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The ITER ECRH&CD system is composed by 24 gyrotrons at 170 GHz that will deliver 20 MW at the plasma. Up to 6.7 MW will be injected in the empty vacuum vessel at the beginning of each plasma discharge to provide the gas breakdown. In that phase and when the plasma absorption is non ideal, a certain level of EC non-absorbed power, usually addressed as stray radiation, will be present. The EC stray radiation interaction with ITER first wall and diagnostics has been described in preliminary works. A more refined assessment is here described, following update of the diffuse stray radiation model and update of the launchers optics parameters. The optical design of the EC equatorial launcher has been entirely redesigned to optimize the power deposition and minimize interaction with the launcher structures. The updated parameters for the 24 launched beams are now available and have been used to estimate the interaction of the beams to be used for the breakdown phase with the tokamak structures. The preliminary stray radiation model described every opening of the tokamak as a “black” hole, that is a perfect power sink. Refining this crude description, using for the openings a “grey” hole model, provide a better agreement with benchmarks from other alternative models. Examples of stray radiation estimates performed for various ITER structures, systems and diagnostics are discussed.

First plasma operation in ITER will start after completing the assembly of the tokamak vessel and the installation of the main sub-systems, but prior to the installation of the blanket modules and the divertor cassettes. Utilization of temporary limiters and divertor replacement structures will provide a poloidal and toroidal reference side position to the plasma edge to protect the vacuum vessel and other components already installed during operations. An additional set of mirrors is required to reflect the power injected from the upper launcher towards the plasma resonance for EC-assisted breakdown of the plasma up to a beam dump needed to trap and absorb the power of the beams after crossing the plasma in order to reduce the stray radiation escaping back into the vacuum chamber down to less than 10% of the total power. The quasi-optical system has been designed, with shape and size of the mirrors compliant with the requirements provided by ITER for their installation, realization and plasma performances, resulting in two standard focussing mirrors and one grating mirror. The beam dump consists of a box with five metal plates, the first providing a spreading of the high incident power and the others coated with absorbing material with thickness distribution studied to gradually reduce the power during the multiple reflections inside the box, avoiding damages to the coating itself. This work focuses on the validation of the quasi-optical design of the mirrors and the assessment of the dump performances, based on a multi-bounces model developed ad-hoc for this purpose. The study includes a tolerance analysis for the beam dump to include the effect of uncertainties in the thickness of the absorbing coating and misalignments of the mirrors, to verify the performances of the dump also when operating in different conditions with respect to the nominal ones.

The ITER first plasma operations will occur with no full blanket and divertor components installed. Machine protection is required and First Plasma Protection Components (FPPC) have been designed to shelter the vacuum vessel and other components from the plasma itself and from high power Electron Cyclotron Resonance Heating (ECRH) foreseen for first plasma breakdown. ECRH protection components will be installed to protect in-vessel structures from direct and stray radiation as EC beams will be used for plasma breakdown. This paper is focused on the design of these components including dedicated mirrors to shape and redirect the beams to the EC resonance location in the magnetic field null region and then into a beam dump located in an equatorial port, where exceeding EC radiation will be trapped and dumped. Two mirrors and one grating mirror have been designed to provide the required shaping and directions for the beams coming from the upper launcher towards EC resonance and dump. The beam dump consists of five large plates affixed in an equatorial port. Guidelines that drove the design of the quasi-optical system, characteristics of the mirrors, resulting launched beams and concept developed for the beam dump will be here described.

The ITER EC system includes both an equatorial port launcher and upper port launchers as RF power injection devices, The waveguide switch in the transmission line (TL) is used to select the operating launcher. Its operation is required even during plasma operation, primarily for the mid-pulse switch operation. Changing the waveguide switch requires that the gyrotron stop RF power during the switch operation since the direction is changed by mechanical movement of the mirror position which takes a few seconds. Since the ITER EC system is based on a multi-subsystem concept, each subsystem has its own subsystem control unit (SCU). The EC main controller supervises all subsystem controllers. Hence, cooperative operation requires the sharing of the information of both RF power status and switch status between the gyrotron SCU and TL SCU via the main controller. Since the design of the inter-subsystem control scheme is a key issue for ITER EC system control, its evaluation is required. At QST, the gyrotron and ITER-relevant TL test stand were utilized for demonstration of mid-pulse switch operation. For this purpose, SCUs for each subsystem and the main controller were developed using the ITER-relevant control system. The operation of the mechanical switch during gyrotron pulse was demonstrated. During the 150 s operation of the high power gyrotron at 400 kW level, the waveguide switch in the TL was operated to change the direction of RF power. The time duration for the switch operation with inter-subsystem control scheme took 1.5 s in total. The synchronizing of RF power suspend and resume with switch motion has therefore been realized, and RF power direction control during the gyrotron operation was successfully demonstrated.

