmonic resonant layer, etc.. Comparison between the codes DISEMAG and TORIC in evaluating the absorption on electrons and minority ions has also been performed and shows a good agreement of the results. A new ion heating scheme based on three ions mixture has also been proposed in DTT.

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.

Hot spot is a serious challenge limiting long pulse operation with LHCD (lower hybrid current drive) in tokamak. In order to mitigate the hot spot in guard limiter and improve LHCD capability, effect of edge plasma density on inducing hot spot and current drive has been studied in EAST. The temperature in the guard limiter of the LH antenna, inducing hot spot directly, increases with edge density and LH power. Studies show that the hot spot is mainly ascribed to the heat flux in front of LH antenna. Further simulation indicates that such spot corresponds to the peak position of edge density due to local LH electric field. In addition, due to the stronger parametric instability (PI) behavior in the case of higher edge density, the current drive capability decreases with edge density. Strike-point splitting behaviors appear with density increase, in agreement with current profiles in the edge region and the reduction of total driven current, suggesting that more power is deposited in edge region, which does more contribution to hot spot. Above studies offer one possible idea to optimize the edge density so as to satisfy the coupling, mitigate the heat flux in the guard limiter, and improve current drive capability for fusion device.

In the first phase of the Divertor Tokamak Test Facility (DTT), the goal of coupling a RF power of 3 MW to the plasma will be reached with a single ICRH module constituted by 4 generators, 2 or 4 External Matching Units (EMU) and 2 antennas. The operation with high levels of RF power into ELMy H-mode plasmas requires a robust and reliable Load-Tolerant EMU. Currently the most appropriate solution, ensuring a VSWR at the generators under the safety threshold of 1.5:1, will be based on a Wideband 2-section coaxial Hybrid Coupler operating in the frequency range of 60−90 MHz. HFSS and CST-MS simulation software was used for the optimization of the characteristic impedances and lengths of the transversal and longitudinal branches aimed to achieve excellent performance (amplitude and phase balance at the output ports, and low reflection at the input port) over the entire frequency range. The analysis has been carried out for Zo = 50 Ω and 30 Ω coaxial transmission lines in terms of S-parameters and electric field. Several good solutions with an amplitude imbalance about ±0.25 dB and a return loss lower than −20 dB in a large part of the frequency range will be presented. Then a simulation of the overall (end-to-end) EMU (Wideband Hybrid Coupler with shifter, stub and service stub in each of the 2 output branches) exploiting a suitable circuital solver of Ansys Electronics tool of HFSS, has been implemented. The resilience of the EMU has been tested determining the VSWR in input and the amplitude balance at the outputs for a typical resistive and reactive loading variation occurring during ELMs at f = 75 MHz and compared with the results at 65 MHz and 85 MHz.

A simplified, flat, plasma-loaded, ion-cyclotron antenna has been simulated using three different modeling approaches and a comparison of simulation outputs is presented in this paper. Several plasma profiles with different density gradients and distances between the antenna and the cutoff density have been used as benchmark for the numerical simulations, which have been run at 30 MHz by means of two tools. The one is the TOrino Polythecnic Ion-Cyclotron Antenna (TOPICA) code that has been repeatedly used and validated for this type of problems. The other tool is the finite-element method (FEM) of a commercial software, where inhomogeneous anisotropic materials can be defined and the perfectly matched layer technique can be adopted. In this tool, two models of antenna load have been implemented: an equivalent dielectric, locally matching the perpendicular propagation constant of the fast wave in the plasma, and a cold plasma model. The simulation results by such models are compared with the results by TOPICA, which relies on a 1-D inhomogeneous, hot plasma.

In this study we are investigating the physics of ion-cyclotron resonance heating (ICRH) plasma interaction in the Divertor Tokamak Test facility, on the basis of the plasma and tokamak parameters characterizing the machine, as well as the antenna design. An assessment of the ICRH scenarios which involves (i) frequency choice, (ii) power spectrum, (iii) minority H and/or He-3 heating, (iv) deuterium second harmonic heating, (v) fast particles energies, has been carried out. Well assessed numerical tools have been used for the solution of the relevant electromagnetic wave equation coupled to the quasi-linear Fokker-Planck equation for the ion distribution function, and the Torino Polytechnic Ion Cyclotron Antenna has been used to calculate the launcher design and the coupling performance of the antenna and the wave spectrum in presence of a plasma load. The wave spectrum especially represents an important input parameter for the numerical codes calculating the propagation and absorption of the ICRH wave. In this work the numerical results are reported and discussed with particular attention to the determination of the distribution function of the accelerated ions and its implication in the fast particle physics.

The Divertor Tokamak Test (DTT) facility will be built to study a solution to the issue of power exhaust in conditions relevant for DEMO. The Italian DTT tokamak, by coupling to plasma up to 45 MW of additional power, will reach the needed condition of power flow to the divertor of 15 MW/m. The selected Heating Systems to achieve this goal are Electron Cyclotron Heating (ECH), Ion Cyclotron Heating (ICH) and Negative Neutron Beam Injector (NNBI). The power will be installed in two stages: a day(-1) configuration with a coupled power of 25 MW and a second step where the completion of the 45 MW will be realized in 4 years from the day-(1.) At first stage 16 MW of ECH power and 4 MW of ICH will be installed, making the DTT plasma dominated by RF heating. The EC system is based on 170 GHz, 1 MW gyrotron, while for the transmission line a Quasi Optical approach has been chosen, with the feature to install the multibeam mirrors (8 beams on each one) under vacuum. The goal is to reduce the overall losses at similar to 10% avoiding atmospheric absorption and selecting the proper polarization for the longest section. The power will be injected into the tokamak using front steering individual antennas and capable to real time steer all the beams for the tasks assigned to EC waves. The first module of the ICH systems will be based on transmitters, capable of a wide frequency range (60-90 MHz), connected, though standard coaxial cables and RF components, to two movable antennas inserted in the equatorial ports of DTT. The selected range is done to exploit different heating schemes. The choice of the antenna type will be based on reliability (i.e. power density) rather than on its performance in term of peak coupled power. This led to choose a two-strap antenna with a power density of 3.5 MW/m(2), shaped to fit the DTT scrape-off plasma and with an adjustable radial position. An external matching system is envisaged to cope with fast variation of antenna loading, e.g. due to edge localized modes.

