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The launcher mirrors will allow the injection of microwave power from the gyrotrons in the plasma of the Divertor Tokamak Test facility, under construction in Italy. The active cooling of the elliptical M1 (curved and not-steerable) and M2 (flat and steerable) mirrors, both subject to a significant heat load, is investigated in this article. Several cooling solutions are analyzed: a spiral cooling for M1, both in a massive and in a lightened configuration, and also a radial-azimuthal cooling for M2, both in a massive and in a lightened configuration. Figures of merit are computed for the different configurations.

Most of the present and future magnetic fusion devices rely on high power mm-waves injection for a variety of applications like plasma breakdown, heating and control. Since under specific operating conditions the mm-waves might not be fully absorbed, potentially damaging in-vessel components, the detection of the non-absorbed radiation is recommended. Among the options for stray radiation detection, we consider here the possibility to exploit the solid body bolometer and its use for JT-60SA. Analysis of the residual power fraction expected for low absorption plasma conditions is presented, considering the dependence on the Electron Cyclotron Resonance Heating (ECRH) system and plasma parameters of JT-60SA. The residual power hitting the wall is evaluated looking at direct illumination and cross-polarization effects. The possibility to integrate a sensor for EC stray radiation, adapting the solution of the detector developed for ITER, is the main outcome of this study. In the ITER bolometer the differential thermocouple principle is considered to single out the nuclear heating and the microwave loads. A key component of the detector is the metal body with a ceramic coating layer, optimized for 170 GHz and 60 GHz. Mm-waves characterization of different coating samples has been performed to identify the optimal solution for JT-60SA considering its multi-frequency ECRH system (138 GHz, 110 GHz, 82 GHz), and will be here presented. The experience from JT-60SA equipped with these sensors will be also useful for their future use in ITER.

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.

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.

The DTT tokamak, whose construction is starting in Frascati (Italy), will be equipped with an ECRH system of 16 MW for the first operation phase and with a total of 32 gyrotrons (170 GHz, ≥ 1 MW, 100 s), organized in 4 clusters of 8 units each in the final design stage. To transmit this large number of power beams from the gyrotron hall to the torus hall building a Quasi-Optical (QO) approach has been chosen by a multi-beam transmission line (MBTL) similar to the one installed at W7-X Stellarator. This compact solution, mainly composed of mirrors in “square arrangement” shared by 8 different beams, minimizes the mode conversion losses. The single-beam QOTL is used to connect each gyrotron MOU output to a beam-combiner mirror unit and, after the MBTL, from a beam-splitter mirror unit to the exvessel and launchers sections located in the equatorial and upper ports of 4 DTT sectors. A novelty introduced is that the mirrors of the TLs are embodied in a vacuum enclosure, using metal gaskets, to avoid atmospheric absorption losses and microwave leaks. The TL, designed for up to 1.5 MW per single power beam, will have a total optical path length between 84 m and 138 m from the gyrotrons to the launchers. The main straight section will travel along an elevated corridor ~10 m above the ground level. The development of the optical design reflects the constraints due to existing buildings and expected neutron flux during plasma operation. In addition, the power throughput of at least 90% should be achieved.

The EC-system of the TCV tokamak is progressively being upgraded with the addition of two MW-class dual-frequency gyrotrons (84 and 126 GHz/2s/1MW). In order to connect the two gyrotrons to the existing low field side and top launchers, new waveguide routing from gyrotron hall to TCV tokamak was designed and dedicated Matching Optics Units (MOU) have been developed. The internal optics of the system have been determined aiming at optimal coupling to the HE11 waveguides. The laws of quasi-optics were used to find quadratic surfaces to shape an incoming Gaussian beam representative of the gyrotron output into a beam matching the proper field distribution at the waveguide entrance and with HE11 content compatible with the system requirements. A solution with one flat (movable) mirror and two shaping mirrors was found and characterized with the physical optics code GRASP. The resulting field distribution is then truncated and projected onto the HE11 component to evaluate the design solution (coupled power at the waveguide entrance >98.4% and HE11 content >96.8% for both frequencies). The model and the results of this analysis will be presented and compared to a model based on the Rayleigh–Sommerfeld scalar diffraction integral. GRASP was also used to evaluate preliminary misalignment effects in terms of coupled power to the waveguide and the MOU design moved to the manufacturing phase. In parallel to the gyrotron integration and to extend the level of flexibility of the TCV EC-system, a modular closed divertor chamber is developed, requiring the X3 top-launcher to be redesigned. Preliminary antenna conceptual design studies including new curvature to cope with the requirements of modularity and flexibility will be presented.

