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A new ECRH launcher mainly aimed at experiments on MHD stabilization has been recently installed in FTU and integrated with a Real Time controller. The new plug-in antenna is capable of injecting two beams over wide angular ranges, ±40 in toroidal and ±15 < θ < 44 in poloidal direction with fast mirror speed. The power is delivered to the launcher with two additional transmission lines (TL) connected to the existing ECRH system with remotely controlled switches. After the optical alignment of the TLs, a campaign of tests on the steering antenna was made with the launcher ex- and in-vessel. A procedure to calibrate the absolute mirror angular position was developed and successfully used. First high power pulses were carried out on plasma target delivering 2 × 0.3 MW EC power for 200 ms. © 2013 CNR-IFP Instituto di Fisica Plasma.

In this paper the architecture of the MHD control system for the Frascati Tokamak Upgrade (FTU) is presented. This control is aimed to stabilize tearing modes (TM) in the plasma by EC heating and current drive. TMs are detected and spatially localized by correlating several ECE channels and Mirnov coil signals; two independent antennas are steered in order to track the TM instability. To guarantee short locking time a fast real-time plasma equilibrium code is used to locate the instability when there is no TM evidence and real-time ray tracing code is used to estimate the location where EC power is deposited. A set of hardware consisting of FPGA and DSP modules on PXI bus for executing the diagnosis and control algorithms is proposed. The overall system layout is introduced together with a description of the data communication among the processors.

In any system designed for electron cyclotron (EC) heating (ECH) and EC current drive in fusion plasmas, the launcher is the matching element between the plasma and the transmission line. Only an appropriate launcher achieves efficient use of the gyrotron power for the many different high-power EC H&CD applications. The frontier is now set at ≈4 MW of launched power at 110 to 140 GHz for ≈10 s, to be further moved to ≈10 MW, 1000 s in the near future. ITER will push the limit to 20 MW, 170 GHz. The workhorse of the antenna system is the front steering setup consisting of a movable mirror, or a mirror array, in front of the hot plasma, which provides for full flexibility in the EC H&CD applications. However, because of the concern associated with cooled and movable parts in a hostile environment, an arrangement with movable mirrors positioned far from the vessel port, and connected to the plasma by imaging waveguides, is being developed as a remote steering backup solution. In a reactor, where flexibility is much less relevant than reliability, the situation could reverse. Techniques for a radial scan of the deposition layer different from front beam steering are discussed in this paper. The ideal goal would be a 100% coupling of the launched EC power, to occur within ≈2% ofthe plasma size and through pipes of size negligible with respect to the vessel, without negative impact on plasma periphery in spite of the high power densities transmitted through the edge.

The development of high power gyrotrons for plasma physics research needs properly matched calorimetric loads able to absorb and measure the power, which nowadays is foreseen to be as high as 2 MW during CW operations. To this end IFP/CNR has developed a family of matched loads useful in the mm-wave frequency band for applications ranging from a few ms to CW in pulse length. The different loads in the family, made of an integrating sphere with a partially reflecting coating on the inner wall, are characterized by having the same absorbing geometry for the incoming beam and a different heat removal system for the specific application. Some important advances have been recently achieved from the point of view of the uniformity of power distribution on the absorbing wall and of the load construction. With the high precision achieved in the coating thickness a better control of the heating power distribution is possible by proper shaping of the local reflectivity, in addition to the shaping of the mirror dispersing the input beam. A more sophisticated model describing the power distribution has been developed, taking into account a variable thickness of the absorbing coating, the proper shape of the spreading mirror, the frequency of the incoming radiation and the shape of the input beam. Lower coating thickness is shown to be preferable, for a given local reflectivity, from the point of view of a lower peak temperature and thermal stress. The paper describes a load with variable coating thickness along the meridian of the sphere, showing a uniform power deposition on the inner walls. The cooling pipe is completely electroformed on the spherical copper shell, ensuring the maintenance of the correct curvature of the inner surface and fast heat conduction from the absorbing coating to the water through the thin copper body. For CW use all heated parts of the load must be cooled and this is achieved with 16 electroformed spiral channels. Both short pulse loads (0.1–1 s) and the CW version at 2 MW, 170 GHz, are described in the paper. High power tests on short-pulse loads have been done using a double frequency gyrotron, 105 GHz/600 kW for 0.5 s and 140 GHz/800 kW for 1 s. Also a method for emulating 2 MW conditions while using 1 MW gyrotron has been applied to test the load to be used for the European 2 MW coaxial cavity gyrotron development programme.

