Table of contents

This meeting belongs to a series of topical events which the International Atomic Energy Agency (IAEA, Vienna) organizes in a regular basis on crucial aspects of nuclear fusion research, or related in particular to ITER physics or a technological application relevant to the nuclear fusion reactor. Each Technical Meeting series has a specific object; the events are called on a two–three years basis and are recommended by the IAEA advisory body for Fusion, the International Fusion Research Council (IFRC) . The object of the IAEA-TM held in Como, Italy, 2–5 May 2005, was the application in ITER of powerful Electron Cyclotron waves in the millimeter wave frequency range for plasma Heating and noninductive Current Drive. The meeting was the third on this subject. There were 42 presentations to an audience of about 60 delegates from 16 countries.

The main goal of this series of IAEA-TM is to bring together specialists of the different branches involved in the project, in the effort of the best understanding of the limits and capabilities of each one of the different fields of research and development. Millimeter-wave source developers, millimeter-wave system designers and plasma physicists, theoreticians and experimentalists in all of the fields, exposed their way of addressing the problem in plenary sessions attended by all participants. Discussions on the different topics of gyrotron development, launcher options and physics application were continued in forums following the presentations.

The specialist reader will find in this volume in particular the latest developments concerning the frequency, the output power and the efficiency of the gyrotrons which are now being considered the preferred type of high power millimeter wave generators for ECH/ECCD applications in the fusion reactor. The debate on the launcher of the EC waves, in the form of Gaussian beams, is presently very active, with a few options on the table to be merged in one optimized and integrated design. Concerning the applications to the ITER plasma, most attention was dedicated to the use of ECCD to actively control and stabilize Magneto-Hydrodynamic instability Modes, with an increasing attention to the need and the crucial issues of the automation of such a control in the reactor.

The general reader should find an up-to-date presentation of the potential and of the limits of the use of high power millimeter waves in fusion plasmas, and have a vivid image of the interplay between sources, systems and plasma physics in such an application.

The conference was held under the scientific responsibility of IAEA, Vienna, and hosted and organized in Como, Italy, by the Istituto di Fisica del Plasma, Consiglio Nazionale delle Ricerche, Milan.

We report operation of a 110 GHz gyrotron with 1.67 MW of output power measured in short pulses (3µs) at an efficiency of 42% in the TE_22,6 mode. We also present a preliminary design of a 1 MW, 120 GHz gyrotron for ITER start-up with an efficiency greater than 50%.

Recent activities on the research and development of 170 GHz gyrotron were presented. In the 170 GHz gyrotron experiment in JAERI, pulse duration had been limited by the beam current decrease due to the emission cooling of the electron emitter. A preprogramming control of the heater power was introduced to keep the beam current constant and to avoid the oscillation mode shift to a lower mode during the operation. And, a built-in mode converter system was improved to reduce the stray radiation. In the preliminary experiment results of 0.2 MW/480 sec and 0.13 MW/600 sec were obtained. The pulse extension experiment at higher power will be followed. In parallel, a study of the high order mode oscillation was carried out using a short pulse gyrotron (∼1 msec). The oscillation mode is TE_31,12, which allows 1.5 MW level CW operation. A stable oscillation of TE_31,12 was demonstrated at the power of 1.56 MW. The maximum efficiency was 30% at 1 MW output. This result indicates that the high order mode up to the level of TE_31,12 will be acceptable to increase the power or to reduce the heat load on the cavity wall at 1 MW output.

Recent tests of two different megawatt-class gyrotrons at CPI provide a design baseline for the 120 GHz, 1 MW CW gyrotrons required by ITER for plasma start-up. The two gyrotron designs include a 140 GHz, 900 kW CW device that was delivered to the Max Planck Institute for Plasma Physics in Greifswald, Germany, and a 110 GHz, 1.3 MW CW gyrotron that was recently tested at CPI. Both gyrotrons utilize many of the same features that are foreseen for the ITER 120 GHz gyrotrons. These features include: single-anode magnetron injection electron guns, high-efficiency internal converters that produce a pure TEM₀₀ output mode, 88-mm-diameter diamond output windows and single-stage depressed collectors. In tests at CPI, the 140 GHz gyrotron achieved 500 kW power levels for 700-s pulses and 930 kW for short pulses. Long-pulse tests of the gyrotron, at beam currents above the 25-A test-set limit at CPI, were performed in Greifswald and resulted in output power levels of up to 900 kW for 30-minute pulses. In short-pulse operation at CPI, the 110 GHz gyrotron has achieved power levels of 1.28 MW. In long-pulse tests of the gyrotron, power levels of 500 kW were obtained for 10-s pulses.