The ITER Plasma Control System (PCS) approaching the second phase of development is now more deeply investigating alternative solutions for the various controls aimed at operations up to 15MA with low auxiliary heating in L-mode. The control functions in the PCS are strictly linked to the performance of the ITER actuators and diagnostics. The capabilities of those systems need to be carefully validated against the control needs. System and performance requirements shall be consistent with the control schemas and where limitations or restrictions are identified, it is necessary to provide alternative solutions. To guarantee this consistency, a set of interface documents is being prepared. Those interfaces for each of the plant systems that can impact the PCS activities detail the requirements specifically needed for control and report also the functional relationship between the two systems. The PCS has also to consider areas not actively part of plasma control that might affect or limit PCS operations (i.e. forces in the superconducting coils). This paper reports the main outcome from the interfaces definition. The actuator boundaries and plant systems constraints impacting the PCS design will be presented. For the sensors the challenge is the derivation of real-time measurement requirements in relation to the separate diagnostic requirements and their respective interface with the PCS. The complex organization of data integration with the PCS will be discussed.

A set of components will be installed during ITER fist plasma operation to protect the vacuum vessel and other in-vessel auxiliary systems from the plasma and from the stray radiation injected at the Electron Cyclotron harmonics to generate breakdown and sustain burn-through. This paper focuses on the quasi-optical design of the system of three mirrors redirecting the microwave beams coming from the Electron Cyclotron Resonance Heating (ECRH) upper launcher to the plasma resonance after proper shaping. In particular, the system consists of two shaped mirrors and one grating mirror. The non-absorbed EC power is then intercepted and absorbed into a beam dump located in one equatorial port.

The ITER Electron Cyclotron (EC) plant is a complex system, essential for plasma operation. The system is being designed to supply up to 20 MW of power at 170 GHz; it consists of 24 RF sources (or Gyrotrons) connected by switchable transmission lines to four upper and one equatorial launcher. The complexity of the EC plant requires a Plant Controller, which provides the functional and operational interface with CODAC and the Plasma Control System and coordinates the various Subsystem Control Units, i.e. the local controllers of power supplies, Gyrotrons, transmission lines and launchers. A conceptual design of the Electron Cyclotron Control System (ECCS) was developed, starting from the collection of the user requirements, which have then been organized as a set of operational scenarios exploiting the EC system. The design consists in a thorough functional analysis, including also protection functions, and in the development of a conceptual I&C architecture. The main aim of the work was to identify the physics requirements and to translate them into control system requirements, in order to define the interfaces within the components of the ECCS. The definition of these interfaces is urgent because some of the subsystems are already in an advanced design phase. The present paper describes both the methodology used and the resulting design. (C) 2015 Elsevier B.V. All rights reserved.

High-power microwaves generated by gyrotrons will be extensively used in ITER for a variety of purposes such as assisting plasma breakdown, plasma heating, current drive, tearing mode suppression and as a probing beam for the Collective Thomson Scattering diagnostic. In a number of these schemes absorption of the microwaves by the plasma will not be full and in some cases there could be no absorption at all. This may result in a directed beam with a high microwave power flux or – depending on location and plasma conditions – an approximately isotropic microwave power field. The contribution of electron cyclotron emission to these power densities is briefly discussed. Exposure to in-vessel components leads to absorption by metals and ceramics. In this paper microwave power densities are estimated and, following a brief review of absorption, thermal loads on in-vessel components are assessed. The paper is concluded by a discussion of the current approach to control such loads.