The activities in Europe towards the development of the EC power sources for ITER are centered on the development of a 170GHz, 2MW, CW coaxial cavity gyrotron of collector potential depressed (CPD) type. A gyrotron with a higher unit power than the ITER reference (I MW) would yield a reduction of the installation costs, a more compact launcher design and, if required, an increase of the power delivered through one port. Tests proving the principle were successfully performed on a short-pulse experimental gyrotron delivering up to 2.2 MW in single mode. Following this success, a coordinated and fully consistent programme of development has been launched. The first industrial 2 MW prototype is now at an advanced stage of construction. The associated superconductive magnet producing 6.86 T on the cavity axis is also being procured. Dummy loads suitable for short and CW operation are also part of the development effort. Finally, a new EC test facility, with the features necessary for the testing of the gyrotron up to full power in CW, has been established and includes a fully solid-state power supply system. (c) 2007 Published by Elsevier B.V.

The EU will be providing the largest contribution to the ITER electron cyclotron (EC) heating and current drive (H&CD) system (20 MW, CW at 170 GHz). The contribution includes one third of the H&CD gyrotrons, their associated power supplies and four upper port launcher antennas. In all areas of participation, the EU EC partnership (coordinated by the European Fusion Development Agreement) aims toward advancing the technology, while staying within a specified cost envelope. This is portrayed in the co-axial gyrotron development that offers the potential to double the output power per source (2.0 MW), increasing the delivered power for a fixed number of auxiliary systems. The EU partnerships also attempt to increase performance for the entire EC system, in particular the launching antennas. The proposed front steering launcher design offers greater control of MHD activity than the previous remote steering design and opens up the possibility of an enhanced performance UL. The EC physics requirements are repartitioned between the upper and equatorial launchers for a synergetic balance, which increases the EC physics capabilities while relaxing some of the engineering requirements. (c) 2007 Elsevier B.V. All rights reserved.

To combine powers from a set of gyrotrons and to switch the combined wave beam between a number of output channels, various kinds of multiplexers can be used. Especially, narrow-band frequency diplexers in connection with small frequency-shift keying of gyrotrons can be used to switch the millimeter-wave power between two output channels. This technique can, for example, be used for fast beam steering for synchronous stabilization of rotating neoclassical tearing modes in tokamaks. Beam steering can be performed by a multistage multiplexer, provided that phase-controlled sources are available. In the paper, various concepts for fast directional switches as well as their integration into transmission lines are discussed. Calculations and low-power measurements of prototypes are presented. A resonant diplexer experiment is at present being prepared to be tested at high power in the electron cyclotron resonance heating (ECRH) system for Wendelstein 7X (W7-X). Requirements and techniques for frequency control of the gyrotrons are discussed, and the results of preliminary frequency modulation experiments are shown. Finally, future prospects for the application of diplexers in large ECRH systems are discussed.

The control of rotating magnetic instabilities in a tokamak can be achieved by localized injection of radio-frequency (RF) driven current. Real-time detection of the radial location of the unstable "magnetic islands" and of the RF power deposition has been achieved in a device designed, built, and tested on the FTU tokamak.

A tokamak plasma with internal transport barriers (ITBs) is the best candidate for a steady ITER operation, since the high energy confinement allows working at plasma currents (Ip) lower than the reference scenario. To build and sustain an ITB at the ITER high density (>=1020 m-3) and largely dominant electron (e-) heating is not trivial in most existing tokamaks. FTU can instead meet both requests, thanks to its radiofrequency heating systems, lower hybrid (LH, up to 1.9 MW) and electron cyclotron (EC up to 1.2 MW). By the combined use of them, ITBs are obtained up to peak densities ne0 > 1.3 × 1020 m-3, with central e- temperatures Te0 ? 5.5 keV, and are sustained for as long as the heating pulse is applied (>35 confinement times, ?E). At ne0 ? 0.8 × 1020 m-3 Te0 can be larger than 11 keV. Almost full current drive (CD) and an overall good steadiness is attained within about one ?E, 20 times faster than the ohmic current relaxation time. The ITB extends over a central region with an almost flat or slightly reversed q profile and qmin ? 1.3 that is fully sustained by off-axis lower hybrid current drive. Consequent to this is the beneficial good alignment of the bootstrap current, generated by the ITB large pressure gradients, with the LH driven current. Reflectometry shows a clear change in the turbulence close to the ITB radius, consistent with the reduced e- transport. Ions (i+) are significantly heated via collisions, but thermal equilibrium with electrons cannot be attained since the e--i+ equipartition time is always 4-5 times longer than ?E. No degradation of the overall ion transport, rather a reduction of the i+ heat diffusivity, is observed inside the ITB. The global confinement has been improved up to 1.6 times over the scaling predictions. The ITB radius can be controlled by adjusting the LH power deposition profile that is affected mostly by the q value of the discharge, while the ITB strength can be varied through central EC heating. FTU experiments have shown that ITER-like e-ITBs are achievable.

