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Non‐oxidative methane activation is carried out in a microwave plasma reactor for coupling to higher hydrocarbons. Fourier transform infrared spectroscopy (FTIR) was used to measure absolute concentrations of the major hydrocarbon species. Hydrogen concentration was also independently inferred from pressure‐based change in molar flow measurements. By closing both the carbon and hydrogen balance, from stoichiometry of the reactions, the amount of deposits was obtained as well. Additionally, core gas temperatures up to 2500 K were measured with Raman scattering when nitrogen acted as probing molecule in sample mixture discharges. At low gas temperatures, ethane and ethylene were significant products based on plasma chemistry, with ethane selectivities reaching up to 60%. At higher gas temperatures, thermal effects become stronger shifting the selectivity toward acetylene and deposits, resembling more with equilibrium calculations. The energy efficiency of the methane conversion reached up to 15% from which 10% represented coupling efficiency to higher hydrocarbons. It is concluded that there is an interplay between plasma and thermal chemistry where plasma generates radicals and final distribution is set by thermodynamics. CH4 ->plasma-> CH3 ->coupling-> C2H6 ->dehydrogenation-> C2H5 ->….

A suite of diagnostics is proposed to characterize microwave plasma dissociation of CO2: laser scattering, Fourier transform infrared spectroscopy, and passive emission imaging. It provides a comprehensive performance characterization as is illustrated on the basis of experiments in a 2.45 GHz, 1 kW microwave reactor with tangential gas injection. For example, two operating regimes are identified as function of pressure: the diffuse and constricted plasma mode. Their occurrence is explained by evaluation of microwave propagation, which changes with the electron-heavy particle collision frequency ve−h. In the diffuse mode, gas temperatures of 1500–3500 K are determined. The measured conversion degree, specific energy input, and temperature are summarized in a two-temperature thermal model, which is solved to obtain the gas temperature at the periphery of the reactor and the size of the hot zone. Solutions are found with edge temperatures of hundreds of K, and hot zone fractions which agree with the measured behavior. The agreement shows that non-thermal processes play only a marginal role in the measured parameter space of the diffuse discharge. In the constricted mode, the radial plasma size is independent of power. A skin depth equal to the plasma size corresponds to electron densities of 1018–1019 m−3. Temperatures in the central filament are in the range 3000–5000 K. Both discharge modes are up to 50% energy efficient in CO production. Rayleigh signals increase in the afterglow, hinting at rapid gas cooling assuming that the gas composition remains unchanged.

New linear plasma devices are currently being constructed or planned in the Trilateral Euregio Cluster (TEC) to meet the challenges with respect to plasma surface interactions in DEMO and ITER: i) MAGNUM-PSI (FOM), a high particle and power flux device with super-conducting magnetic field coils which will reach ITER-like divertor conditions at high magnetic field, ii) the newly proposed linear plasma device JULE-PSI (FZJ), which will allow to expose toxic and neutron activated target samples to ITER-like fluences and ion energies including in vacuo analysis of neutron activated samples, and iii) the plasmatron VISION I. a compact plasma device which will be operated inside the tritium lab at SCK-CEN Mol, capable to investigate tritium plasmas and moderately activated wall materials. This contribution shows the capabilities of the new devices and their forerunner experiments (Pilot-PSI at FOM and PSI-2 Julich at FZJ) in view of the main objectives of the new TEC program on plasma surface interactions. (C) 2011 Forschungszentrum Julich, Institut fur Energieforschung-Plasmaphysik. Published by Elsevier B.V. All rights reserved.

A basic requirement for ITER equipment to meet is a high level of reliability, because ITER operation time is precious and radioactive operation leaves limited scope for repair. In order to reduce the risk of failure during ITER operation an effective risk mitigation strategy is necessary. This paper presents such strategy for the ITER electron cyclotron upper port launcher (ECUPL). A preliminary ECUPL risk analysis identifies possible failure modes. A probabilistic risk assessment quantifies the risk of failure using a 4 x 4 impact-likelihood matrix. Impact is quantified through technical, cost and schedule elements. Likelihood depends on the risk mitigation strategy adopted. A cost benefit analysis determines the most cost effective risk mitigation strategy. An essential element in risk mitigation is the testing of equipment prior to installation on the ITER machine. This paper argues the need for low- and highpower millimetre wave tests carried out on the fully assembled ECUPL It presents a conceptual design for a dedicated on-site test bed that can handle 2 of 8 microwave beams at 2 MW long pulse operation. (C) 2010 Elsevier B.V. All rights reserved.