openaire: EC/H2020/633053/EU//EUROfusion An Electron Cyclotron Wave (ECW) system will be used in the European DEMO for the stabilization of Neoclassical Tearing Modes (NTMs). In order to avoid movable mirrors in the harsh environment close to the plasma and to simplify the NTM launcher integration, the tuning of the ECW deposition location can be achieved by launching frequency-tunable ECWs from fixed mirrors while the frequency is tuned in discrete steps of 2–3GHz. An overview of the frequency step-tunable ECW system for NTM stabilization is presented. The design considerations are discussed based on the current DEMO baseline parameters and the status of technologies. A simulation of NTM stabilization with an idealized frequency tunable ECW system on an analytical NTM model is shown. The simulation takes into account a realistic tuning speed based on the present technology and considers the current NTM launcher configurations in DEMO. A simple sweeping strategy is adapted for the control of frequency. Various uncertainties, which will affect the feasibility, need to be further investigated.

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 facility (DTT) will be a superconductive machine that aims at studying the power exhaust issue in a DEMO relevant environment. A significant power across the separatrix (compared to the radius of the machine), PSep/R ≅15 MW/m is of primary importance to fulfill this requirement and it can be obtained by injecting to the plasma up to 45 MW of additional power. This amount of power will be reached with a mix of Heating and Current Drive (HCD) systems: Electron Cyclotron Resonance Heating (ECRH), Ion Cyclotron Resonance Heating (ICRH) and Negative Neutral Beam Injection (NNBI). Power supply and wave generators of HCD will be hosted in remote-control buildings around the DTT Tokamak hall, and thus several ducts have been designed to connect the systems to the machine. The evacuated transmission lines will pass through dedicated holes in the main building making almost empty and big penetrations which cause a large neutron streaming through the walls. This work is devoted to a three-dimensional neutronics study, in support of the DTT licensing procedure, addressing the neutron and gamma streaming through the ECRH penetrations in order to optimize and determine the effectiveness of several proposed shielding options required to protect workers, population and auxiliary systems. Neutron and gamma transport simulations were carried out with the MCNP5 Monte Carlo code. Surface sources of neutrons and gammas have been used to propagate the radiation outside the penetrations. The ADVANTG hybrid transport code has been also used for variance reduction techniques in order to obtain statistically significant results in regions located several meters away from the plasma source. The results of the calculations are presented and discussed as well as the effectiveness of the proposed options to mitigate the radiation streaming and to reduce the doses outside the tokamak hall.

Heating and current drive (H&CD) systems are being investigated for a demonstration fusion power plant DEMO to deliver net electricity for the grid around 2050. Compared to ITER, which has to show the generation of 500-MW thermal power, the target of DEMO is the successful production of 300 to 500 MW electrical power to the grid and to aim for a self-sufficient tritium fuel cycle. Three H&CD systems are under development for DEMO in Europe, the electron cyclotron (EC) system, the neutral beam injection (NBI) system, and the ion cyclotron system. Based on present studies for plasma ramp-up, ramp-down, and flat top phases, to be further validated in more detailed simulations, the assumed total launched power needed from the H&CD system in DEMO is in the range of 50-100 MW, to be provided for plasma heating and control. This paper describes the design and Research and Development status of selected H&CD systems, considered for their deployment in the EU DEMO. It was always considered that different H&CD configurations and design variants will have an impact on the performances for the whole fusion plant. It shall be noted that the basis for the H&CD integrated design and system development is the actual version of the European fusion electricity roadmap. The project also elaborates on H&CD efficiency improvements which will reduce the recirculating power fraction in the future fusion power plants. Different studies under investigation will be discussed such as for NBI the photoneutralization and for EC novel concepts for gyrotron multistage-depressed collector.

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.

The feasibility of getting “D” shaped plasma configurations in FTU, with a possible X point close to the first wall, has been investigated with the aim of achieving an H-mode regime in this machine. This regime could allow both evaluating the thermal effects on the liquid lithium limiter due to the possible Edge Localized Modes and studying the L-H transition properties in low recycling conditions due to the presence of lithium.. An alternative design for the magnetic system in FTU has been also proposed, to realize an X-point inside the plasma chamber, close to the Liquid Lithium Limiter.Preliminary experiments with elongated configurations and limited ECRH additional heating power did not allowed approaching the L-H transition but they were used to develop a proper elongation control. This controller allowed guaranteeing the vertical stability in elongated configurations despite the reduced power available for the horizontal field coils in FTU. The elongation was stably keep over 1.2, while the lithium limiter was very close to the last close flux surface. Neither limiter damages nor plasma pollution were observed in these configurations.A possible alternative connection of the poloidal field coils in FTU is here proposed, with the aim of achieving a true X-point configuration with a magnetic single null well inside the plasma chamber and strike points on the lithium limiter. A preliminary assessment of this design allowed estimating the required power supply upgrade and showed its compatibility with the existing mechanical structure and cooling system, at least for plasmas with current up to 300 kA and flat-top duration up to 4s. Keywords: FTU, Liquid lithium limiter, L-H transition, X-point, Plasma elongation