L'energia prodotta dalla fusione nucleare nelle stelle ha reso possibile la vita cosi come oggi la conosciamo. Probabilmente, sui tempi lunghi, la sopravvivenza delle condizioni ambientali che permettono la vita sul nostro pianeta potra essere garantita dalla realizzazione di una fonte di energia basata sulla fusione. La sfida e molto complessa e richiede uno sforzo interdisciplinare al limite delle conoscenze attuali. Il nostro paese ha raccolto questa sfida fin dalla declassificazione della ricerca sulla fusione alla fine degli anni '50 del secolo scorso. In questi decenni si e costruita una rete di eccellenze che lega enti di ricerca, universita ed industrie ponendo il nostro paese in prima linea nella ricerca sulla fusione per capacita teoriche, sperimentali e realizzative.

Recently, experimental effect of density fluctuation in edge region on LHCD (lower hybrid current drive), another candidate related to parasitic effect, has been observed for the first time in EAST. Results show that density fluctuation is affected by RMP (resonant magnetic perturbation) application at density of 3.5x10 19 m -3 . The current drive capability indicated by the loop voltage improves with the decreasing density fluctuation. Meanwhile, the internal inductance enhances, indicating a peaked plasma current profile. Such degradation of LHCD at higher density fluctuation is mainly ascribed to the effect of density fluctuation in edge region on launching wave, which is firstly evidenced by the frequency spectrum measurement, leading to more power deposited in the edge region. Results are encouraging considering that the LHCD tool is essential for control of current profile in reactor grade plasmas.

The solution of the problem of heat exhaust has been pointed out as one of the main challenge towards the realization of magnetic confinement fusion. In the last years, two concepts have been proposed in alternative to the conventional divertor solution adopted for ITER: modification of the magnetic topology in the divertor region and liquid metal as plasma facing component. The role of the Divertor Tokamak Test facility (DTT) in the power exhaust implementation strategy is discussed. The evolution of the project, since the original proposal in 2015 to the present design, is shown. The DTT facility is well integrated in the European strategy and the final decision on the divertor configuration will be made, within 2022-23, on the basis of the indication of the Power Exhaust Group constituted by the EUROfusion Consortium. Finally, the main milestones and the timeline of the project are illustrated.

The Italian Divertor Tokamak Test (DTT) facility, a new machine for fusion research, has been proposed by the Italian Fusion Community to study the problem of power exhaust with different divertor configurations in conditions close to the future DEMO reactor. In DTT three additional heating systems are foreseen: Negative Neutral Beam Injector (NNBI), Electron Cyclotron resonance Heating (ECRH) and Ion Cyclotron Resonance Heating (ICRH). This paper reports a preliminary design of the most promising antenna concept for the DTT ICRH system, and the comparison between two load-resilient external matching schemes. The ICRH system has a modular structure and, in the first phase of DTT, the goal of coupling a RF power of 3 MW to the plasma in the frequency range 60-90 MHz can be reached with a single module, which mainly consists of two antennas, four external matching units (EMUs) and four generators. As far as the antenna is concerned, an array of 8 straps was found to give better performances, compared to other types of ICRH antennas, in terms of power coupling capability, VSWR and minimum conductance ( G min ). The antenna design has been firstly optimized with CST Microwave Studio at the central frequency of 75 MHz, using a movable water tank to emulate the magnetized plasma loading in front of the array. Then the antenna performances with realistic plasma load conditions have been carefully evaluated with suitable kinetic profiles. The most attractive matching scheme has been assessed by means of HFSS and CST, where a circuit simulation of the overall (end-to-end) RF system has been implemented. ICRH matching networks, realized with high power RF components in rigid coaxial cable (line stretchers, stubs, hybrid couplers, T-junction), are aimed at limiting the VSWR at the generators under the safely threshold of 1.5 : 1 by re-circulating the high reflected power toward the plasma without exceeding the maximum voltage of about 35 kV in the transmission line. Two different EMUs based on the conjugate-T and on the hybrid coupler have been studied by comparing the resilience to the strong loading variation due to the plasma instabilities in terms of S parameters, input VSWR, and electric fields. Finally adverse effects of asymmetries, as well as poloidal and toroidal mutual coupling between straps have been investigated too.

An Italian Divertor Tokamak Test (DTT) facility has been proposed to tackle a major mission of the European roadmap to fusion electricity, i.e. the problem of power exhaust. DTT will rely on an unprecedented amount of additional heating power, including ion-cyclotron resonance frequency (ICRF), to attain a DEMO-relevant power-over-radius ratio crossing the separatrix. This paper reports the status of the ICRF conceptual design, providing the rationale behind main design choices to couple 3 MW to the first DTT plasmafuss (day-1) and up to 10 MW after machine upgrade to full power. The most suitable antenna concepts are identified according to performance documented in literature along with matching schemes able to face fast variation of plasma admittance, and the state-of-the-art of DTT-relevant radiofrequency sources is briefly sketched too. Some preliminary antenna designs are compared on the basis of simulation results, discussing the strengths and weaknesses of each candidate.