The FOM-Institute for Plasma Physics Rijnhuizen is constructing Magnum-PSI; a magnetized (3 T), steady-state, large area (80 cm2) high-flux (up to 1 024 H+ ions m-2 s-1) plasma generator. Magnum-PSI will be a highly accessible laboratory experiment in which the interaction of magnetized plasma with different surfaces can be studied. This experiment will provide new insights in the complex physics and chemistry that will occur in the divertor region of the future experimental fusion reactor ITER. Here, extremely high power and particle flux densities are predicted at relatively low plasma temperatures. Magnum-PSI will be able to simulate these detached ITER divertor conditions in detail. In addition, conditions can be varied over a wide range, such as different target materials, plasma temperatures, beam diameters, particle fluxes, inclination angles of target, background pressures, magnetic fields, etc., making Magnum-PSI an excellent test bed for high heat flux components of future fusion reactors. The design phase of the Magnum-PSI device has been completed. The construction and assembly phase of the device is in progress. In this contribution, we will present the design and construction of the Magnum-PSI experiment. The status of the vacuum system, the 3 T superconducting magnet, the plasma source, the target plate and manipulator, and additional plasma heating will be presented. The plasma and surface diagnostics that will be used in the Magnum-PSI experiment will be introduced. © 2010.

A design is presented for the electron cyclotron (EC) heating and current drive system of the ITER upper port launchers based on the remote steering (RS) concept. In this concept the millimeter-wave beam is steered by a mirror that is located at the back end of the launcher waveguide. The RS concept has the advantage that the mirror steering mechanism can be situated in the secondary vacuum of the ITER machine where neutron flux and beryllium and tritium contamination is reduced. This allows simpler maintenance relative to a system with a plasma-facing steering mechanism. The optimization is carried out on the quasi-optical elements of the system, including the mirror shapes and positions. The design is assessed for its effectiveness in stabilizing the neoclassical tearing mode (NTM) over a wide range of ITER reference scenarios. The stabilization performance is quantified in terms of the parameter eta(ntm), expressing the ratio between the peak EC wave-driven current density and the bootstrap current density, which parameter should exceed 1.2. The performance is also evaluated in terms of beam-focusing properties and power loading on the mirrors, and an empirical relation between beam size and eta(ntm) has been established. The performance achieved meets the requirements for NTM stabilization in all but one of the ITER reference scenarios.

The current status of the mechanical design of the remote steering electron cyclotron resonance heating upper port launching system for ITER is presented. Although an alternative front steering launcher has now been selected as the reference design for ITER, the development of a remote steering launcher continues so that it can be used as a backup solution and as a candidate for DEMO. Since earlier proposals of a remote steering launcher could not fulfill the design criteria with respect to physics performance and because a number of engineering issues remained that have proven to be very difficult to solve, a change was applied to its layout. By increasing the length of the square waveguides that form the heart of the remote steering design, the layout of the optics could be further optimized so that the performance could be improved, while a number of engineering issues could be solved. This paper provides a brief description of the previous design followed by the modifications taken in the optical design to improve the physics performance by reducing the beam size at the resonance location. A first indication is given that the expected reduction of beam size at the resonance location is more than 30%, relative to earlier designs.

FOM is developing a full-scale mock-up of the mm-wave upper-port launching system for electron cyclotron resonance heating in ITER. The test set-up foresees in most of the typical ITER operation conditions, e.g. primary vacuum and the secondary vacuum, blanket cooling and component cooling conditions. All relevant operational ITER conditions except neutron radiation can be simulated, including baking conditions. The fabrication of the launching system mock-up will be the last step in the development phase, which will be characterized by the fabrication of test samples and mm-wave tests and calculations. The parameterized modeling of mm-wave beam propagation in CATIA [D.M.S. Ronden, et al., Parameterized modeling of mm-wave beam propagation of the ITER ECRH remote steering upper port launcher in CATIA, this conference] (8th figure in this article) the test and calculation results will be used for the built to print design of the remote steering ECRH launcher for the ITER upper ports [A.G.A. Verhoeven, et al., Design of the Remote Steering ITER ECRH Upper Port Launcher, this conference].