The first unit of the dual-frequency gyrotron, 84126GHz/1MW/2s, for the upgrade of the TCV ECH system has been delivered and is presently being commissioned. During a first phase, long-pulse operation (T RF $\gt 0.5\mathrm{s})$ has been achieved and powers in excess of 0.93MW/1.1s and 1MW/1.2s have been measured in the evacuated RF-load at the two frequencies, 84GHz (TE$_{17,5}$ mode) and 126GHz (TE$_{26,7}$ mode), respectively. Considering the different rf losses in the experimental setup, the power level generated in the gyrotron cavity is in excess of 1.1MW and 1.2MW, with a corresponding electronic efficiency of 35% and 36%. These values are in excellent agreement with the design parameters and would likely lead to a gyrotron total efficiency higher than 50% in case of implementation of a depressed collector. The gyrotron behavior is remarkably reliable and robust with the pulse length extension to 2s presently only limited by external auxiliary systems.

Gyrotron R&D within the EUROfusion Work Package Heating and Current Drive (WPHCD) is addressing the challenging requirements posed on gyrotrons by the European concept for a demonstration fusion power plant (EU DEMO). These requirements, as specified within WPHCD, ask for highly reliable and robust long-pulse operation of the gyrotron, delivering 2 MW of microwave power at frequencies above 200 GHz with a high overall efficiency above 60% and the option for fast frequency step-tunability. To meet these targets, which are clearly beyond today's state-of-the-art, the R&D activities within WPHCD are organized in five main branches: these are the experimental verification of the advanced coaxial gyrotron technology at long pulses, the development of a coaxial gyrotron meeting the EU DEMO requirements, the development of multi-stage depressed collectors for enhanced energy recovery, the development of large broadband diamond windows to allow fast frequency tunability of the gyrotron, and the studies on further innovations and improvement of critical gyrotron components, in view of optimization of performance, reliability, and industrialization. The paper presents the progress made on these activities, the recent results, and the near-term planning. The major recent achievements include (i) the experimental validation of the fabrication of the new, water-cooled components for the longer-pulse coaxial gyrotron at Karlsruhe Institute of Technology, by demonstrating 2.2 MW at 170 GHz with 33% efficiency without depressed collector, (ii) the design of key components for a 2 MW, 170/204 GHz dual-frequency coaxial gyrotron, (iii) the design of a two-stage depressed collector with 77% efficiency, (iv) the first-ever production, in an industrial plasma reactor, of large chemical vapor deposition diamond wafers of 180mm diameter, (v) the procurement of an advanced electron gun with coated emitter edges, and (vi) advances in the theoretical and numerical modelling for investigating improved concepts for the gyrotron beam tunnel and cavity.

In Europe, the research and development with main focus on achieving robust industrial designs of series gyrotrons for electron cyclotron heating and current drive of today's nuclear fusion experiments and towards a future DEMOnstration fusion power plant is constantly progressing. The R&D is following two different paths. Both are complementing each other: Firstly, it is the adaption of the physical design and basic mechanical construction of the reliably operating 140 GHz, 1 MW CW (spec.: 920 kW, 1800 s) gyrotron of the stellarator Wendelstein 7-X (W7-X), Greifswald, Germany. With regards to time and costs it is the target to perform reliable developments of fusion gyrotrons with advanced specifications for today's plasma fusion experiments. Main focus is on the development of the first EU 170 GHz, 1 MW CW (3600 s) gyrotron for the installation in ITER, Cadarache, France. Another adaption is the dual-frequency 126/84 GHz 1 MW (2 s) gyrotron upgrade for the medium size TCV tokamak, Lausanne, Switzerland. Finally, it is the upgrade of the W7-X gyrotron design towards an RF output power per unit of up to 1.5 MW and possible dual-frequency operation by keeping the basic mechanical construction. Additional to the proven design it allows to fit the new 1.5 MW gyrotron into the already existing infrastructure and to reuse existing W7-X gyrotron auxiliaries, e. g. the high-power voltage supply (HV PS) and the superconducting (SC) magnet. The second R&D path is defined by the complementary approach with regards to development risks towards a future gyrotron which shall fulfil the significant more advanced specifications of a future EU DEMO. The starting point is the 2 MW EU/KIT coaxial-cavity gyrotron design. Main requirements are an RF output power of 2 MW CW at above 200 GHz, multiple operating frequencies, frequency step-tunability and a total efficiency above 60 %.

