A. Bruschi

Current drive at plasma densities required for thermonuclear reactors

Progress in thermonuclear fusion energy research based on deuterium plasmas magnetically confined in toroidal tokamak devices requires the development of efficient current drive methods. Previous experiments have shown that plasma current can be driven effectively by externally launched radio frequency power coupled to lower hybrid plasma waves. However, at the high plasma densities required for fusion power plants, the coupled radio frequency power does not penetrate into the plasma core, possibly because of strong wave interactions with the plasma edge. Here we show experiments performed on FTU (Frascati Tokamak Upgrade) based on theoretical predictions that nonlinear interactions diminish when the peripheral plasma electron temperature is high, allowing significant wave penetration at high density. The results show that the coupled radio frequency power can penetrate into high-density plasmas due to weaker plasma edge effects, thus extending the effective range of lower hybrid current drive towards the domain relevant for fusion reactors.

High Plasma Density Lower-Hybrid Current Drive in the FTU Tokamak

s Vol. 19C (European Physical Society, Geneva, 1995), Part III, p. 361. [15] G. Tonon et al., Plasma Phys. Controlled Fusion 35, A105

Design of the EU-1MW gyrotron for ITER

EU is developing a 1 MW cylindrical cavity gyrotron. In the last year the design of the components of the new gyrotron has been finalized while the technological design of the new tube has been defined. In the present paper, the main characteristics of the new EU gyrotron for ITER are presented.

A Real-Time Tracking for Optimal Wave Injection in Overdense Plasma Heating Experiments at 140 GHz in FTU

Experiments on overdense plasma heating through the mode-coupling scheme known as “O-X-B Double Mode Conversion” obtained launching a narrow beam of millimeter waves at 140-GHz frequency and 400-kW power are scheduled for the next experimental campaigns of the FTU tokamak. Such a scheme, not yet demonstrated at electron density higher than the critical one (2.4 ·1020m-3) for the 140-GHz ordinary mode, and consequently at such a high frequency, exploits the conversion of an ordinary polarized wave (O) into the extraordinary (X) one, followed by a subsequent conversion to Bernstein (B) waves, which are then absorbed by the plasma. In the specific case of FTU, the overall efficiency of this scheme is mainly determined by the coupling efficiency between the O- and the X-wave, which can occur only for ordinary polarized radiation propagating in a very narrow angular range at the cutoff region. The simulations performed with a single ray tracing show that the required precision in the injection of the wave into the plasma is very high and an angular deviations of ±1° with respect to the optimal injection, in either vertical or horizontal direction, implies a 50% drop in the power transmitted to X-mode. Moreover, the application of models able to take into account the real shape of the incident beam shows that the maximum reachable efficiency, under optimal wave injection, is expected not to exceed 45% of the EC power injected, while the diameter of the angular window corresponding to 50% of power conversion keeps ≤ 3° in both polidal and toroidal directions. The new E CH&CD launcher, now being installed in FTU, will be able to provide the angular precision required for the steering. The basic idea of a control algorithm, aimed to track in real time the optimal angular window for the wave injection in experiments on O-X-B mode conversion, is presented in the paper. The control will use the stray gyrotron radiation as observable, which is detected by a set of sniffer probes located at different toroidal positions in the FTU vessel.

DIAGNOSING THE ELECTRON DISTRIBUTION FUNCTION WITH OBLIQUE ELECTRON CYCLOTRON EMISSION ON FTU

This paper addresses the problem of diagnosing the electron distribution function by measuring Electron Cyclotron Emission (ECE) at different angles with respect to the magnetic field. This diagnostic set-up, that is often referred to as oblique ECE, provides a unique method for measuring the electron distribution function with good energy and space resolution at once. First oblique ECE measurements have been performed on FTU during onaxis Electron Cyclotron Resonance Heating (ECRH) applied on the current ramp-up phase of the discharge. The experimental results appear to be consistent with theoretical FokkerPlanck calculations, showing the existence of a deformation of the bulk of the electron distribution function.

FaDiS, a fast switch and combiner for high-power millimetre wave beams

The virtual German-Italian-Russian Helmholtz- Institute Advanced ECRH for ITER runs a program FADIS aimed to develop a system for combining outputs from a set of gyrotrons and toggling the combined beam with kHz-frequency between two launcher for dedicated electron cyclotron heating and current drive in fusion experiments. A diplexer prototype fed by a 140 GHz gyrotron with ~0.5- MW, ~ 1-s pulses operated in good agreement with theory.

Review of the Innovative H&CD Designs and the Impact of Their Configurations on the Performance of the EU DEMO Fusion Power Plant Reactor

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.

From W7-X towards ITER and beyond: Status and progress in EU fusion gyrotron developments

In Europe, significant progress in gyrotron research, development and manufacturing has been made in 2014, starting from the successful continuation of the 1 MW, 140 GHz gyrotron production for the stellarator Wendelstein 7-X (W7-X) at Greifswald, Germany and the accelerated development of the EU 1 MW, 170 GHz conventional cavity gyrotron for the ITER tokamak at Cadarache, France. Based on that, a physical design activity was started which shall lead to a dual frequency gyrotron for TCV, Lausanne, Switzerland. Within the European fusion development consortium (EUROfusion), advanced gyrotron research and development has started towards a future gyrotron design which shall fulfil the needs of DEMO, the nuclear fusion demonstration power plant that will follow ITER. Within that research and development, the development of advanced design tools, components, and proper test environment is progressing as well. A comprehensive view over the status and prospects of the different development lines shall be presented.

Experiments on magneto-hydrodynamics instabilities with ECH/ECCD in FTU using a minimal real-time control system

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 actuators 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.

Guidelines for internal optics optimization of the ITER EC H&CD upper launcher

The importance of localized injection of Electron Cyclotron waves to control Magneto-HydroDynamic instability is well assessed in tokamak physics and the set of four Electron Cyclotron (EC) Upper Launchers (UL) in ITER is mainly designed for this purpose. Each of the 4 ULs uses quasi-optical mirrors (shaping and planes, fixed and steerable) to redirect and focus 8 beams (in two rows, with power close to 1 MW per beam coming from the EC transmission lines) in the plasma region where the instability appears. Small beam dimensions and maximum beam superposition guarantee the necessary localization of the driven current. To achieve the goal of MHD stabilization with minimum EC power to preserve the energy confinement in the outer half of the plasma cross section, optimization of the quasi-optical design is required and a guideline of a strategy is presented. As a result of this process and following the guidelines indicated, modifications of the design (new mirrors positions, rotation axes and/or focal properties) will be proposed for the next step of an iterative process, including the mandatory compatibility check with the mechanical constraints.