The millimeter-wave system of the remote steering launcher at the upper port level is composed of beam-lines that are rated for 2-MW continuous-wave operation at 170 GHz. In each beamline, a torus window is located between the entrance to the in-vessel square corrugated waveguide and the steerable mirrors in the launcher back end. In the reference design, the maximum steering angle of 12 deg imposes a 27-mm off-center beam shift to the window disk center, which in turn leads to asymmetrical heating of the window. This raises particular concerns of enhanced thermomechanical stresses in the window and in the metallic window cuffs. In order to qualify the optical, mechanical, and thermohydraulic design, high-power short-pulse and thermohydraulic tests were performed using a prototype chemical vapor deposition diamond torus window developed and manufactured at Forschungszentrum Karlsruhe. It was proven that arcing did not occur even under maximum millimeter-wave power levels available (up to 0.53 MW) and that the millimeter-wave beam profile was fully maintained. A test facility allowed thermohydraulic studies of the window cooling system with parameters characteristic for component cooling water loops at ITER (p w =1.0 MPa, T w = 40°C).

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 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 strong non-equilibrium conditions provided by the plasma phase offer the opportunity to beat traditional thermal process energy efficiencies via preferential excitation of molecular vibrations. Simple molecular physics considerations are presented to explain potential dissociation pathways in plasma and their effect on energy efficiency. A common microwave reactor approach is evaluated experimentally with Rayleigh scattering and Fourier transform infrared spectroscopy to assess gas temperatures (exceeding 104 K) and conversion degrees (up to 30%), respectively. The results are interpreted on a basis of estimates of the plasma dynamics obtained with electron energy distribution functions calculated with a Boltzmann solver. It indicates that the intrinsic electron energies are higher than is favorable for preferential vibrational excitation due to dissociative excitation, which causes thermodynamic equilibrium chemistry to dominate. The highest observed energy efficiencies of 45% indicate that non-equilibrium dynamics had been at play. A novel approach involving additives of low ionization potential to tailor the electron energies to the vibrational excitation regime is proposed.

A storage scheme for Renewable Energy (RE) based on the plasmolysis of CO2 into CO and O2 has been experimentally investigated, demonstrating high energy efficiency (>50%) combined with high energy density, rapid start-stop and no use of scarce materials. The key parameter controlling energy efficiency has been identified as the reduced electric field. Basic plasma parameters including density and temperature are derived from a simple particle and energy balance model, allowing parameter specification of an upscale 100 kW reactor. With RE powered plasmolysis as the critical element, a CO2 neutral energy system becomes feasible when complemented by effective capture of CO2 at the input and separation of CO from the output gas stream followed by downstream chemical processing into hydrocarbon fuels.

The construction phase of the linear plasma generator Magnum-PSI at the FOM institute DIFFER has been completed and the facility has been officially opened in March 2012. The scientific program to gain more insight in the plasma–wall interactions relevant for ITER and future fusion reactors has started. In Magnum-PSI, targets of a wide range of materials and shapes can be exposed to high particle, high heat flux plasmas (>1024 ions m−2 s−1; >10 MW/m2). For magnetization of the plasma, oil-cooled electromagnets are temporarily installed to enable pulsed operation until the device is upgraded with a superconducting magnet. The magnets generate a field of up to 1.9 T close to the plasma source for a duration of 6 s. Longer exposure times are available for lower field settings. Plasma characterizations were done with a variety of gases (H, D, He, Ne and Ar) to determine the machine performance and prepare for subsequent scientific experiments. Thomson scattering and optical emission spectroscopy were used to determine the plasma parameters while infrared thermography and target calorimetry were used to determine the power loads to the surface. This paper reports on the status of Magnum-PSI and its diagnostic systems. In addition, an overview of the plasma parameters that can be achieved in the present state will be given.