Millimeter-wave diagnostics at JET, mainly reflectometry, are employing state of art electronics, but are limited in performance by the existing waveguides and antenna system that are inadequate and obsolete. The use of long run waveguides with high losses and non-optimized antennas (in some cases, not pointing to relevant plasma positions) lead to difficult measurement conditions for reflectometry. The new access system presented in this article has been designed to improve the performance of reflectometry measurements and enable the installation of antennas for oblique viewing ECE. These two new antennae will allow the ECE radiation to be collected at different angles with respect to the magnetic field. This set-up, known as oblique ECE [1] , is expected to be extremely useful in improving the interpretation of ECE temperature measurements in all fusion experiments with significant additional heating. For reflectometry, there is an urgent need to improve the edge density measurements as both the lithium beam and Thomson scattering exhibit limitations of resolution at lower densities. The project proposal states that the expected improvement in reflectometry S/N ratio is 30 dB. If realized, this will allow broad band reflectometry, for the measurement of the electron density profile, for the first time in JET.

An increased power handling of rf-beam loads is required by the development of high power gyrotrons for fusion research, as well as the achievement of full capability to work under vacuum conditions, which is imposed by the use of evacuated transmission lines. New techniques were developed at IFP/CNR, leading to two prototypes. One is capable of 1 MW CW operation, while the second one is designed for precise measurements at short-pulses (0.1 s) at higher peak power (2 MW). Preliminary tests at 140 GHz, high power and long pulse (500 kW, 10 s) were performed. (c) 2005 Published by Elsevier B.V.

An important focus of the present research on thermonuclear fusion based on magnetic confinement is the control of rotating magnetic islands of low helical mode numbers m, n that are the nonlinear stage of resistive instabilities developing at magnetic surfaces where q = m/n. These perturbations are often driven by external resonant 'error fields' or mode coupling effects and maintained by local distortions of the current profile. The islands can be reduced in width or completely suppressed by a current driven by electron cyclotron waves accurately located within the island. A requisite for an effective control action is the ability to identify the relevant state variables (radial location, island amplitude, frequency and phase) in 'real time' and vary accordingly the control variables (wave beam power modulation and direction). Traditionally magnetohydrodynamic instabilities are monitored by external magnetic signals that however give no information about the radial location of the perturbation within the plasma. Here we describe the design of a real-time diagnostics/control device using an original algorithm for processing electron cyclotron emission signals to monitor temperature fluctuations and EC wave power deposition.

Significant progress has been made on ASDEX Upgrade during the last two years in the basic understanding of transport, in the extension of the improved H-mode in parameter space and towards an integrated operating scenario and in the development of control methods for major performance limiting instabilities. The important features were the understanding of particle transport and the control of impurity accumulation based on it, the satisfactory operation with predominantly tungsten-clad walls, the improved H-mode operation over density ranges and for temperature ratios covering (non-simultaneously) the ITER requirements on ν*, n/nGW and Te/Ti, the ELM frequency control by pellet injection and the optimization of NTM suppression by DC-ECCD through variation of the launching angle. From these experiments an integrated scenario has emerged which extrapolates to a 50% improvement in n T τ or a 30% reduction of the required current when compared with the ITER base-line assumptions, with moderately peaked electron and controllable high-Z density profiles.

Two experiments on DIII-D have been performed with the purpose of searching for evidence of a critical electron temperature gradient or gradient scale length. Both experiments employed off-axis electron cyclotron heating to vary the local value of inverse temperature scale length 1/L-Te, = -del T-e/T-e while holding the total heating power and thus edge temperatures constant. The incremental or heat pulse thermal diffusivity was monitored as a means of identifying the existence of a threshold or critical value of 1/L-Te, k(crit), where a large increase in heat pulse diffusivity may occur while going from below to above kcrit. No evidence of a critical value was found, but the existence of one cannot be ruled out by the experimental results. The results can be used to place an upper bound on k(crit). If k(crit) exists, the experimental results indicate kcrit < 3.8 m(-1) at rho = 0.45 and k(crit) < 2.5 m(-1) at rho = 0. 29 corresponding to temperature scale lengths relative to the plasma minor radius of LTe/a > 0.43 and 0.65, respectively. Over most of the plasma radius (rho >= 0.3) the temperature profile in the L-mode discharges studied exceeds these upper bounds on k(crit). A transport model based on a critical gradient is consistent with experimental observations but is not unique. Models other than one based on the existence of a critical gradient are also consistent with experimental observations.

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.

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 rdep, 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 rdep and risland 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 rdep, 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.

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 PΩ 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° < < 12°.

This chapter reports the main physics results obtained with three radio-frequency-injection systems. The frequency of 8 GHz for the lower hybrid (LH) current drive (CD) (LHCD) system was chosen to explore CD at high density: full CD has been demonstrated for central densities up to 1.4 x 10{sup 20} m{sup -3} at 0.5 MA with an applied power up to 2.0 MW. The Frascati Tokamak Upgrade (FTU) database shows CD efficiencies from 0.1 to 0.3 x 10{sup 20} AW{sup -1} m{sup -2}. In combined experiments with electron cyclotron (EC) waves (140 GHz, up to 1.2 MW), a suprathermal absorption by the fast electron tail generated by LHCD has been observed in both downshifted and upshifted interaction regimes, with the resulting electron cyclotron current drive (ECCD) ranging from 20 to 100 kA, depending on experimental conditions. With pure EC resonance heating, the narrowness of the radial power deposition profile has been exploited, resulting in strong local electron heating. Results in high-density regimes are also presented. The third system (433 MHz, 0.5 MW) is the first to test ion Bernstein wave (IBW) coupling with a waveguide antenna. The experiment operates at high frequency, avoiding the occurrence of nonlinear phenomena at the edge.more » Improved confinement regimes resulting in a central peaking of the pressure profiles have been achieved with P{sub IBW} up to 0.4 MW. Modeling and experimental results are summarized.« less