The external walls of the Tokamak building, made of thick concrete, provide the nuclear shielding for operators working in adjacent buildings and for the environment. There are a series of openings to these external walls, devoted to ducts or pipes for ventilation, waveguides and transmission lines for heating systems and diagnostics, cooling pipes, cable trays or busbars. The shielding properties of the wall shall be preserved by adequate design of the openings in order not to affect the radiological zoning in adjacent areas. For some of them, shielding properties of the wall are not affected because the size of the network is quite small or the source is far from the opening. But for most of the openings, specific features shall be considered. Even if the approach is the same and the ways to shield can be standardized, specific analysis is requested in any case because the constraints are different.

The electron cyclotron (EC) heating and current drive (H&CD) system developed for the ITER is made of 12 sets of high-voltage power supplies feeding 24 gyrotrons connected through 24 transmission lines (TL), to five launchers, four located in upper ports and one at the equatorial level. Nearly all procurements are in-kind, following general ITER philosophy, and will come from Europe, India, Japan, Russia and the USA. The full system is designed to couple to the plasma 20 MW among the 24 MW generated power, at the frequency of 170 GHz, for various physics applications such as plasma start-up, central H&CD and magnetohydrodynamic (MHD) activity control. The design takes present day technology and extends toward high-power continuous operation, which represents a large step forward as compared to the present state of the art. The ITER EC system will be a stepping stone to future EC systems for DEMO and beyond. The development of the EC system is facing significant challenges, which includes not only an advanced microwave system but also compliance with stringent requirements associated with nuclear safety as ITER became the first fusion device licensed as basic nuclear installations as of 9 November 2012. Since the conceptual design of the EC system was established in 2007, the EC system has progressed to a preliminary design stage in 2012 and is now moving forward toward a final design.

The RF heating and current drive (H&CD) systems to be installed for the ITER fusion machine are the electron cyclotron (EC), ion cyclotron (IC) and, although not in the first phase of the project, lower hybrid (LH). These systems require high voltage, high current power supplies (HVPS) in CW operation. These HVPS should deliver around 50 MW electrical power to each of the RF H&CD systems with stringent requirements in terms of accuracy, voltage ripple, response time, turn off time and fault energy. The PSM (Pulse Step Modulation) technology has demonstrated over the past 20 years its ability to fulfill these requirements in many industrial facilities and other fusion reactors and has therefore been chosen as reference design for the IC and EC HVPS systems. This paper describes the technical specifications, including interfaces, the resulting constraints on the design, the conceptual design proposed for ITER EC and IC HVPS systems and the current status.

The transmission line (TL) subsystem associated with the ITER electron cyclotron heating and current drive system has reached the conceptual design maturity. At this stage the responsibility of finalizing the design has been transferred from the ITER Organization to the U.S. Domestic Agency. The purpose of the TL is to transmit the microwaves generated by the 170-GHz gyrotrons installed in the radio-frequency building to the launchers located in one equatorial and four upper tokamak ports. Each TL consists of evacuated HE11 wave guides, direct-current breaks, power monitors, mitre bends, polarizers, switches, loads, and pumping sections and will have a typical length that ranges from 100 to 160 m. Overall transmission efficiency could be as high as 92% depending on the specific path between a given gyrotron and launcher. All components are required to be 2-MW compatible, and their layout and organization have been optimized for simplifying the maintenance accessibility and monitoring the primary tritium barrier integrity. Two different TL layouts are at the moment under study, to accommodate the two alternative options for the European sources: four 2-MW units or eight 1-MW units. In this paper the actual design is presented and the technical requirements are discussed.

An electron cyclotron system is one of the four auxiliary plasma heating systems to be installed on the ITER tokamak. The ITER EC system consists of 24 gyrotrons with associated 12 high voltage power supplies, a set of evacuated transmission lines and two types of launchers. The whole system is designed to inject 20 MW of microwave power at 170 GHz into the plasma. The primary functions of the system include plasma start-up, central heating and current drive, and magneto-hydrodynamic instabilities control. The design takes present day technology and extends towards high power CW operation, which represents a large step forward as compared to the present state of the art. The ITER EC system will be a stepping stone to future EC systems for DEMO and beyond. The EC system is faced with significant challenges, which not only includes an advanced microwave system for plasma heating and current drive applications but also has to comply with stringent requirements associated with nuclear safety as ITER became the first fusion device licensed as basic nuclear installations as of 9 November 2012. Since conceptual design of the EC system established in 2007, the EC system has progressed to a preliminary design stage in 2012, and is now moving forward towards a final design. The majority of the subsystems have completed the detailed design and now advancing towards the final design completion. (C) 2014 Elsevier B.V. All rights reserved.