Power-to-gas is a storage technology aiming to convert surplus electricity from renewable energy sources like wind and solar power into gaseous fuels compatible with the current network infrastructure. Results of CO2 dissociation in a vortex-stabilized microwave plasma reactor are presented. The microwave field, residence time, quenching, and vortex configuration were varied to investigate their influence on energy- and conversion efficiency of CO2 dissociation. Significant deterioration of the energy efficiency is observed at forward vortex plasmas upon increasing pressure in the range of 100 mbar towards atmospheric pressure, which is mitigated by using a reverse vortex flow configuration of the plasma reactor. Data from optical emission shows that under all conditions covered by the experiments the gas temperature is in excess of 4000 K, suggesting a predominant thermal dissociation. Different strategies are proposed to enhance energy and conversion efficiencies of plasma-driven dissociation of CO2.

High-power millimetre wave beams employed on ITER for heating and Current drive at the 170 GHz electron cyclotron resonance frequency require agile steering and tight focusing of the beams to suppress neoclassical tearing modes. This paper presents experimental validation of the remote steering (RS) concept of the ITER upper port millimetre wave beam launcher. Remote steering at the entrance of the upper port launcher rather than at the plasma side offers advantages in reliability and maintenance of the mechanically Vulnerable steering system. A one-to-one scale mock-up consisting of a transmission line, mitre bends, remote steering unit, vacuum window, square corrugated waveguide and front mirror simulates the ITER launcher design configuration. Validation is based on low-power heterodyne measurements of the complex amplitude and phase distribution of the steered Gaussian beam. High-power (400 kW) short Pulse (10 ms) operation under vacuum, diagnosed by calorimetry and thermography of the near- and far-field beam patterns, confirms high-power operation, but shows increased power loss attributed to deteriorating input beam quality compared with low-power operation. Polarization measurements show little variation with steering, which is important for effective current drive requiring elliptical polarization for O-mode excitation. Results show that a RS range Of LIP to -12 to +12 can be achieved with acceptable beam quality. These measurements confirm the back-up design of the ITER ECRH&CD launcher with future application for DEMO. (C) 2009 Elsevier B.V. All rights reserved.

A Fast Fourier Transform (FFT) based wide range millimeter wave diagnostics for spectral characterization of scattered millimeter waves in plasmas has been successfully brought into operation. The scattered millimeter waves are heterodyne down-converted and directly digitized using a fast analog-digital converter (ADC) and a compact Peripheral Component Interconnect (cPCI) computer. Frequency spectra are obtained by FFT in the time domain of the intermediate frequency signal. The scattered millimeter waves are generated during high power Electron Cyclotron Resonance Heating (ECRH) experiments on the TEXTOR Tokamak and demonstrate the performance of the diagnostics and, in particular, the usability of direct digitizing and Fourier transformation of millimeter wave signals. Major benefit of the new diagnostics is a tunable time and frequency resolution due to post-detection, near-Real-Time processing of the acquired data. This diagnostics has a wider application in astrophysics, earth observation, plasma physics and molecular spectroscopy for the detection and analysis of millimeter wave radiation, providing high-resolution spectra at high temporal resolution and large dynamic range. Such a diagnostics also has the potential to detect Electron Cyclotron Emission (ECE) and to be used in Real-Time ECE feedback control systems. © 2009 IEEE.

The design of a remote steering (RS) variant of the electron cyclotron heating (ECH) upper port launcher (UPL) for neo-classical tearing mode (NTM) stabilization in ITER is being finalised. The RS variant is the fall-back option for the reference design, which is a front steering (FS) UPL. This paper presents the changes made in the design of the RS UPL with respect to the previous baseline (the October 2005 baseline) design. The changes are motivated by a demand to improve the reliability of the launching system by solving a number of pressing issues. To accomplish this, a reduction in the steering range of the launcher was proposed. This reduction yields more tolerable head loads on the focusing elements for a given performance. Modifications of the launcher layout allow for an increase of the toroidal injection angle, which improves the NTM stabilization efficiency. A mode converter can reduce the stray radiation in the vicinity of the steering mechanism down to tolerable levels. Abandoning the concept of the mini-duct leads to a reduction of stray radiation concerns and allows for the implementation of improved alignment facilities of the individual steerable mirror units. The work shows that a launcher based on the RS principle is feasible. This launcher can be made reliable, flexible and easily accessible for maintenance purposes accepting the consequent limitations in the achievable steering range inside the ITER plasma. (C) 2009 Elsevier B.V. All rights reserved.