Transport studies are presented in this chapter. Global scaling studies have been performed using several transport codes. Ohmic plasmas are found to follow the ITER97 L-mode scaling. Transport coefficients are discussed for improved confinement scenarios achieved in the Frascati Tokamak Upgrade (FTU): the repetitive pellet enhanced plasma mode, showing neoclassical confinement with H-factors up to 1.6, and the electron internal transport barriers (ITBs) with large transport barriers and H-factors up to 1.3. Heat transport models have been tested using electron cyclotron resonance heating (ECRH), steady or modulated, as a probe. The electron temperature stiffness observed in the main bulk of steady FTU plasmas can be interpreted both with a critical gradient transport model and with a model based on the existence of canonical profiles. ECRH has also been used to benefit from the improved confinement generally associated with low or negative magnetic shear, and large electron temperatures have been achieved in these conditions. Profile resiliency is observed so that heat transport is not consistent with a constant thermal diffusivity. Experimental optimization is discussed together with the analysis of transport coefficients. Thorough discussions of impurity transport are given, both for intrinsic and injected (from laser blow-off) impurities. Code simulation and experimentalmore » data are compared for a series of FTU experiments focusing on the improved confinement modes (pellets and ITBs). A moderate inward pinch velocity is generally required to reproduce the data.« less

High-frequency wave systems with high-power density launching capability have been the preferred choice to heat the Frascati Tokamak Upgrade (FTU) because of physics arguments (electron heating at ...

The main physics results achieved in the recent years in the Frascati Tokamak Upgrade (FTU) are reviewed. The main focus of research has been the development of performance plasmas at high densities (up to 4 x 10{sup 20} m{sup -3}), high magnetic field (up to 8 T) and plasma current (up to 1.6 MA), that are therefore in a domain of relevance for burning physics experiments such as ITER. The main tools consist in the development of plasma conditioning techniques and the use of various electron heating and current drive systems. Improved confinement regimes have been developed, including (a) the production of steady electron internal transport barriers at high density and electron temperature (up to central electron temperature of 11 keV at a central density of 0.9 x 10{sup 20} m{sup 3}), (b) the production of repetitive pellet enhanced plasma modes with deep pellet deposition leading to a substantial increase of the neutron yield (and a record FTU value of the fusion product n{sub i}T{sub i}{tau}{sub E} up to 0.8 x 10{sup 20} m{sup -3} keV.s), and (c) the production of radiation improved modes at high magnetic field. Main results on the supporting physics program will also be given inmore » the domain of plasma wave physics (lower hybrid current drive, electron cyclotron resonance frequency, ion Bernstein waves), heat and impurities transport, and magnetohydrodynamic studies.« less

Perturbative and steady-state heat transport of FTU tokamak in current ramp-up discharges are investigated by means of modulated electron cyclotron heating (MECH). Perturbative and steady-state transport experiments are coherent with an electron heat transport which switches from low to high values when electron temperature gradient length reaches a threshold value 1/L_Tc. The threshold value 1/L_Tc is shown to be proportional to the ratio s/q. The experimental findings are compared to predictions of an empirical model based on the assumption of a threshold gradient length, L_Tc (1/L_T = |grad Te/Te|), in the electron temperature Te below which electron thermal diffusivity, chi_e, switches from low to high values. Plasma responses to steady state and MECH are modelled assuming the electron diffusivity as chi_PB = chi_0 + alpha Te^(3/2) (1/L_T - 1/L_Tc)^(1/2); here Te^(3/2) reflects the gyro-Bohm assumption, chi_0 represents the heat transport for 1/L_T < 1/LT_c and the term (1/L_T - 1/L_Tc)^(1/2), which sets in for 1/L_T > 1/L_Tc , mimics an extra transport possibly due to electron temperature gradient (ETG) modes. In agreement with ETG threshold 1/L_Tc is shown to be correlated with the magnetic shear s.

For the electron cyclotron resonance heating (ECRH) system on the frascati tokamak upgrade (FTU), composed of four 0.5 MW gyrotrons pulsed for 0.5 s, two different calorimetric loads were designed to measure the mm-wave power at 140 GHz propagating either as a gaussian beam or as an HE(11) mode carried by an oversized corrugated waveguide. The first one, designed for low reflectivity and I MW power handling capability, is now routinely used for gyrotron testing, conditioning and optimization on the generator side of the transmission line at 500 kW, 0.5 s. The basic design of the load, with a substantial upgrading of the cooling system for adequate heat removal, has full CW operation capability. The other one, whose cylindrical shape allows positioning as a waveguide replacement, is occasionally used when tests on the transmitted power along the 40 m of waveguide are needed. The loads are described together with low power measurements of reflected radiation. (C) 2001 Elsevier Science BN. All rights reserved.