A 2 MW, CW, 170 GHz coaxial cavity gyrotron is under development in cooperation between European Research Institutions (FZK Karlsruhe, CRPP Lausanne, HUT Helsinki) and the European tube industry (TED, Velizy, France). The design of critical components has recently been examined experimentally at FZK Karlsruhe with a short pulse (∼ few ms) coaxial cavity gyrotron. This gyrotron uses the same cavity and the same quasioptical (q.o.) RF-output system as designed for the industrial prototype and a very similar electron gun.

In ITER, EC heating and current drive (H&CD) is foreseen not only as a principal auxiliary system for plasma heating and as assist for plasma start-up, but is considered essential in meeting the key requirement of neoclassical tearing mode (NTM) stabilisation, by localized current drive. In the reference ECH design, ITER requires a total of 20 MW/CW power at 170 GHz using gyrotrons with a unit power of 1 MW. A higher power per unit (2 MW/gyrotron) would result in a strong reduction of the cost of the whole ECRH system, and would also relax the room constraints on the launcher antenna design. In view of the capability of coaxial cavity gyrotrons demonstrated with short pulse experiments at FZK, the European Fusion Development Agreement (EFDA) has started in 2003 the development of an industrial 170 GHz 2 MW/CW coaxial cavity gyrotron, in a collaborative effort between European research associations CRPP/EPFL, FZK, TEKES and Thals Electron Devices (TED). The development plan includes three steps to reach successively 2 MW/1s, 2 MW/60s and finally 2 MW/CW operation. The procurement of the first prototype is in progress and it scheduled to be delivered during the first quarter of 2006. The experimental tests of the prototypes will be carried out at CRPP/EPFL, where an ITER relevant test facility is presently under construction and will be achieved during the second half of 2005. The test facility is designed to be flexible enough, allowing the possible commissioning of tubes with different characteristics, as well the tests of the launcher antenna at full performances.

The JT-60U electron cyclotron range of frequency (ECRF) heating system is utilized to realize high performance plasma. Its output power is 4 MW at 110 GHz. The output power of the gyrotron used in the heating system can be controlled by changing its anode voltage. Then, a compact anode voltage controller was developed to modulate the injected power into plasmas. This controller achieved the modulation frequency of 12 - 500 Hz with modulation factor of 80 % at 0.7 MW of output power. Additional function of this controller also could make the pulse duration longer from 5 s to 16 s at 0.5 MW. For the long pulse operation, temperature rise of the DC break made of Alumina ceramics in the gyrotron was estimated from the measured temperature rise of the coolant and its maximum temperature was about 140 deg. From the analysis of this temperature rise, DC break materials should be changed to low-loss materials for extending to 30 s of the pulse duration. The stabilization of neoclassical tearing mode (NTM) was demonstrated by ECRF heating using the real-time system in which ECRF beams were injected to the NTM location predicted from ECE measurement every 10 ms. The ECRF beams were scanned poloidally with the steering mirrors of the antennas.

An ECRH (Electron Cyclotron Resonance Heating) system capable of delivering 2.4 MW CW is presently under construction at CEA (Commissariat à l'Energie Atomique) Cadarache, for the Tore Supra experiment to provide plasma heating and current drive by Electron Cyclotron Resonance interaction.

Due to some limitations observed on the first series tube which achieved 300 kW output power for 110 s, a new study carried out in a collaboration between TED (Thales Electron Devices), the Association Euratom-CEA and the Association Euratom-Confédération Suisse has led to the construction of a new modified gyrotron.

The new gyrotron, with a new launcher profile and a better cooling system is now installed in this test bed. A clear improvement in the time required to condition the tube has been observed. On the other hand poor mode purity in the output beam has resulted in the need to implement a cooling system for the waveguides transmitting the power to the dummy load. The gyrotron tests have been temporarily suspended while a new system for the automatic filling of the cryostats with liquid nitrogen and helium is being installed.