Electron cyclotron current drive will become the main scheme on ITER for the stabilization of neoclassical tearing modes (NTMs) and the control of sawtooth oscillations. The effectiveness of this scheme forms the basis for the requirements of the ITER Upper Port Launcher. These requirements include the need for steering the mm-wave beam to the pertinent magnetic flux surfaces where the instability develops. Several concepts have been explored; front steering (FS) where a system of movable mirrors is situated at the plasma-facing end of the launcher; remote steering (RS) where the movable mirror system is situated at the back end of the launcher system. The advantage of the RS concept is that moving parts are placed away from the plasma in a secondary vacuum system, less exposed to plasma neutron flux, thus offering advantages in reliability and safety of the machine. Also, contamination of the beam steering unit by Be and T is less of a problem. However, beam-focusing properties of the RS launcher deteriorate with increased steering angle, which limits the NTM stabilization performance as compared with FS. In this paper, recent improvements in the RS design are presented. The stabilization efficiency calculated reaches values that meet most of the ITER requirements. Improvements include non-spherical front mirrors and a tapered square corrugated waveguide. Rather than steering all mm-wave beams over the entire vertical height range specified, RS launchers with a limited steering range are explored, yielding a performance improvement of 25% with an acceptable mm-wave heat load of 5.7 MW m−2 on the front mirrors. This result leads to a dual launcher concept with one set of beams dedicated to control NTMs at the outward lying q = 2 surface and another set of beams dedicated to control NTMs at the inward lying q = 3/2 surface in combination with sawteeth control at the innermost q = 1 surface.

An electron cyclotron emission (ECE) receiver inside the electron cyclotron resonance heating (ECRH) transmission line has been brought into operation. The ECE is extracted by placing a quartz plate acting as a Fabry-Perot interferometer under an angle inside the electron cyclotron wave (ECW) beam. ECE measurements are obtained during high power ECRH operation. This demonstrates the successful operation of the diagnostic and, in particular, a sufficient suppression of the gyrotron component preventing it from interfering with ECE measurements. When integrated into a feedback system for the control of plasma instabilities this line-of-sight ECE diagnostic removes the need to localize the instabilities in absolute coordinates. (C) 2008 American Institute of Physics. [DOI: 10.1063/1.2976665]

Torus window prototypes for electron cyclotron (EC) heating and current drive (H&CD) systems are developed for different launcher concepts in accordance with the specific transmission requirements. For ITER, single-disk windows are designed for 2 MW power transmission at fixed frequency (170 GHz) in two variants. For the RS launcher, the remote steering unit is placed close to the torus window in the launcher back-end which calls for a large window aperture (95 mm) to avoid beam vignetting at the extreme steering angles of ±12°. For the FS launcher with the steering mechanism placed in the front shield of the launcher, the aperture and the disk thickness are reduced as the torus window is integrated into a cylindrical waveguide configuration with an inner diameter of 63.5 mm. The reduced absorbed power and cooling paths allow to consider indirect cooling as an alternative to edge cooling which eliminates the risk of tritium contact to the cooling water in case of crack formation in the diamond disk. For the multifrequency torus window at ASDEX Upgrade (aperture: 80 mm), a double-disk configuration is realised with the disk separation of 5 mm, fine tuneable over ±1 mm. Prototypes of the high power windows were manufactured. First results of the current thermo-hydraulic and of the mm-wave performance tests support the feasibility of the concepts. For the ITER torus windows, the tool development is described which provides on-site replacement by automated cutting and welding tools.

At the TEXTOR tokamak a 140 GHz, 800 kW, 10 s gyrotron is employed for studies on ECW heating and ECCD. A key program is the suppression of tearing modes, which is aided by the unique facility at TEXTOR to program magnetic structures during a discharge (DED). A scheme is under development which aims to detect the perturbations on the electron cyclotron emission caused by these instabilities via the same line of sight as is used by the ECRH beam, and to use this information for feedback control of the instability. A brief overview of TEXTOR and relevant island parameters is given as well as a description of the upgraded ECRH system. This is followed by a description of the design of the transmission line coupler and receiver for island detection which constitutes the main topic of the paper. The final section sketches the feedback controller that processes the receiver signals, and controls the moveable launcher and the gyrotron power.