A 24 MW Electron Cyclotron (EC) system operating at 170 GHz and 3600 s pulse length is to be installed on ITER. The EC plant shall deliver 20 MW of this power to the plasma for Heating and Current Drive (H&CD) applications. The EC system is designed for plasma initiation, central heating, current drive, current profile tailoring, and Magneto-hydrodynamic control (in particular, sawteeth and Neo-classical Tearing Mode) in the flat-top phase of the plasma. A preliminary design review was performed in 2012, which identified a need for extended application of the EC system to the plasma ramp-up, flattop, and ramp down phases of ITER plasma pulse. The various functionalities are prioritized based on those applications, which can be uniquely addressed with the EC system in contrast to other H&CD systems. An initial attempt has been developed at prioritizing the allocated H&CD applications for the three scenarios envisioned: ELMy H-mode (15 MA), Hybrid (similar to 12 MA), and Advanced (similar to 9 MA) scenarios. This leads to the finalization of the design requirements for the EC sub-systems.

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

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

Compact directional couplers-polarimeters were developed for the circular corrugated waveguide transmission lines of the joint IFP-ENEA Electron Cyclotron Resonance Heating experiment at 140 GHz, 2 MW on the Frascati Tokamak Upgrade in Frascati.A linear array of cutoff holes was drilled in the mirrors of the quasi-optical miter bends. The radiated pattern preserves the symmetry and polarization of the waveguide mode in the plane of the array. The direction of propagation is preserved too.Two pairs of standard gain horns with detectors are placed in the plane of the array along the propagation axes of incident and reflected radiation to detect both linear components. The whole assembly is enclosed in a shielded anechoic box of suitable geometry.

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.

Summary form only given. A 24MW CW Electron Cyclotron Heating and Current Drive (EC HC present day systems are operating in relatively short pulses (≤10s) and installed power levels of ≤4.5MW.

We describe a false case of infection caused by Dicrocoelium dendriticum, a cosmopolite trematode that can infect human bile ducts but tends to live in cattle or other grazing mammals. Our aim is to stress the relevance of adequate diagnostic methods and of exact medical history in order to detect any possible clinical case.

The ITER electron cyclotron (EC) system having capability of up to 26 MW generated power at 170 GHz is being procured by 5 domestic agencies via 10 procurement arrangements. This implies diverse types of equipment and complex interface management. It also places a challenge on control system architecture to entertain the constraints of procurement slicing and meeting the overall functional requirement. The envisioned architecture is to use the local control units (supplied with each procurement) and a supervisory plant controller (by ITER). This offers a reliable control configuration for such delicate and complex EC plant system. The control system is envisioned to monitor the whole plant and perform automated tasks that are today performed via direct human intervention. For example, the automated gyrotron conditioning and active control of the EC plant to respond to requests from the plasma control system (PCS). This later aspect requires rapid shut down of the gyrotrons and power supplies, deviation of the actuators to direct the power from an equatorial to upper launcher and then restart of the power generation for rapid stabilization of the magneto hydrodynamic (MHD) instabilities that occur in high performance plasma operation. The plant controller will be designed for optimized performance with the PCS and the feedback control system used to actively control the power (with modulation capability up to 5 kHz) and launching direction for MHD stabilization. (C) 2011 ITER Organization. Published by Elsevier B.V. All rights reserved.

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An engineering study has been poformed on the ITER electron cyclotron transmission lines with the aim of optimizing its conceptual design. The support types and optimum spacing, cooling, vacuum, seismic, and gravitational effects were reviewed. For the vacuum system it was shown that two pumps per line, with a capacity of 50 l/s, are sufficient. It was explained that the temperature variation inside the building is the predominant factor that influences the thermal expansion of the lines. The support strategy is one of minimizing the number of constraints. Variation in support interspacing reduces the degree of harmonic disturbances. The section of transmission line inside the ITER port cell was identified as critical with regards to occurrence of deformation and stresses. Potential solutions are described. The use of seismic breaks is discussed in light of the differences in foundation and structure of the ITER tokamak building and assembly hall. It is proposed that this interface be studied in more detail, after more data is available on the behavior of these buildings. The geometry of individual supports should be simple, with the fewest possible adjustments. The supports are designed to allow small movements of the waveguide to compensate for the thermal expansion or contraction. The transmission line system can be made for optimum alignment during nominal operating temperatures by prestressing during installation.