The experience gained from tests operations including some of the problems related both to auxiliary equipment and to the control of the gyrotrons will be presented with a special focus on long pulse related issues.

Electron Cyclotron Resonance Heating (ECRH) is the main heating method for the Wendelstein 7-X Stellarator (W7-X), which is under construction at IPP-Greifswald.

A 10 MW ECRH plant with CW-capability at 140 GHz is under construction to meet the scientific objectives. The microwave power is generated by 10 gyrotrons with 1 MW each two gyrotrons are operational at IPP in Greifswald. The tubes are equipped with a single-stage depressed collector for energy recovery and operate with an output power modulation between 0.3 and 1 MW with a sinusoidal frequency of up to 10 kHz which is achieved by modulating the depression voltage and is an interesting feature for NTM control at ITER. The general features of the ECRH-plant such as frequency power, cw-capability, flexibility and the experimental experience are of high relevance for the ITER system.

Each gyrotron is fed by two high-voltage sources. A high-power supply for driving the electron beam and a precision low-power supply for beam acceleration. The high-power facility consists of modular solid state HV-supplies (−65 kV 50/100 A) providing fast power control and high flexibility. The low-power high-voltage source for beam acceleration is realized by a feed back controlled high-voltage servo-amplifier driving the depression voltage. A protection system with a thyratron crowbar for fast power removal in case of gyrotron failure by arcing is installed. Both the high power and low-power high-voltage sources have the capability to supply a 2 MW ITER gyrotron without any modification. Analogue electronic devices control the fast functions of the high-voltage system for each gyrotron and a hierarchy of industrial standard PLCs and computers supervise the whole ECRH-plant.

The launcher structural design integration is presented for the '8-beamline' remote steering (RS) reference launcher. It covers the design of the blanket shield module (BSM) which closes the gap in the regular blanket structure and of the internal neutron shield in the main launcher structure. Nuclear heating rates are provided by nuclear calculations; including heat loads from the mm-wave components, a conceptual design is defined for the cooling system fed by the ITER blanket cooling water. The thermo-mechanical performance of the first wall panel which is welded to BSM side walls is proven by FEM modelling. The disassembly of the port plug and its internals is based on an axial access scheme. Boundary conditions to neighbouring ITER systems are quantified in terms of shut-down dose levels for hands-on access, helium generation in the vacuum vessel and dimensions of the cut-out required in the lower regular blanket module. The adaptation of the shielding and cooling configuration is discussed for the two advanced beamline variants which are the 6 beamline RS dogleg option and the front steering (FS) launcher.

EC launcher components for the ITER have been designed and developed. Thermomechanical analysis of a front shield module and a steering mirror for the ITER equatorial launcher were carried out. Maximum temperature and induced stress of the front shield module are estimated to be 232 °C and 351 MPa, respectively. Maximum temperature of 302 °C is expected at the steering mirror surface under the 1 MW-CW operation. Maximum induced stress at the mirror surface and the inner surface of the cooling tube are 150 MPa and 248 MPa, respectively. These values are less than allowable level. A cyclic loading test of the spiral cooling tube to supply cooling water to the steering mirror revealed the durability over one million rotational cycles. A diamond window with copper coated edge was developed as a reliable torus window in terms of safety. The result of the 1 MW relevant experiment of the window shows no degradation of the high power transmission performance. It is thermomechanically and experimentally verified that most of the EC launcher components that has been developed for the ITER are acceptable.

An ECRH (electron-cyclotron resonance heating) launching system for the ITER upper ports is being designed. The aim of the system is to inject Electron Cyclotron Waves (ECW) in the ITER plasma in order to stabilize neoclassical tearing modes (NTM). Each of the four upper-port launchers consists of six mm-wave lines capable of transmitting high power up to 2 MW per line at 170 GHz. In order to exploit the capability of ECW for localized heating and current drive over a range of plasma radii in ITER, the ECH&CD upper port launcher must have a beam steering capability. The Remote Steering (RS) principle has great advantages, because it enables to avoid steerable mirrors with flexible cooling lines at the plasma-facing end of the launcher. The principle consists of a long, corrugated, square waveguide having the steerable optics placed outside of the first confinement boundary of the vacuum vessel. All vulnerable components are far away from the hostile plasma environment. Furthermore, the RS launching system enables to do maintenance on the system during shutdown, without affecting the torus vacuum and the blanket cooling circuits.