The evidence of Parametric Decay Instabilities (PDI) excited by the ECH power injected in O-Mode has been explored in FTU Tokamak, using the Collective Thomson Scattering (CTS) diagnostic. The experiments show evidences to support the hypothesis of low-threshold excitation of waves generated by PDI mechanisms, formerly proposed in the case of 2nd harmonic X-mode injection in TEXTOR and ASDEX-U. Theoretical analysis predicts low-threshold parametric decay also for O-mode pump-wave injection, which can be injected in FTU at frequencies close to the first Harmonic EC resonance. Experiments were made at different magnetic fields, injecting the 140 GHz probe and observing the emission from the second antenna of the EC launcher in poloidally symmetric and asymmetric configurations, in presence of MHD islands. The signal is detected by the CTS radiometers, with a fast digitizer allowing the spectral reconstruction at very fine time and frequency scales. Different types of emissions are studied in detail, comparing them with the magnetic island rotation frequency in different plasma conditions. In order to locate the plasma volume originating the emissions, a new antenna and receiving line has been installed.

The integration of the heating and current drive (HCD) systems in the EU DEMO tokamak must address a number of issues, namely space constraints in the tokamak building, remote handling requirements, breeding blanket penetration, neutron and photon radiation shielding, compliance of penetrations of the primary vacuum with safety and vacuum criteria, and a large number of loading conditions, in particular heat, electromagnetic (EM), and pressure loads in normal and off-normal conditions. A number of pre-conceptual design options for the vacuum vessel (VV) port and the port-plug are under assessment and need to be verified against all requirements and related criteria. The identification of the functional (or physics) requirements of the HCD systems remains an on-going process during the pre-conceptual design phase, hence some initial assumptions had to be made as a basis for development of the design of the VV ports and the HCD port plugs. The paper will provide an overview of present margins in the functional/physics requirements and the rationale behind the assumptions made in order to facilitate development of the pre-conceptual design options. Furthermore it will introduce the initial design concepts of the electron cyclotron (EC) Launchers and the neutral beam (NB) injectors integrated in equatorial ports. The NB duct design in DEMO and related issues such as transmission and re-ionization losses will be also addressed.

A new receiving antenna for collecting signals of the Collective Thomson Scattering (CTS) diagnostics in FTU Tokamak has been recently installed. The squared corrugated section and the precisely defined length make it possible to receive from different directions by remotely steering the receiving mirrors. This type of Remote-Steering (RS) antennas, being studied on FTU for the DEMO Electron Cyclotron Heating (ECH) system launch, is already installed on the W7- X stellarator and will be tested in the next campaign. The transmission of the signal from the antenna in the tokamak hall to the CTS diagnostics hall will be mainly realized by means of oversized circular corrugated waveguides carrying the hybrid HE11 (quasi-gaussian) waveguide mode, with inclusion of a special smooth-waveguide section and a short run of reduced-size square-corrugated waveguide through the tokamak bio-shield. The coupling between different waveguide types is made with ellipsoidal focusing mirrors, using quasi-optical matching formulas between the gaussian-shaped beams in input and output to the waveguides. In this work, after a complete study of feasibility of the overall line, a design for the receiving line will be proposed, in order to realize an executive layout to be used as a guideline for the commissioning phase. © 2018 ENEA. Published by IOP Publishing Ltd on behalf of Sissa Medialab.

The engineering design and R&D of auxiliary heating systems and their sub-systems are a key activities in the frame of the present conceptual design phase for a first of a kind DEMOnstration Fusion Power Plant in order to develop a system capable of achieving and controlling burning plasmas. In the frame of the EU DEMO reference design, the R&D activities consider the injection of about 50 MW of Electron Cyclotron (EC) power to support and assist different plasma phases. As the project is still in the conceptual phase, a range of options for gyrotrons, transmission lines and antennas is under assessment taking into account the guidelines for the integration of the EC system in a nuclear reactor and a maximal achievable reliability and availability of the EC power during operation.

The upgrade of the EC-system of the TCV tokamak has entered in its realization phase and is part of a broader upgrade of TCV. The MW-class dual-frequency gyrotrons (84 or 126GHz/2s/1MW) are presently being manufactured by Thales Electron Devices with the first gyrotron foreseen to be delivered at SPC by the end of 2017. In parallel to the gyrotron development, for extending the level of operational flexibility of the TCV EC-system the integration of the dual-frequency gyrotrons adds a significant complexity in the evacuated 63.5mm-diameter HE11 transmission line system connected to the various TCV low-field side and top launchers. As discussed in [1], an important part of the present TCV-upgrade consists in inserting a modular closed divertor chamber. This will have an impact on the X3 top-launcher which will have to be reduced in size. For using the new compact launcher we are considering employing a Fast Directional Switch (FADIS), combining the two 1MW/126GHz/2s rf-beams into a single 2MW rf-beam.