Barrier windows are critical components in electron cyclotron plasma heating/current drive (ECRH/ECCD) systems for fusion plasma experiments in which high frequency and high power microwave beams are used as additional heating or plasma profiles control. Even though material characteristics and thickness are chosen in order to reduce the absorbed power to a minimum, the window temperature is of major concern in high power, long pulse or CW operation. An infrared diagnostic device has been implemented on one of the boron nitride windows of the 140 GHz, 0.5 s, 0.5 MW per line ECRH system of the FTU tokamak. The evolution of the temperature of the barrier window can be monitored by acquisition of several IR emission profiles during the gyrotron pulse. Moreover, information on the beam pattern quality can be inferred. (C) 2001 Elsevier Science BN. All rights reserved.

The quasi-optical beam launching antenna for the Electron Cyclotron Resonance Heating (ECRH) experiment at 140 GHz, 1.6 MW to the plasma, on the Frascati Tokamak Upgrade (FTU) has been successfully exploited in experimental operations. It provides four beams 400 kW each, with independent poloidal and toroidal steering capability and a maximum power density of 60 kW/cm2 at the plasma edge. The beam radius in the plasma is ≈20 mm, allowing a very high localisation of the absorbed power. The main characteristics of the antenna are: • Four movable launching mirrors under vacuum, far from the plasma edge (no movable parts near the plasma). • Oblique toroidal injection capability at fixed angles obtained with reflections on two fixed stainless steel plates, gold plated, inserted at the sides of the port. • Vacuum gate separating the main vacuum from the appendix containing the two matching mirrors and the barrier window (for safety and maintenance). • All movements are transferred outside the vacuum chamber through bellows and linear displacement. Actuators and encoders are in air. • Full capability to adjust small movements of the machine during cool-down and warm-up, preserving the alignment. Pattern measurements at low power before installation were performed to characterise the effects of the real system on the ideal shape and the polarisation of the launched beams, with particular attention to diffraction effects. Final control of the shape of the beams and the alignment of the mirrors in the system is performed after installation, under vacuum and with the tokamak at liquid nitrogen temperature, with the aid of a retractable probe and using a thermopile as the RF detector.

Plasma absorption at the fundamental EC resonance of e.m. waves at 140 GHz, 0.8 MW, Ordinary mode, provides in the FTU tokamak a localized (minimum deltar similar or equal to2.5 cm) and intense (up to similar or equal to 100 MW/m(3)) heat source for the electrons. It is shown that at high density (similar or equal to 10(20) m gamma - (-3)) ions are heated (DeltaT(i)/T-i = 0.25) by collisional e - i energy transfer and assuming ion thermal diffusivity about three times larger (at most) than predictions in order to account ;for the observed T-i. Off-center localization (r(abs)/a similar or equal to 0.2) of the ECRH heat source changes the electron temperature (and current density) profile in a way to fully stabilize sawteeth. On-axis absorption accelerates the sawtooth crash repetition rate if P-ecrh less than or equal to P-oh, but with P-ecrh > P-oh the sawtooth period is increased. A critical shear model for the sawtooth crash repetition rate is shown to be in agreement with its experimental dependence on resonance position and heating power. In discharges characterized by a flat temperature profile in the ohmic phase, ECRH may de-stabilize tearing modes of different order, with a strong degradation of core energy confinement. (C) 2001 Published by Elsevier Science B.V.

A general overview of the Frascati tokamak upgrade (FTU) machine for the period 1997–1999 is given. The status of the major systems and the relevant upgrades are discussed together with some data on the operating experience, the main scientific results and the near term programme.

Electron temperatures in excess of 8 keV have been obtained by electron-cyclotron-resonance heating on FTU plasmas at peak densities up to $8\ifmmode\times\else\texttimes\fi{}{10}^{19}\mathrm{m}{}^{\ensuremath{-}3}$. The magnetic shear in the plasma core is low or negative, and the electron heat diffusivity remains at, or below, the Ohmic level $(0.2\mathrm{m}{}^{2}/\mathrm{s})$, in spite of the very large heating power density $(10--20\mathrm{MW}/\mathrm{m}{}^{3})$ which produces extremely high temperature gradients (up to 120 keV/m). The ion heat transport remains at the neoclassical level.

We have produced a prototype broadband, low-sidelobe conical corrugated feed horn suitable for measurements of the Cosmic Microwave Blackground (CMB) radiation in the frequency band 120–150 GHz. The antenna is a first prototype for the Low Frequency Instrument array in ESA's PLANCK mission, a space project dedicated to CMB anisotropy mesurements in the 30–900 GHz range. We describe the fabrication method, based on silver electro-formation, and present the two-dimensional antenna beam pattern measured at 140 GHz with a milimeter-wave automated scalar test range. The beam has good symmetry in the E and H planes with a far sidelobe level approaching –60 dB at angles ∼ 80°. An upper limit to the return loss was measured to be –21 dB.