The operational experience of present day high power EC systems is nominally only for a few seconds. This is a long way away from the thousands of seconds required for ITER. It would be beneficial to the ITER program that EC system components be tested to full parameters prior to committing to producing the full set of components. The planned growth in the EC system on DIII-D over the next few years provides the opportunity to assemble a test stand for ITER EC components. By building the DIII-D hardware to the ITER specifications it will allow ITER to gain beneficial prototyping experience on a working tokamak, prior to committing to building the hardware for delivery to ITER.

Electron Cyclotron Resonance Heating (ECRH) system will play an important role in plasma formation, heating and current drive in the Superconducting Steady state Tokamak (SST-1). Commissioning activity of the machine has been initiated. Many of the sub-systems have been prepared for the first plasma discharge. A radial and a top port have been allotted for low field side (LFS) and high field side (HFS) launch of O and X- modes in the plasma.

The system is based on a gyrotron source operating at a frequency of 82.6±0.1GHz (GLGD-82.6/0.2) and capable of delivering 0.2 MW / 1000s with 17% duty cycle. The transmission line consisting of ∼15 meters length 63.5mm corrugated wave guide, DC break, wave guide switch, mitre bend, polariser, bellows that terminates with a vacuum barrier CVD window. A beam launching system used to steer the microwave beam in the plasma volume is connected between the end of the transmission line and the tokamak radial and top ports. A VME based real time data acquisition and control (DAC) system is used for monitoring, acquisition and control. Hard-wired interlock operates a rail-gap based crowbar system in less than 10µs under any fault condition. Burn patterns are recorded at various stages in the transmission line. The gyrotron is tested for ∼200 kW / 1000s operation on a water dummy load. Transmission line is tested at various power levels for long pulse operation. The paper highlights the experimental results of successful commissioning of the system.

In the course of the development of a remote steering ECRH upper port launcher for ITER, it became clear that a modification could be introduced in the conceptual design in order to solve a number of structural weaknesses. Up to that point, all conceptual layouts were based on each remote steering beamline having a single front mirror placed in front of the square waveguide to aim the beam towards its resonance surface in the plasma. By placing an additional mirror per line inside the front shield of the upper port plug - effectively creating a dogleg routing - a number of structural issues were solved. This modification allows for a decrease of the heat load on the front mirrors and a shift downwards of the launching point. Additionally, through correct placement and focusing of the mirrors, the front shield penetration could be reduced by a factor of 4 and the cut in the blanket module below the upper port could be reduced significantly, while the level of overall performance could be increased as well. In order to visualise this new concept accurately, a more detailed design of the beam propagation was required. Through concerted effort within our institute, two different approaches were made to come to this new design; further advancements of the 3D-model and an Excel-based 2D simulation. This dual approach, together with beam tracing calculations done by affiliated institutes have indicated that the dogleg can prove to be a reliable design for a RS upper port launcher.

The optics of the ECRH Upper Launcher in ITER based on the Remote Steering concept needs special attention, since any focussing element in front of the waveguide has combined effects on the range of steering angles achievable and the beam width in the plasma region. The effects are studied in detail for a setup composed by 8 beams per port (three ports), for a spherical and a hyperbolic mirror surface. Gaussian beam analysis is compared to beam pattern calculations with the optical physics code GRASP, in order to verify the validity of gaussian optics approximation. The standard description with simply astigmatic beams, not adequate in more complex systems as the proposed two-mirror set-up, requires approximations, which are compared with the generalized astigmatic beam description. The ohmic losses at the end mirrors and the related localized heating due to the very large power density cause deformations that depends on the design of the cooling circuit. The distortion of the beam shape has been evaluated in a realistic case of mirror cooling with a small-channel system. The quantification of the effect depends on the precise evaluation ohmic losses and their enhancement in the long term due to the surface deterioration.

This paper reports the results of the high-power test of a remote steering launcher mock-up at 140 GHz, which were performed at the ECRH installation for the future stellarator W7-X at IPP Greifswald. The mock-up test system consists of a 6.62 m long corrugated square waveguide with a steerable optic at the entrance and various diagnostics at the exit of the waveguide. A straight and a dog-leg version of the launcher were investigated.