A demonstration fusion power plant (DEMO) producing electricity for the grid at the level of a few hundred megawatts is included in the European Roadmap [1]. The engineering design and R&D for the electron cyclotron (EC), ion cyclotron and neutral beam systems for the DEMO reactor is being performed by Work Package Heating and Current Drive (WPHCD) in the framework of EUROfusion Consortium activities. The EC target power to the plasma is about 50 MW, in which the required power for NTM control and burn control is included. EC launcher conceptual design studies are here presented, showing how the main design drivers of the system have been taken into account (physics requirements, reactor relevant operations, issues related to its integration as in-vessel components). Different options for the antenna are studied in a parameters space including a selection of frequencies, injection angles and launch points to get the best performances for the antenna configuration, using beam tracing calculations to evaluate plasma accessibility and deposited power. This conceptual design studies comes up with the identification of possible limits, constraints and critical issues, essential in the selection process of launcher setup solution.

The proposed Divertor Test Tokamak, DTT, aims at studying power exhaust and divertor load in an integrated plasma scenario. Additional heating systems have the task to provide heating to reach a reactor relevant power flow in the SOL and guarantee the necessary P SEP /R together with adequate plasma performances. About 40 MW of heating power are foreseen to have P SEP /R ≥ 15 MW/m. A mix of the three heating systems has been chosen, assuring the necessary flexibility in scenario development. An ECRH system at 170 GHz will provide 10 MW at plasma for several tasks, such as: bulk electron heating, localized CD, avoidance of impurity accumulation and MHD control. In addition 15 MW of ICRH in the range 60–90 MHz will provide the remaining bulk plasma heating power, on both electrons and ions. ICRH, in minority scheme, will produce fast ions, allowing the study of fast particle driven instabilities like alphas in D-T burning plasmas. The heating schemes foreseen in DTT are 3He and H minority as well as Deuterium 2nd harmonic. The addition of 15 MW of NBI, later in the project, could provide a mainly parallel fast ion distribution to simulate the alpha heating scheme of a reactor. The NBI primary aim is to support plasma heating during the flat top phase when the need of central power deposition and the minimization of the shine-through risk suggests a beam energy around 300 keV. In the first phase of the DTT project the available power will be at least 25 MW, to be increased during the lifetime of the machine.

During the last years the Collective Thomson Scattering (CTS) system installed on the Frascati Tokamak Upgrade (FTU), operating at 140 GHz, has been exploited to investigate different kinds of phenomena: high-frequency daughter waves by Parametric Decay Instabilities, which, under certain conditions, are recently presumed to be conceivable even under routine ECRH scenarios, and bulk ions thermal emissions stimulated by a 140 GHz gyrotron beam. A new receiving line, giving rise to a second further line of sight from the plasma during the scattering experiments and integrated in the CTS antenna, has been recently completed and installed. Such new line will be useful to recognise signals originating in the volume of beams cross from the ones coupled outside or also to monitor the gyrotron's spectrum in real-time during the shot. The present status of this new installation is shown in the paper.

During the ongoing pre-concept phase (2014 – 2020) for a possible future European DEMOnstration Fusion Power Plant (DEMO) the activities within EUROfusion WP HCD EC Gyrotron R&D and Advanced Developments are focusing on options for near-term solutions, and, at the same time, on long-term even more advanced options. The near-term target for DEMO is to realize pulsed operation. According to the current baseline it will probably use an EC system operating at 170 GHz and 204 GHz is being assessed, whereas the long-term target aims for steady-state operation and frequencies for current drive up to 240 GHz. Common targets for both are an RF output power per unit of significantly above 1 MW (target: 2 MW) and a total gyrotron efficiency of significantly higher than 60 %. Frequency step-tunability and multi-purpose/multi-frequency operation have to be considered. Those targets shall be achieved by considering the coaxial-cavity gyrotron technology and advanced technologies for key components (e.g. CVD diamond-disk Brewster angle window). Advanced simulation and test tools are complementing the research and developments. Gyrotron development is additionally supported by a significant investment into a new multi-megawatt long-pulse gyrotron test stand which is under final installation at KIT currently.

The upgrade of the electron cyclotron (EC) system of the TCV tokamak has entered in its realization phase. The MW-class dual-frequency gyrotron design (84 or 126GHz/2s), based on the 140GHz W7-X gyrotron, is finalized and a contract for the manufacturing of two units has been placed with Thales Electron Devices (TED). For maintaining a high level of operational flexibility of the EC-system the integration of the dual-frequency gyrotrons adds a significant complexity in the evacuated HE11 transmission line system connected to the various TCV low-field side and top launchers. Specific aspects of the gyrotrons design and their integration into the EC-system are described. The EC-system upgrade is an essential step of a larger extent upgrade of TCV.