It is well known that fusion plasma operations can be limited in standard scenarios at high β by resistive instabilities, called neoclassical tearing modes (NTMs), which degrade the plasma confinement leading to a loss of plasma energy, and in some cases to disruption. The avoidance of their onset is a very important issue that has been largely investigated in many tokamaks. In particular, it has been shown that these modes, shaped as magnetic islands, can be triggered by finite seed perturbations associated with long sawtooth crashes. The control of the sawtooth periods is then a key step in the physics of plasma confinement in fusion devices: the shortening of these periods can reduce any triggered large seed island below the NTMs' growth threshold allowing maximum β values and high plasma performances to be achieved. A powerful tool for sawtooth control is the use of high localized electron cyclotron heating (ECH) and current drive (ECCD), capable of modifying the plasma current density and effecting the sawtooth period. Modulated ECH and ECCD have been used as triggers of sawtooth crashes to test conditions for an a priori constant sawtooth period. In the FTU high magnetic field compact tokamak (R0 = 0.93 m, a = 0.3 m, B0 = 4-8 T) similar experiments have been performed with an ECRH system of four gyrotrons operating at 140 GHz and delivering 0.5 MW each. Repetitive pulses of EC power from 1 to 2 gyrotrons up to 0.8 MW for 500 ms have been used to investigate the sensitivity of sawtooth periods during long and short EC switching on and off phases in view of a real time EC control system soon to be working in FTU. A new type of locking of the sawtooth periods to the EC modulation has been observed for deposition inside the q = 1 radius for EC on phase smaller than the ohmic period. In this paper, sawtooth period shortening and locking by ECH and co-ECCD inside the q = 1 radius is investigated and reproduced by numeric simulations. © 2014 EURATOM.

The design of the ITER electron cyclotron launchers recently reached the preliminary design level - the last major milestone before design finalization. The ITER ECH system contains 24 installed gyrotrons providing a maximum ECH injected power of 20 MW through transmission lines towards the tokamak. There are two EC launcher types both using a front steering mirror; one equatorial launcher (EL) for plasma heating and four upper launchers (UL) for plasma mode stabilization (neoclassical tearing modes and the sawtooth instability). A wide steering angle range of the ULs allows focusing of the beam on magnetic islands which are expected on the rational magnetic flux surfaces q=1 (sawtooth instability), q=3/2 and q=2 (NTMs). In this paper the preliminary design of the ITER ECH UL is presented, including the optical system and the structural components. Highlights of the design include the torus CVD-diamond windows, the frictionless, front steering mechanism and the plasma facing blanket shield module (BSM). Numerical simulations as well as prototype tests are used to verify the design (C) 2013 Karlsruhe Institute of Technology (KIT). Published by Elsevier B.V. All rights reserved.

In the electron cyclotron frequency range the handling of high power is critical. In some cases an unpredictable amount of stray radiation can reach some components or accumulate in localized regions, with risk of damages caused by thermal overloads, and any uncontrolled reflection represents a danger for the sources. A possibility to mitigate the problem consists in covering some regions exposed to radiation with absorbers. Enhanced absorption of stray radiation lowers requirements on active protection systems in microwave diagnostics. The released heat can be extracted by dedicated cooling systems. The chromium oxide (Cr2O3), largely tested at IFP-CNR, has been routinely used as internal coating for matched loads. The performances of a variable thickness coating has been tested at high power at Karlsruhe Institute of Technology (KIT), with a 140 GHz gyrotron of the W7-X ECRH system and an averaged power density absorbed at the coating surface higher than 1 MW/m2 for 3 min. Also boron carbide (B4C) has been tested at low power and patented as a millimeter-wave absorber. In the paper, the results of some tests performed on these coatings are given, together with some simulations of the absorption capability based on low power measurements on samples. Finally, some calculations are presented for a coating obtained combining together Cr2O3 and B4C. © 2013 CNR-IFP Instituto di Fisica Plasma.

A real-time data acquisition and elaboration system is being implemented to control the new ECH launcher recently installed at FTU (Frascati Tokamak Upgrade). The system is aimed at controlling different kinds of magnetohydrodynamic instabilities, in particular the deleterious 312 and 211 (neoclassical) tearing modes, (N)TM, and the saw teeth period in order to prevent the seeding of NTMs. The complete system is presented here together with preliminary offline and real-time tests. (C) 2001 Elsevier Science. All rights reserved. (C) 2013 Published by Elsevier B.V.