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.

A collaborative effort between European research Associations and Thales Electron Devices (TED) has been launched by the European Fusion Development Agreement (EFDA) in 2003, aiming at the development of an industrial 170 GHz/2 MW/CW coaxial cavity gyrotron. The first prototype is expected to reach 2 MW/1s and is presently being tested in Lausanne at a dedicated test facility. The test facility has been designed to be flexible: allowing the possible commissioning of tubes with different characteristics, as well the tests of a version of the ITER upper launcher antenna at full performances. The test facility has been commissioned during the first test phases on the gyrotron which include: high voltage stand-off, coaxial insert alignment and cathode conditioning without depressed collector power supply at a slightly reduced electron beam power of 85 kV/78 A/2 ms. The short pulse (

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

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Experiments on the effects of simultaneous injection of LH and ECH power have been carried out in FTU, with up to PLH=1 MW and PECH=0.75 MW. In the absence of the cold ECH resonance in the plasma, the ECH power is absorbed on the suprathermal electron tail sustained by LHCD, as predicted theoretically, and causes the central electron temperature Te0 to increase up to 1.2 keV. With the cold resonance at the center, in a sawteeth and MHD free plasma, almost fully sustained by LHCD, a quasi-stationary electron transport barrier is built-up in a region r/a

Gaussian beam steering (2 beams, ≈400 kW each) and toroidal field tuning are used to control ECRH localization for MHD stabilization and energy transport studies on FTU tokamak. It is shown that by locating the absorption of one beam inside the m=2 island, and of a second one close to the s.t., inversion radius, both instabilities can be suppressed, with a significant impact on energy confinement. MHD suppression allows the analysis of the underlying energy transport. It is shown that the electron temperature and pressure profiles are stiff when ECRH is localized off-axis, and that a finite amount of heat pumping against the temperature gradient must be accounted for to explain steady-state profiles at high density (ne0,line≈1020 m−3). The resistance of the thermal response to change with the heating source term can be analyzed in terms of two models implying profile stiffness: the effect of a critical temperature gradient, possibly related to ETG modes, and the tendency of a tokamak plasma to stick to pr...

The EC system at 140 Ghz, 2 MW, being implemented on FTU tokamak for performing electron heating, profile control and current drive, is composed by four gyrotrons with 0.5 s pulse length capability, and four hybrid mirror/waveguide transmission lines. The launching system is capable of poloidal/toroidal beam steering through a set of tiltable in-vacuum mirrors, coupling to the O-mode at the fundamental resonance, with full control of the e.m. field polarization. Experiments have been made with two gyrotrons, by launching 0.8 MW to the plasma. In the high density regime (local n{sub e} from {approx_equal}0.8 up to 2 10{sup 20} m{sup -3}) the e-i energy transfer is significant, and ion heating is observed through the enhancement of the neutron emission in Deuterium plasmas. The highest electron heating is observed when on-axis ECRH is performed in the current ramp-up, sawtooth-free phase. In steady-state conditions, saw-tooth period is increased up to stabilization by off-axis ECRH. Localized ECCD is also performed to assist in the shaping of the current density profile.

Localized ECRH/ECCD has been performed on FTU with the specific aim of shaping the temperature/current density profile for controlling MHD activity. In particular, m=1 mode dynamics and stability are studied during the heating phase, when the steady-state, sawtoothing conditions of the ohmic plasma are modified by a strong off-axis electron heating. Depending on the position of the absorbing layer and on the density of the ohmic target plasma, sawteeth are temporarily suppressed, still in the presence of a slowly growing m=1 oscillation, or a complete quenching of any m=1 activity can be achieved. When direct current, driven by EC waves launched at +/-10 degrees from the perpendicular in the up-shifted scheme, is added to the pure beating effect, different shapes and periods of relaxation phenomena in the plasma core are observed. Although low values of I-ECCD/I-p are expected because of rather high values of Z(eff) during heating, (I-ECCD/I(p)approximate to 0.04 for P-EC =800KW, Z(eff)=5 and I-p=360kA), different local changes in the magnetic shear are thought to be responsible of these differences. Time-dependent calculations of current profiles, including transport and diffusion effects, and an analysis of the onset and stabilization of sawtooth oscillations based on the time evolution of sl(the magnetic shear value at q=1) compared with that of a suitable critical value are performed.