A 3D model has been developed in CATIA that is capable of accurate visualization of a mm-wave beam along the entire free-space section of a transmission line. This development has been applied to the design of the remote steering ECRH upper port launching system for ITER. Because the model uses parameterized equations to calculate the quasi-optical beam propagation, it can show the effects of a change of any of the transmission line components (location, orientation, or optical properties). Furthermore, it takes into account the effects the square waveguide has on the propagating beams as a function of its steered angle. The beam size at its target area inside the ITER plasma has proven to be a good indicator of the launcher's performance. It has also become clear that the 3D visualisation of the beam propagation in the structural design can result in a much more efficient use of the confined space inside each upper port plug—space that is also needed to locate sufficient neutron shielding and cooling structures, as well as to allow access for remote-handling tools.

The design of the mm-wave system for the ITER upper ports is being carried out with the aim to inject Electron Cyclotron Waves (ECW) in the plasma in order to stabilize neoclassical tearing modes (NTM). Four upper ports are equipped with six mm-wave lines, each capable of transmitting up to 2 MW at 170 GHz. The required beam-steering capabilities es in our approach are achieved by Remote Steering (RS). New design efforts are presented that improve the current-drive capability of the RS set-up substantially by doubling the square waveguide length to around 9 m.

The association FOM-Rijnhuizen, together with EFDA ECHULA group, develops a plug-in 170 GHz ECW launcher system, based on the remote steering (RS) principle. At this principle the steerable mirror is placed before the input of a Square Corrugated Waveguide (SCW), with no plasma facing moving parts [1]. A high-power testing program was performed on a mockup, in cooperation with FZK, to validate the relative power-loss, steering range, beam quality and polarization. Recently the design has evolved in using a longer SCW, of about 8 to 10 meter, to improve the physics performance to guarantee NTM stabilization in the important ITER scenarios.

The Euratom 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, designed to stabilise the neoclassical tearing modes (NTM) at the q=3/2 and q=2/1 surfaces by inducing off-axis current drive. The launching system enables up to eight mm-wave beams per port plug to be directed from the feeding waveguides through corrugated square waveguides (/spl ap/ 44/spl times/44 mm/sup 2/) 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 HE/sub 11/ mode in 63.5 mm circular corrugated waveguide. (The taper will be replaced by a gyrotron source in high power testing). 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/spl deg/ and 12/spl deg/ through the CVD diamond window into the copper square corrugated waveguide. Antenna patterns from the exit aperture of the square waveguide are scanned using a 3-axis scanner. Low-power measurements (Takahashi et al., 2003) are a preparatory phase for high-power tests, currently prepared in FZK. In the low-power measurements, waveguide performance will be inferred from antenna pattern measurements. Of interest are slight differences in beam parameters (angle, waist etc.) relative to the input beam.

The design work for the ITER ECRH remote-steering upper-port launcher is described. The goal of the system is the stabilization of neoclassical tearing modes (NTMs). The required scanning range in the plasma (more than 20 degrees) is achieved remotely, far away from the hostile plasma environment. This is done behind a diamond window in a separate vacuum environment. Scanning of the beam is done by a mirror that both rotates and translates to steer the beam in the proper way through the window and an isolation valve into a square corrugated waveguide. At the output of this 4.4 m long waveguide, the beam comes out at the same angle and is reflected by a fixed front mirror into the plasma [A.G.A. Verhoeven, et al., The ITER remote-steering ECW upper-port launcher, in: N. Hiromoto (Ed.), The 28th International Conference on Infrared and MM Waves, Otsu, Japan, September 29-October 3, 2003 (JSAP Catalog No. 03123 1)]. (c) 2005 Elsevier B.V. All rights reserved.

A 3D model is of vital interest to the entire development process of the remote steering (RS) electron cyclotron resonance heating (ECRH) top launcher. The aim is to have a digital model that continuously incorporates the latest developments of the overall design. This paper will present the results of the effort put in creating this 3D model and will discuss details about parameterized computer aided design (CAD), which enables the designer to integrate newly acquired insights on a much shorter timescale. Also, the detailed design of the steerable mirror mechanism will be presented. (c) 2005 Elsevier B.V. All rights reserved.

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.

An ECRH (Electron Cyclotron Resonance Heating) system has been designed for JET in the framework of the JET Enhanced-Performance project (JET-EP) under the European Fusion Development Agreement (EFDA). Due to financial constraints it has recently been decided not to implement this project. Nevertheless, the design work conducted from April 2000 to January 2002 shows a number of features that can be relevant in preparation of future ECRH systems, e.g., for ITER.