The high-power tests of the straight setup have been performed with powers up to P₀ = 700 kW (typically 500 kW) and pulse lengths of up to 10 seconds.

For both polarizations (parallel and perpendicular to the steering plane), no arcing was observed in spite of the fact, that the experiments were performed under ambient atmospheric conditions. After the integration of 2 mitre bends in the setup, arcing limited the usable parameter range.

The ohmic loss P_Ω of the waveguide was measured via the temperature increase of the waveguide wall, and was used to calibrate the calculated angular dependence of the total ohmic losses of the waveguide. Short-pulse radiation pattern measurements with thermographic recording show high beam quality and confirm the steering range of −12° < ϕ < 12°.

The long pulse capability of Tore Supra and its ECRH system makes it an ideal machine to prove steady state feedback control as required in ITER. Although Neoclassical Tearing Modes (NTMs) have not yet been observed on Tore Supra, the control of other MHD modes represents a very similar task from a control point of view and the stabilisation of such modes for long periods using ECRH will provide essential experience for the implementation of such control schemes on ITER. For this work to progress on Tore Supra, it must be possible to vary the injection angles in real time under feedback control from measured plasma parameters. At Tore Supra the front mirror position - and hence the injection angles - is adjusted using stepper motors controlled through a serial link. The use of a serial link limiting the sampling time for the control system to 50-100 ms and the dynamic response of the stepper motors results in a system frequency response <5 Hz. For following the evolution of the current profile this seems fully acceptable though it could prove somewhat slow for reacting to fast beta changes. In any case the time constants typically associated with growth rate of NTMs on ITER would not require a significantly faster system. Initial commissioning of such real time control was performed in 2004. Unfortunately, following a limited number of cycles and prior to using the system for plasma experiments, one water bellows in the antenna ruptured indicating a major weakness in the design of the water-cooling system for the antenna. As a consequence the water-cooling of the mobile mirror has been redesigned. Careful calculations and subsequent tests have been used to optimise the trajectories of the flexible water connections and a more robust layout is being implemented, ready for operation in the summer 2005.

The ECRH ITER upper port antenna's role is to stabilize the neoclassical tearing mode (NTM) on either the q = 2 or 3/2 rational flux surfaces, which requires a narrow current deposition profile (j_CD) over a wide range along the resonance surfaces. The width of j_CD should be equivalent to the marginal island width to fully stabilise the NTM. Two antenna concepts are under consideration for the upper port launcher: front steering (FS) and remote steering (RS). The FS launcher decouples the steering and focusing aspects using a two mirror system (one focusing and one steering), achieving a wider steering range and higher current density for NTM stabilisation than required by ITER, offering a threefold increase in NTM stabilization efficiency over the RS launcher. The improved physics performance has motivated the further design study of the FS launcher aiming toward a build to print launcher. The present design is compatible with ⩾2.0 MW CW operation and 8 beams per port plug. A frictionless backlashfree system is envisioned for the steering mechanism. An overview of the launcher design, the calculated physics performance and the possibility of using the upper port launcher for extended physics applications (beyond NTM stabilisation) are discussed.

A FS launcher is being designed for the ITER upper port, which offers enhanced physics performance over the RS launcher. A two mirror system is used to decouple the focusing and steering aspects of the launcher and provide a relatively small beam waist (<20 mm) projected far into the plasma (>1.6 m from the steering mirror). The resulting NTM stabilization efficiency (maximum CD density divided by the local bootstrap current >1.6) is above marginal for the q = 2 and 3/2 rational flux surfaces of the relevent ITER equilibria (scenarios 2, 3a and 5) and a factor of ∼3 relative to an equivalent RS launcher. The performance of the FS launcher strongly depends on the reliability of the steering mechanism, which is used to rotate the plasma facing steering mirror. CRPP has designed a frictionless steering mechanism assembled in a compact cartridge capable of up to ±10° rotation (corresponding to a poloidal steering range of up to ±20° for the microwave beam around a fixed axis of rotation) that offers a high operation reliability despite the close proximity to the thermal and neutron flux coming from the ITER plasma.