Most of the interest in ECRH experiments is linked to the high localization of EC waves absorption in well known portions of the plasma volume. In order to take full advantage of this capability a reliable code has been developed for beam tracing and absorption calculations. The code is particularly important for oblique (poloidal and toroidal) injection, when the absorbing layer is not simply dependent on the position of the EC resonance only. An experimental estimate of the local heating power density is given by the jump in the time derivative of the local electron pressure at the switching ON of the gyrotron power. The evolution of the temperature profile increase (from ECE polychromator) during the nearly adiabatic phase is also considered for ECRH profile reconstruction. An indirect estimate of optical thickness and of the overall absorption coefficient is given by the measure of the residual e.m. power at the tokamak walls. Beam tracing code predictions of the power deposition profile are compared with experimental estimates. The impact of the finite spatial resolution of the temperature diagnostic on profile reconstruction is also discussed.

The 140GHz, 2MW, 0.5s ECRH system on FTU tokamak integrates closed waveguide transmission lines (≈30 m) with quasi optical systems at both ends for efficient coupling from the 4 gyrotrons to the 4 waveguides and from these to the plasma through a single access port. Poloidal and toroidal control of the beam’s launching angles and polarization is performed without movable components close to the plasma. Most of the components of each generation and transmission system were designed to operate at a power level higher than 0.5 MW, and a possible up-grade to a full 1 MW, 0.5 s capability is discussed.

The electron cyclotron emission (ECE) diagnostic on FTU tokamak is routinely performed with a Michelson interferometer with spectral range extending up to 1300 GHz. The diagnostic allowed accurate electron temperature measurements during the recent 140 GHz electron cyclotron resonance heating (ECRH) experiments on FTU. Very accurate measurements have been performed on a wide range of electron temperatures and profile peaking. The ECE measurements have been compared with Thomson scattering and with observations of x-ray spectra from highly stripped molybdenum ions. The suprathermal emission in these conditions has been studied.

The European Union is working toward providing 2-MW, coaxial-cavity, continuous-wave (cw) 170-GHz gyrotrons for ITER. Their design is based on results,from an experimental preprototype tube having a pulse length of several milliseconds, in operation at Forschungszentrum Karlsruhe (FZK) for several years now. The first industrial prototype tube was designed for cw operation but, in a first phase, aimed at a pulse length of 1 s at the European Gyrotron Test Facility in Lausanne, Switzerland, as part of a phased testing/development program. (1 s, 60 s, cw). The first experimental results of the operation of this prototype gyrotron are reported here. The microwave generation was characterized at very short pulse length (

The main magnetohydrodynamic (MHD) activities affecting the Frascati Tokamak Upgrade (FTU) high-field plasmas with limiter configuration are sawtooth relaxations and tearing modes.The period of sawtooth relaxations can be increased in FTU both by electron heating and by pellet particle deposition near the sawtooth inversion radius; both methods lead to full stabilization in proper conditions. The sawtooth period can be shortened as well by central heating.The influence of localized electron cyclotron resonance heating (ECRH) on the stability of m = 2 tearing modes has been studied in FTU by means of radial and power scans. Heating between the plasma center and the island location increases the island size, while heating at the island location produces mode stabilization if ECRH power exceeds a threshold value. These sawtooth and tearing mode studies show that some control of both phenomena can be achieved.Double-tearing modes in the form of regular, sawtooth-like relaxations have been observed in discharges with reversed magnetic shear. The development of these instabilities is particularly interesting in FTU as it happens in the absence of injected momentum.Long-lived m = 1 island structures are frequently observed following pellet deposition near the inversion radius; particle accumulation around the O-point enhances diagnostic sensitivity, thus allowingmore » fine studies of island dynamics.MHD spectroscopy has revealed the existence of coherent waves at frequencies well above the drift-tearing range in thermal plasmas. In addition, broadband turbulence has been observed both in ohmic and in radio-frequency-heated plasmas. The amplitude of turbulent fluctuations increases with heating power and is anticorrelated with the neutron yield.« less