During the last years anomalous signals have been found in the spectra analyzed with the Collective Thomson Scattering (CTS) diagnostics in tokamak devices like TEXTOR, ASDEX and FTU. The signal frequency, down-shifted with respect to the probing one, suggested the explanation of their origin in terms of Parametric Decay Instability (PDI) processes correlated with the presence of magnetic islands and taking place even for values of the pumping wave power below the threshold predicted by conventionalmodels. A threshold lower or comparable with the power levels routinely used could pose limitations to the use of the ECRH in fusion devices. An accurate characterization of the conditions for the occurrence of this phenomenon and of its consequences is thus of primary importance. The CTS diagnostics of FTU has been recently improved for this aim with a new fast data acquisition, allowing spectra reconstruction by direct FFT of the intermediate frequency signal, added in parallel to the pre-existing multichannel spectrum analyzer. A section with polarizing mirrors has been introduced in the quasi-optical transmission line to control the polarization of the received signals. Exploiting the front-steering configuration available with the real-Time launcher of FTU, the setup allows studying the anomalous emissions in conditions of density and wave injection similar to ITER. Given the possibility of the CTS in FTU to operate in both O-and X-mode, a second radiometer is being installed, to detect simultaneously signals scattered in both polarizations. The upgrades mentioned above are presented in this work as well as the first data obtained in 2014, during the experimental activity carried out under an EUROfusion Enabling Research project.

The ITER scenarios and the project of DEMO involve stable operation above the Greenwald density, which justifies efforts to understand and overcome the density limit, this last observed as a disruptive termination of tokamak discharges and a thermal crash (with no disruption) of stellarator and reversed-field pinch (RFP) ones. Both in the tokamak and the RFP, new findings show that the high density limit is not governed by a unique, theoretically well-determined physical phenomenon, but by a combination of complex mechanisms involving two-fluid effects, electrostatic plasma response to magnetic islands and plasma-wall interaction. In this paper we will show new evidence challenging the traditional picture of the 'Greenwald limit', in particular with reference to the role of thermal instabilities and the edge radial electric field Er in the development of this limit. © 2015 EURATOM.

Anomalous emissions were found over the last few years in spectra of Collective Thomson Scattering (CTS) diagnostics in tokamak devices such as TEXTOR, ASDEX and FTU, in addition to real CTS signals. The signal frequency, down-shifted with respect to the probing one, suggested a possible origin in Parametric Decay Instability (PDI) processes correlated with the presence of magnetic islands and occurring for pumping wave power levels well below the threshold predicted by conventional models. A threshold below or close to the Electron Cyclotron Resonance Heating (ECRH) power levels could limit, under certain circumstances, the use of the ECRH in fusion devices. An accurate characterization of the conditions for the occurrence of this phenomenon and of its consequences is thus of primary importance. Exploiting the front-steering configuration available with the real-time launcher, the implementation of a new CTS setup now allows studying these anomalous emission phenomena in FTU under conditions of density and wave injection geometry that are more similar to those envisaged for CTS in ITER. The upgrades of the diagnostic are presented as well as a few preliminary spectra detected with the new system during the very first operations in 2014. The present work has been carried out under an EUROfusion Enabling Research project. A shorter version of this contribution is due to be published in PoS at: 1st EPS conference on Plasma Diagnostics © 2015 Istituto di Fisica del Plasma - CNR.

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Since 2014 under the umbrella of EUROfusion Consortium the Work Package Heating and Current Drive (WPHCD) is performing the engineering design and R&D for the electron cyclotron (EC), ion cyclotron and neutral beam systems of the future fusion power plant DEMO. This presentation covers the activities performed in the last two years on the EC system conceptual design, as part of the WPHCD, focusing on launchers, transmission lines, system reliability and architecture.

A Cyclotron Auto-Resonance Maser (CARM) is under study at the ENEA Research Laboratories in Frascati as a electron cyclotron heating (ECH) source in plasma magnetic confinement devices like Tokamaks. The aim of the project is the realization of a test source at 250 GHz to verify heating mechanisms in high magnetic field Tokamaks and at the second harmonics ECH frequencies for which a higher energy exchange efficiency has been predicted.

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The oblique electron cyclotron emission (ECE) diagnostic installed at JET allows simultaneous analysis of the ECE spectra along three lines of sight (with toroidal angles of 0°, ~ 10°, and ~ 20°) and two linear polarizations for each oblique line of sight. The diagnostic is capable of measuring EC emission over the band of 75-800 GHz with 5 ms time resolution and 7.5 GHz spectral resolution, and it is designed to investigate the features of ECE spectra related to electron distribution in the thermal velocity range. Instrumental accuracy was assessed using sources at different temperatures (77-900 K) and with plasma emission. ECE from high temperature plasmas and in the presence of fast ions has been compared to simulations performed with the modeling code SPECE, setting an upper limit to possible discrepancies from thermal emission.