The association FOM-Rijnhuizen is developing a plug-in 170 GHz ECW launcher system for use in the upper-ports of ITER, based on the remote steering (RS) principle. The launching system enables up to eight mm-wave beams per port plug to be directed from the feeding waveguides through corrugated square waveguides (≈44 × 44 mm²) inside the port plug to the plasma. A mock-up system has been manufactured, capable of vacuum & high-power short-pulse operation. The transmission line consists of a taper, to launch an HE₁₁ mode in 63.5 mm circular corrugated waveguide. The circular waveguide is used as input for the steering mechanism, housed in a vacuum vessel, which directs the beam under an angle between −12° and 12° into the copper square corrugated waveguide. Antenna patterns from the exit aperture of the square waveguide are scanned using a 3-axis scanner.

The reference diameter for the ECH transmission lines is presently 63.5 mm. Analyses of the heat generation and removal in 63.5 mm corrugated waveguide components were reported [R.A. Olstad, et al., Proc. of IAEA TM on ECRH Physics and Technology for ITER, Kloster Seeon, Germany, 2003, http://ipp.mpg.de/tmseeon; R.A. Olstad et al., ''ECH MW-level CW Transmission Line Components Suitable for ITER,'' to be published in Fusion Eng. & Design (2005)]. Those analyses concluded that the temperature of all components could be kept to acceptable levels, even with operation at 2 MW cw per transmission line. Recently interest has been expressed in the community about the possible advantages of using a smaller diameter waveguide for ITER, particularly because of limited space available at both the equatorial and upper launchers. In addition to ameliorating the space constraints, there could be large cost savings for a modest diameter reduction in certain transmission line components, particularly gate valves and CVD diamond window assemblies at the entrance to the launchers.

Results of a tradeoff study on ITER waveguide diameter are reported. Diameters considered range from 45 mm to 63.5 mm; consideration is also given to the possibility of tapering down to 31.75 mm at the launchers. The most critical issue for smaller diameter components is the increased losses and increased power densities. These lead to more demanding cooling provisions and higher operating temperatures for components such as miter bends, power monitor miter bends, bellows, dc breaks, waveguide switches, and waveguide sections adjacent to miter bends. In addition, the overall transmission efficiency of the ITER transmission lines would be reduced.

For high power transmission in ECRH systems, window materials have to combine ultra low mm-wave loss (tanδ < 10⁻⁴) with an outstanding resistance against crack formation. Specially challenging are the CVD diamond ''torus'' windows which form the primary tritium confinement to the mm-wave system of the ECRH Upper Launcher for plasma stabilization at ITER and which have to be designed as a compact structure allowing transmission of 2 MW beams at 170 GHz. Special emphasis is given to the window development for the remote steering (RS) launcher, as in contrast to the alternative front steering (FS) launcher, it requires additional capability for off-axis transmission.

Radiation shielding analyses have been performed for the ECRH system in the ITER upper port to complete the neutron streaming analysis performed previously. The analyses aimed at assessing and optimising the performance of the radiation shield to prove that the shielding requirements can be met by the proposed design variants. The radiation transport calculations have been performed by means of the Monte Carlo programme MCNP in 3D geometry using the standard ITER neutronics model with ECRH launcher and plug integrated into the upper port. The interface programme MCAM was used to convert the 3D ECRH launcher models available from the CAD-system for use with MCNP-calculations. It was shown that the launcher design with the proposed radiation shield can satisfy the design limits for the radiation loads to both the launcher and the neighbouring components such as the Vacuum Vessel and the TF coils. Radiation dose levels were assessed for reactor shutdown at the rear side of ECRH launcher at locations where personnel access for maintenance may be required. The shutdown dose rate calculations were also performed in 3D geometry by applying the rigorous 2-step (R2S) method and comparing the results to those obtained with the direct 1-step (D1S) method. The R2S method includes activation calculations for the launcher materials by means of the inventory code FISPACT. It was proven that the shutdown dose rates inside the port with straight waveguides will be below the ITER radiation limit of 100 µSv/hr after 10 days decay time.