A peculiarity of the quasi-optical propagation of a Gaussian beam of electromagnetic waves is that it requires a treatment taking account, step by step, of all of its rays together. The dispersion relation from which such a ray- tracing may be deduced is in fact characterized not only by the medium where the beam is launched but also by the intrinsic beam structure. The behaviour of three-dimensional Gaussian beams in the electron-cyclotron frequency range is considered, with particular attention paid to their propagation in toroidal plasmas of fusion interest (with ITER-like parameters and various magnetic equilibrium configurations), to the generality of their launching conditions and to their power deposition profiles around resonant layers.

The Electron Cyclotron (EC) fast launcher for real time control experiments has been installed on FTU and characterized to be fully integrated in a real-time MHD control module under development. The launcher scheme is based on a two module sys- tem, symmetric with respect to the equatorial plane of FTU, with a front steering concept and the launched beams are real-time controllable both in poloidal and toroidal directions. Specific design parameters, defined by the FTU MHD dynamics (typically island size and q profile changes), are the beam dimensions with zooming capabilities, the steering range and mirror speed with the most demanding requirement on poloidal speed of the Steering Mirror (SM) = 1 in 10 ms. A set of tests has been done to verify the system perfor- mance. High power tests of the launcher have been done on a 500 kA, 0.5 10 20 m -3 and 5.3 T plasma with 2 300 kW of EC power delivered to the plasma. The steering mechanism was tested under the automatic control system and showed a dynamic response in line with the requirements. Results of these tests will be presented in the paper.

The automatic controller developed for MHD instability control on FTU via the real time (RT) EC launcher [1, 2] is based on the a-priori estimate of the instabilities location and on the fast elaboration of the stream of diagnostic data. A fast data acquisition system, based on existing standard FTU diagnostics, collects thermal and magnetic signals by a 12-channels ECE polychromator [3] and by a set of 22 Mirnov coils respectively. Moreover, the EC time waveform from directional couplers along the transmission line is acquired as well. This set of signals is processed in a timescale significantly shorter than the typical time step of the controller (1ms). RT elaboration algorithms aim at supplying the controller [4] with reliable information about the existence and the spatial location of the instability and about the actual ECRH deposition layer in plasma. The main algorithms blocks are ECE-ECE, ECE-Mirnov and ECE-ECRH cross-correlation, and SVD (Singular Value Decomposition) of Mirnov signals. The capabilities of effective detection, false positive resilience, and mode discrimination will be discussed through the application to actual plasma data.

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New FTU ohmic discharges with a liquid lithium limiter at I P = 0.7–0.75 MA, B T = 7 T and n e0 ⩾ 5 × 1020 m−3 confirm the spontaneous transition to an enhanced confinement regime, 1.3–1.4 times ITER-97-L, when the density peaking factor is above a threshold value of 1.7–1.8. The improved confinement derives from a reduction of electron thermal conductivity (χe) as density increases, while ion thermal conductivity (χi) remains close to neoclassical values. Linear microstability reveals the importance of lithium in triggering a turbulent inward flux for electrons and deuterium by changing the growth rates and phase of the ion-driven turbulence, while lithium flux is always directed outwards. A particle diffusion coefficient, D ∼ 0.07 m2 s−1, and an inward pinch velocity, V ∼ 0.27 m s−1, in qualitative agreement with Bohm–gyro-Bohm predictions are inferred in pellet fuelled lithized discharges. Radio frequency heated plasmas benefit from cleaner plasmas with edge optimized conditions. Lower hybrid waves penetration and current drive effects are clearly demonstrated at and above ITER densities thanks to a good control of edge parameters obtained by plasma operations with the external poloidal limiter, lithized walls and pellet fuelling. The electron cyclotron (EC) heating system is extensively exploited in FTU for contributing to ITER-relevant issues such as MHD control: sawtooth crash is actively controlled and density limit disruptions are avoided by central and off-axis deposition of 0.3 MW of EC power at 140 GHz. Fourier analysis shows that the density drop and the temperature rise, stimulated by modulated EC power in low collisionality plasmas are synchronous, implying that the heating method is the common cause of both the electron heating and the density drop. Perpendicularly injected electron cyclotron resonance heating is demonstrated to be more efficient than the obliquely injected one, reducing the minimum electric field required at breakdown by a factor of 3. Theoretical activity further develops the model to interpret high-frequency fishbones on FTU and other experiments as well as to characterize beta-induced Alfvén eigenmodes induced by magnetic islands in ohmic discharges. The theoretical framework of the general fishbone-like dispersion relation is used for implementing an extended version of the HMGC hybrid MHD gyrokinetic code. The upgraded version of HMGC will be able to handle fully compressible non-linear gyrokinetic equations and 3D MHD.

The new Electron Cyclotron Resonance Heating and Current Drive launcher for real-time control experiments on FTU consists of two antennas with front fast-steering mirrors, aiming to test new strategies for MHD stabilization and plasma heating. The description and experimental identification of the mirror dynamics, for the design of both an optimized position controller and a model predictive protection system, are the main objectives of this paper. Each mirror is steered by a couple of AC brushless motors for toroidal and poloidal movements and each motor is controlled by a drive with embedded PI speed and torque controllers. A position controller, based on plasma feedback, is realized externally with a set of hardware and software also described in this paper. Several tests have been carried out to evaluate the system dynamic performance compared with the target specifications and to identify a state-space model of the mechanical system to be used for a model predictive protection, whose aim is to avoid that the mirror goes out of the workspace boundaries. (C) 2011 Elsevier B.V. All rights reserved.