The new Oblique ECE diagnostics of JET allows simultaneous measurement along three lines of sight at different toroidal angles. JET pulses, in which modulated LHCD power was applied, were analyzed by means of the SPECE emission code, which takes into account the fast electron tail generated by LH. The code results were compared to the ECE spectra from the oblique ECE diagnostic. The match between the computed and the experimental data from the three lines of sight provides an estimate of the density of superthermal electrons and of their maximum energy from which the fraction of the plasma current driven by LHCD is derived.

Since the 2012 IAEA-FEC Conference, FTU operations have been largely devoted to runaway electrons generation and control, to the exploitation of the 140GHz electron cyclotron (EC) system and to liquid metal limiter elements. Experiments on runaway electrons have shown that the measured threshold electric field for their generation is larger than predicted by collisional theory and can be justified considering synchrotron radiation losses. A new runaway electrons control algorithm was developed and tested in presence of a runaway current plateau, allowing to minimize the interactions with plasma facing components and safely shut down the discharges. The experimental sessions with 140GHz EC system have been mainly devoted to experiments on real-time control of magnetohydrodynamic (MHD) instabilities using the new EC launcher with fast steering capability. Experiments with central EC injection have shown the onset of 3/2 and 2/1 tearing modes, while EC assisted breakdown experiments have been focused on ITER start-up issues, exploring the polarization conversion at reflection from inner wall and the capability to assure plasma start-up even in presence of a large stray magnetic field. A new actively cooled lithium limiter has been installed and tested. The lithium limiter was inserted in the scrape-off layer, without any damage to the limiter surface. First elongated FTU plasmas with EC additional heating were obtained with the new cooled limiter. Density peaking and controlled MHD activity driven by neon injection were investigated at different plasma parameters. A full real-time algorithm for disruption prediction, based on MHD activity signals from Mirnov coils, was developed exploiting a large database of disruptions. Reciprocating Langmuir probes were used to measure the heat flux e-folding length in the scrape-off layer, with the plasma kept to lay on thea internal limiter to resemble the ITER start-up phase. New diagnostics were successfully installed and tested, as a diamond probe to detect Cherenkov radiation produced by fast electrons and a gamma camera for runaway electrons studies. Laser induced breakdown spectroscopy measurements were performed under vacuum and with toroidal magnetic field, so demonstrating their capability to provide useful information on the surface elemental composition and fuel retention in present and future tokamaks, such as ITER. © 2015 EURATOM.

Deviations of the electron function velocity distribution from Maxwellian behavior in the high energy range have been extensively studied during the past two decades. A brief review of the experimental techniques and findings on the subject is given in this paper. There are indications that the electron distribution function in thermonuclear plasmas could be significantly different from the Maxwellian one even near the thermal velocity range when particular circumstances occur. These reasons motivate a renewed effort in the measurements that could detect low energies distortions, among which is the new Oblique Electron Cyclotron Emission diagnostics that has entered operations during 2006 experimental campaign at JET and aims to resolve relatively small differences in the ECE spectra taken at three toroidal angles. Such measurements allow insight into the characteristics of the electron distribution function. First results of such experimental system are discussed in this paper, along with emission simulations.

The LFI instrument, onboard the PLANCK satellite, consists of 22 pseudo-correlation radiometers. They are composed of a front end unit working at 20 K and a back end unit working at 300 K. The passive components of these radiometers are the feed horns, collecting the radiation from the sky, the ortho mode transducers that split the incoming radiation into two perpendicular linearly polarized signals, and the waveguides that connect the front end to the back end unit. In this work we describe their characterization.

Experiments on real time control of magneto-hydrodynamic (MHD) instabilities using injection of electron cyclotron waves (ECW) are being performed with a control system based on only three real time key items: an equilibrium estimator based on a statistical regression, a MHD instability marker (SVDH) using a three-dimensional array of pick-up coils and a fast ECW launcher able to poloidally steer the EC absorption volume with dρ/dt = 0.1/30 ms maximum radial speed. The MHD instability, usually a tearing mode with poloidal mode number m and toroidal mode number n such that m/n = 2/1 or 3/2 is deliberately induced either by neon gas injection or by a density ramp hitting the density limit. No diagnostics providing the radial localization of the instabilities have been used. The sensitivity of the used MHD marker allows to close the control loop solely on the effect of the actuator's action with little elaboration. The nature of the instability triggering mechanism in these plasma prevents that the stabilization lasts longer than the ECW pulse. However when the ECW power is switched on, the instability amplitude shows a marked sensitivity to the position of the absorption volume with an increase or decrease of its growth rate. Moreover the suppression of the dominant mode by ECRH performed at high plasma density even at relatively low power level facilitates the development of a secondary mode. This minimized set of control tools aim to explore some of the difficulties which can be expected in a fusion reactor where reduced diagnostic capabilities and reduced actuator flexibility can be expected. © 2015 EURATOM.