The potential of continuous plasma sustainment of Large Helical Device (LHD) was successfully demonstrated by a one hour discharge with 110 kW electron cyclotron heating (ECH) power. The ECH power of frequency 84 GHz generated by a continuous-work (CW) gyrotron is transmitted through an evacuated waveguide transmission line and injected to the LHD vacuum vessel using waveguide antenna. The plasma density was kept at about 1.5 × 10¹⁸/m³ and the electron temperature at the plasma center over 1 keV. Due to the low injection power the density was not so high but the plasma was quite stable. The power injection was terminated manually at 3900 seconds from the limitation of the setting of data acquisition, not for any troubles on devices.

In this paper the results of the experiments of combined injection of LH waves and EC waves performed in the FTU tokamak are reported. Such experiments were mainly devoted to study and to control the access to the advanced tokamak scenarios in plasma conditions close to ITER parameters (B_t≈5T or higher, n_e,line≈10²⁰ m⁻³). Two different absorption mechanisms were used for the EC waves. In the first one the cold resonance absorption of EC waves launched with a toroidal angle was used to induce small modification of the current profile mainly maintained by LH waves. In the second one the absorption through Doppler shift due to the fast electron tails generated by LHCD was used.

In the moderate magneticfield of TCV (1.5T), the recently installed X3 system (3 gyrotrons @118 GHz, 0.45 MW/each, 2s) broadens the operational space with the possibility of heating plasmas at high density, well above the cutoff density of the X2 system (X2 cutoff at n_e = 4.2 × 10¹⁹m⁻³). To compensate the significantly weaker absorption coefficient compared to the absorption of X2, the top-launch injection allows to maximize the ray path along the resonance layer thus maximizing the optical depth. To maintain the maximum absorption in plasma discharges with a dynamic variation of both density (refraction) and temperature (relativistic shift) a real time control system on the poloidal injection angle has been developped and succesfully tested on TCV. With a total injected power of 1.35 MW and using the mirror realtime control, full-single pass absorption has been measured in an L-mode plasma. A significant fraction of the absorbed power is associated to the presence of suprathermal electrons generated by the X3 wave itself. In X3 heating experiments of H-mode plasmas it has been possible to enter into a different ELMy regime compared to the ohmic/low-power-heating ELMy regime. In these experiments a significant increase of the plasma energy is obtained with nearly fullsingle pass absorption. Results on the comparison of the absorbed fraction calculated with the TORAY-GA ray-tracing code and the beam-tracing code, ECWGB, which includes difiraction effects, are discussed.

Proof of principle of feed-back controlled Electron Cyclotron Heating and Current Drive (ECH/ECCD), aiming at automatic limitation (or suppression) of Neoclassical Tearing Modes amplitude, has been achieved in a number of present machines. In addition to Neoclassical Tearing Mode stabilization, more applications of well-localized ECH/ECCD can be envisaged (saw-tooth crash control, current profile control, thermal barrier control, disruption mitigation). However, in order to be able to take a step forward towards the application of these techniques to burning plasmas, some crucial issues should be more deeply analyzed: multi-beam simultaneous action, control of deposition radii r_dep, diagnostic of plasma reaction. So far the Electron Cyclotron Emission has been the most important tool to get localized information on plasma response, essential for both r_dep and r_island recognition, but its use in very hot burning plasmas within automatic control loops should be carefully verified. Assuming that plasma response is appropriately diagnosed, the next matter to be discussed concerns how to control r_dep, since all techniques so far used, or proposed (plasma position, toroidal field, mechanical beam steering, gyrotron frequency tuning) have limitations or drawbacks. Finally, simultaneous multiple actions on many actuators (EC beams), concurring to automatic control of one single parameter (e.g. NTM amplitude) might be a challenging task for the controller, particularly in view of the fact that any effect of each beam becomes visible only when it is positioned very close to the right radius. All these interlinked aspects are discussed in the paper.

The performance of different options for the ITER ECRH upper launcher is analysed in terms of NTM stabilisation efficiency, which is the main task for this system. The driven current at the (3,2) and (2,) resonant surfaces of ITER scenarii 2, 3a and 5 is calculated and compared with the local bootstrap current density there, which is the drive for the NTM. It is found that in terms of the figure of merit for NTM stabilisation, j_ECCD/j_bs, a launcher concept based on front steering has much better performance than concepts based on remote steering. This can be explained by the smaller spot size of the ECRH beam at the resonant surfaces in the case of front steering. Thus, from a physics point of view, the analysed front steering option is preferable to the remote steering option.