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.

This Paper reviews the design and functionality of the 24MW 170GHz electron cyclotron heating and current drive system being planned for the ITER Tokamak. The sub-systems (power supplies, gyrotrons, transmission lines and launcher antennas) are described based on present day technologies, while on-going R&D provides component and sub-system testing with the possibility of increasing the reliability of the overall EC system. Modifications to the steering ranges of the launching antennas are under investigation that can improve the functional capabilities of the EC system without increasing cost and relaxing the engineering constraints.

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In this work we present the results obtained applying Singular Value Decomposition (SVD) on a Mirnov coil array in FTU discharges where experiments on MHD stabilization or disruption avoidance via ECRH were performed. In these shots the mode analysis has been consolidated off line by means of FFT and SXR‐tomography. Although the Mirnov setting was not necessarily optimised, results show that the algorithm based on SVD is able to detect the mode with a precision equal or better than the FFT, while acting on a smaller time interval. The short execution time required, even in the present preliminary form, suggests that this analysis can be a suitable tool to be implemented in a real time control chain.

Generation of a swing electron cyclotron current drive (swing ECCD), i.e. driving alternated, symmetric, positive or negative local ECCD, during a single discharge and at constant total input EC power, was performed at the Tokamak a Configuration Variable (TCV). The electron temperature is observed to be modulated inside the deposition radius, implying modulation of the electron transport properties. The modulation of ECCD is the only actuator for the observed modifications in the electron transport properties. These exhibit inverted behaviors depending on the deposition location of the co- and counter-ECCD. At more on-axis depositions, swing ECCD results in a higher electron temperature during the co- ECCD phase, whereas at more off-axis depositions, the electron temperature is higher during the counter-ECCD phase. Transport modeling of these discharges shows that the local electron tranport behavior depends on the value of the modulated magnetic shear. The results are transport model independent, confirming the robustness of the magnetic shear modeling, and indicating that the main contribution is due to the ECCD. Moreover, the results are consistent with predictions from gyrokinetic simulations, that the local electron confinement is proportional to the magnetic shear at low shear and inversely at high shear values, s greater than or similar to 1.

The challenge of developing the conceptual design of the ECH Upper Launcher system for MHD control in the ITER plasmas has been tackled by team of European Associations together with the European Domestic Agency ("F4E"). The launcher system has to meet the following requirements: (a) a mm-wave system extending from the interface to the transmission line up to the target absorption zone in the plasma and performing as an intelligent antenna; (b) a structural system integrating the mm-wave system and ensuring sufficient thermal and nuclear shielding; (C) port Plug remote handling and testing capability ensuring high Port plug system availability. The paper describes the reference launcher design. The mm-wave system is composed of waveguide and quasi-optical sections with a front steering system. An automated feedback control system is developed as a concept based on an assimilation procedure between predicted and diagnosed absorption location. The structural system consists of the blanket shield module, the port plug frame, and the internal shield for appropriate neutron shielding towards the launcher back-end. The specific advantages of a double walled structure are discussed with respect to adequate baking, to rigidity towards launcher deflection under plasma-generated loads and to removal of thermal loads, including nuclear ones. Basic Studies of remote handling (RH) to validate design development are initiated using a virtual reality simulation backed by experimental validation, for which a launcher handling test facility (LHT) is set LIP as a full scale experimental site allowing furthermore thermohydraulic studies with ITER blanket water parameters. (C) 2008 Elsevier B.V. All rights reserved.

The development of electron cyclotron resonance heating (ECRH)-electron cyclotron current drive (ECCD) as a tool for suppression of plasma instabilities requires that the millimeter-wave beams used for testing magnetohydrodynamic (MHD) stabilization schemes for ITER be able to follow magnetic island position in real time. In the FTU tokamak, the design of a new ECRH fast-steerable launcher will enable a fast-controlled trolled deposition at a precise poloidal location and the inclusion of the mirror motion in a feedback loop aimed at MHD stabilization. Two of the four existing transmission lines will be switched to the new launcher located in a different equatorial port. It will launch two independent beams with radius in the plasma changeable between 17 and 28 mm, in order to control the deposited power density. Real-time control of the poloidal steering requires high acceleration, speed, and positioning precision of the last mirror. Additionally, oblique toroidal injection at precise angles will allow current profile shaping through controlled ECCD and heating of overdense plasmas (n(e) > 2.4 X 10(20) m(-3)) using electron Bernstein waves. For optimal 04 conversion, the required toroidal angle, estimated with dedicated beam-tracing calculations, is close to +/-38.5 deg, near the upper limit in the toroidal steering angle. The launch requirements and their impact on the launcher design phase are presented in the paper.

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.