In this paper we present the test results of the qualification model (QM) of the LFI instrument, which is being developed as part of the ESA Planck satellite. In particular we discuss the calibration plan which has defined the main requirements of the radiometric tests and of the experimental setups. Then we describe how these requirements have been implemented in the custom-developed cryo-facilities and present the main results. We conclude with a discussion of the lessons learned for the testing of the LFI Flight Model (FM).

The millimetric region of the electromagnetic spectrum is of great interest both to plasma physics and to experimental cosmology. Building on the common technological interest, the tests on passive components of the Low Frequency Instrument of the PLANCK mission are made in IFP’s microwave laboratory. The experimental techniques and sample results are shown.

A Martin Puplett interferometer for electron cyclotron emission (ECE) measurements from JET tokamak plasmas was extended to multichannel operation for simultaneous radial and oblique ECE measurements. This paper describes the new optics and the instrument's performance. [doi:10.1063/1.3660276]

An important goal of the control system in a tokamak is the suppression of magneto-hydrodynamic (MHD) instabilities with low m, n (poloidal and toroidal mode numbers), which can influence the confinement time of energy and particles and possibly lead to plasma disruption. These instabilities, which appear as rotating magnetic islands, can be reduced or completely suppressed by a current driven by electron cyclotron waves (ECW) accurately located within the island. A fundamental requisite for this control technique is the ability to identify the island parameters (amplitude and radial position) and to vary accordingly the ECW deposition location. Here we describe a control scheme of the steering mirror of the ECW source based on the real-time tracking of the island radial position realized using only the electron cyclotron emission (ECE) diagnostics in quasi-in-line configuration, i.e. with toroidal anti-parallel propagation of the ECW and ECE beams. The successful experimental proof of principle of this scheme, tested on the TCV tokamak, is here reported.

JT-60SA envisions electron cyclotron wall conditioning (ECWC), as wall conditioning method in the presence of the toroidal field to control fuel and impurity recycling and to improve plasma performance and reproducibility. This paper reports on Helium ECWC experiments on TCV in support of JT-60SA operation. Nearly sixty Helium conditioning discharges have been successfully produced in TCV, at a toroidal field B T = 1.3 or 1.54 T, with gyrotrons at 82.7 GHz in X2 mode, mimicking ECWC operation in JT-60SA at the second harmonic of the EC wave. Discharge parameters were tuned in order to (i) minimize the time for the onset of ECWC plasmas, thus minimizing absorption of stray radiation by in-vessel components, (ii) improve discharge homogeneity by extending the discharge vertically and radially, and wall coverage, in particular of inboard surfaces where JT-60SA plasmas will be initiated, (iii) assess the efficiency of He-ECWC to deplete carbon walls from fuel. An optimized combination of vertical and radial magnetic fields, with amplitudes typically 0.1 to 0.6% of that of B T, has been determined, which resulted in lowest breakdown time, improved wall coverage and enhanced fuel removal. A standard ohmic D 2-plasma could be then sustained, whereas it would not have been possible without He-ECWC.

In the frame of the TCV Tokamak upgrade, two 84/126 GHz/2 s dual frequency gyrotrons designed by SPC and KIT and manufactured by THALES will be added to the existing EC-System. The first unit has been delivered to EPFL-SPC and tested. In the commissioning configuration, a matching optics unit (MOU) is connected to the gyrotron window. The RF is then coupled to the HE11 mode of a 63.5mm corrugated waveguide and dissipated in a load procured by CNR after 4m of waveguide and 2 miter bends. Owing to the flexible triode gun design giving the possibility to adjust the pitch angle parameter, the specifications were met at both frequencies. At 84 GHz (TE17,5 mode), a power of 0.930 MW was measured in the calorimeter, with a pulse duration of 1.1 s. At the high frequency (126 GHz, TE26,7 mode), a power of 1.04 MW was reached for a pulse length of 1.2 s. Accounting for the load reflection and the ohmic losses in the various subcomponents of the transmission line and the tube, it is estimated that the output power at the gyrotron window is in excess of 1 MW at both frequencies, with an electronic efficiency of 32% and 34% at 84 GHz and 126 GHz respectively. The gyrotron behavior is remarkably robust and reproducible, and the pulse length is limited by external systems that will be improved shortly.

The conceptual design study of a high-power high-frequency Cyclotron Auto-Resonance Maser (CARM) has been undertaken at the ENEA-Frascati research center in 2013. CARMs could be effective mm-wave sources for electron cyclotron waves and plasma diagnostics in magnetic confinement fusion reactors. The project is advancing quickly with the technical design of some major parts now completed and ready to enter the realization phase. Subsystems under conceptual or detailed design are presented together with the outlook of next steps. © 2015 IEEE.