An electron cyclotron resonance heating (ECRH) system has been designed for JET in the framework of the JET enhanced-performance project (JET-EP) under the European Fusion Development Agreement (EFDA). Due to financial constraints it has recently been decided not to implement this project. Nevertheless, the design work conducted from April 2000 to January 2002 shows a number of features that can be relevant in preparation of future ECRH systems, e.g. for ITER. (C) 2003 Elsevier B.V. All rights reserved.

The Fusion Free-Electron Maser (FFF) is the prototype of a high-power, tunable source of mm-wave radiation, for use on fusion plasma devices. In previous experiments a net output power of 730 kW at 206 GHz was generated in short pulses. The present experiment has been equipped with a system to recover the charge and energy of the spent electron beam. We present experimental results which show output of mm-wave radiation at constant power level during the full pulse length, as well as single-frequency operation; even though the cavity is highly overmoded; the latter is reached by effective suppression of spurious modes by the feedback system.

An intermediate frequency (IF) band digitizing radiometer system in the 100-200 GHz frequency range has been developed for Tokamak diagnostics and control, and other fields of research which require a high flexibility in frequency resolution combined with a large bandwidth and the retrieval of the full wave information of the mm-wave signals under investigation. The system is based on directly digitizing the IF band after down conversion. The enabling technology consists of a fast multi-giga sample analog to digital converter that has recently become available. Field programmable gate arrays (FPGA) are implemented to accomplish versatile real-time data analysis. A prototype system has been developed and tested and its performance has been compared with conventional electron cyclotron emission (ECE) spectrometer systems. On the TEXTOR Tokamak a proof of principle shows that ECE, together with high power injected and scattered radiation, becomes amenable to measurement by this device. In particular, its capability to measure the phase of coherent signals in the spectrum offers important advantages in diagnostics and control. One case developed in detail employs the FPGA in real-time fast Fourier transform (FFT) and additional signal processing. The major benefit of such a FFT-based system is the real-time trade-off that can be made between frequency and time resolution. For ECE diagnostics this corresponds to a flexible spatial resolution in the plasma, with potential application in smart sensing of plasma instabilities such as the neoclassical tearing mode (NTM) and sawtooth instabilities. The flexible resolution would allow for the measurement of the full mode content of plasma instabilities contained within the system bandwidth. (C) 2011 American Institute of Physics.

An electron cyclotron resonance heating (ECRH) system has been designed for JET in the framework of the JET enhanced performance project (JET-EP) under the European fusion development agreement. Due to financial constraints it has been decided not to implement this project. Nevertheless, the design work conducted from April 2000 to January 2002 shows a number of features that can be relevant in preparation of future ECRH systems, e.g. for ITER. The ECRH system was foreseen to comprise six gyrotrons, 1 MW each, in order to deliver 5 MW into the plasma (Verhoeven A.G.A. et al 2001 The ECRH system for JET 26th Int. Conf. on Infrared and Millimeter Waves (Toulouse, 10–14 September 2001) p 83; Verhoeven A.G.A. et al 2003 The 113 GHz ECRH system for JET Proc. 12th Joint Workshop on ECE and ECRH (13–16 May 2002) ed G. Giruzzi (Aix-en-Provence: World Scientific) pp 511–16). The main aim was to enable the control of neo-classical tearing modes. The paper will concentrate on: the power-supply and modulation system, including series IGBT switches, to enable independent control of each gyrotron and an all-solid-state body power supply to stabilize the gyrotron output power and to enable fast modulations up to 10 kHz and a plug-in launcher that is steerable in both toroidal and poloidal angles and able to handle eight separate mm-wave beams. Four steerable launching mirrors were foreseen to handle two mm-wave beams each. Water cooling of all the mirrors was a particularly ITER-relevant feature.

An ECRH (Electron Cyclotron Resonance Heating) system has been designed for JET in the framework of the JET Enhanced-Performance project (JET-EP) under the European Fusion Development Agreement (EFDA). Due to financial constraints it has recently been decided not to implement this project. Nevertheless, the design work conducted from April 2000 to January 2002 shows a number of features that can be relevant in preparation of future ECRH systems, e.g., for ITER.