The capabilities of the ITER ECRH upper port launcher to drive a well localized current (co-ECCD) either at the q = 1.5 or at the q = 2 surface by injecting EC waves as first harmonic O-mode at the nominal frequency f = 170 GHz have been already explored for a number of ITER scenarios relevant for neoclassical tearing modes (NTM) stabilization. The analysis, made by including the hybrid scenario 3 and a low q scenario (scenario 5) as well as equilibrium and poloidal beta variations in addition to the Q = 10 reference scenario 2, explored a wide range of relevant surfaces with different kinetic data for each scenario. "Optimal" toroidal and poloidal injection angles for application of co-ECCD on the relevant flux surfaces have been identified for the two rows of mirrors, and performance evaluations by taking into account several beam parameters have been carried out. However, the analysis so far has been done only for one specific value of the vacuum toroidal magnetic field (B₀ ∼5.3 T at R₀ = 6.2 m). In this work the capabilities to drive efficient and well localized co-current for a range of relevant surfaces in ELMy-H mode ITER plasmas at lower magnetic fields are explored by an updated version of the Milano beam tracing code. On the beam trajectories fully relativistic EC absorption and CD are calculated at both first and second harmonic. The changes of the "optimal" launch angles as a consequence of the shift of the resonance to lower values of major radius are shown. The range of magnetic fields for which power deposition and current drive at relevant surfaces becomes impossible for EC waves injected from upper launcher as first harmonic O-mode is pointed out. Further, the effectiveness of the upper launcher for stabilization of (3,2) and (2,1) NTM is investigated in the range of magnetic fields where EC waves at f = 170 GHz should be injected as second harmonic X-mode.

Tokamaks designed for burning plasma operation have significant free energy in the poloidal magnetic field and in the thermal stored energy. Protection of the first wall and the vacuum vessel from the effects of a rapid release of this energy (a disruption) is required. While mitigation of a disruption is feasible, avoidance of the disruption is preferable. Suppression of tearing modes that lead to disruption is a key method of avoidance. Electron cyclotron current drive is a demonstrated technique for suppression of tearing modes. The location of the current drive relative to the tearing mode is the critical parameter for successful suppression. Incorporation of this suppression technique in a machine protection system requires continuous aiming of the electron cyclotron waves (for rapid application to island) and a closed-loop optimization of the effect of the current drive on the mode (for maximum effectiveness). Real-time methods to accomplish both of these tasks have been demonstrated successfully in the DIII-D tokamak.

The TORAY-GA code has been used to calculate the effectiveness of electron cyclotron waves for applications on the International Tokamak Experimental Reactor (ITER) using the steerable launchers of the reference design. This study focuses on the effect on the electron cyclotron current drive of operation of ITER over a range of toroidal magnetic field B_T from 2.35 to 5.3 T. For B_T above 4.3 T and for B_T near 2.6 T where the second harmonic takes the place of the fundamental, the launchers are satisfactory for addressing the main objectives of ECCD in ITER. Between about 4.3 T and 3.3 T, the top launcher is not usable and the midplane launchers can only be used to deposit heat near the outer part of the plasma. The applications for which the midplane launchers can be used for B_T between 3.3 T and 2.6 T depend rather sensitively on the equilibrium and the application

This paper describes a form of the generalised Rutherford Equation for Neoclassical Tearing Modes suited to fit experiments. All the relevant terms contributing to the perturbed current parallel to the magneticfield have been included both in the cases of small or large island limit; we present a fit to ASDEX Upgrade experimental data.

We discuss the use of step-tuneable gyrotrons in present day and future tokamak devices. It is shown that with a two-frequency system, the performance of the planned ITER ECCD system, using the present launcher designs, could be greatly enhanced. With step-tuneable gyrotrons, it should be possible to use fixed launchers without any steering at all.

The influence of steering errors on Neoclassical Tearing Mode Control by means of the ITER ECRH Upper Launcher is numerically and statistically studied and recommendations are given for the alignment. Changes of density and temperature profiles are shown to have a negligible effect on the launch angles and to affect the driven current according to a well known scaling of current drive efficiency. Finally, a study of the radial positions accessible by the Upper and Equatorial Launchers is presented.