We present the design and operation of a suite of Gas Puff Imaging (GPI) diagnostic systems installed on the Tokamak à Configuration Variable (TCV) for the study of turbulence in the plasma edge and Scrape-Off-Layer (SOL). These systems provide the unique ability to simultaneously collect poloidal 2D images of plasma dynamics at the outboard midplane, around the X-point, in both the High-Field Side (HFS) and Low-Field Side (LFS) SOL, and in the divertor region. We describe and characterize an innovative control system for deuterium and helium gas injection, which is becoming the default standard for the other gas injections at TCV. Extensive pre-design studies and the different detection systems are presented, including an array of avalanche photodiodes and a high-speed CMOS camera. First results with spatial and time resolutions of up to ≈2 mm and 0.5 µs, respectively, are described, and future upgrades of the GPI diagnostics for TCV are discussed.

The ITER ECRH system consists of 24 gyrotrons with up to 24 MW installed millimeter wave heating power at 170 GHz, power supplies, control system, transmission lines, one Equatorial and the four Upper Launchers. With its high frequency and small beam focus the ECRH has the unique capability of driving locally current. While the Equatorial Launcher mainly acts for central heating and current profile shaping, the Upper Launchers aim on suppressing MHD instabilities, especially neoclassical tearing modes (NTM) triggering plasma disruptions. The Upper Launchers inject millimeter waves through a quasi-optical section consisting of three fixed and the front steering mirror set. The eight overlapping beams have focal points optimized for suppression of the q = 3/2 and q = 2/1 NTMs. Several project change requests required the redesign of the Upper Launchers and the connected ex-vessel system. This redesign includes a new boundary geometry of the launchers as well as a newly designed cooling system for the Blanket Shield Module (BSM), a modified flange of the BSM to the structural main frame and a refined optical design. Additionally shield blocks with integrated in-vessel waveguides were added and the closure plate with waveguide and supply line feedthroughs was adapted. Further changes, not all caused by PCRs, include newly designed ex-vessel waveguide components with a reduced aperture and redesigned ultra low-loss CVD diamond windows. Finally several components originally foreseen as off-the-shelf components have become part of the design scope. The new launcher design status is presented with selected results on numerical design validation.

The recently completed MST2 upgrade to the Thomson scattering (TS) system on TCV (Tokamak à Configuration Variable) at the Swiss Plasma Center aims to provide an enhanced spatial and spectral resolution while maintaining the high level of diagnostic flexibility for the study of TCV plasmas. The MST2 (Medium Sized Tokamak) is a work program within the Eurofusion ITER physics department, aimed at exploiting Europe's medium sized tokamak programs for a better understanding of ITER physics. This upgrade to the TCV Thomson scattering system involved the installation of 40 new compact 5-channel spectrometers and modifications to the diagnostics fiber optic design. The complete redesign of the fiber optic backplane incorporates fewer larger diameter fibers, allowing for a higher resolution in both the core and edge of TCV plasmas along the laser line, with a slight decrease in the signal to noise ratio of Thomson measurements. The 40 new spectrometers added to the system are designed to cover the full range of temperatures expected in TCV, able to measure electron temperatures (Te) with high precision between (6 eV and 20 keV) . The design of these compact spectrometers stems originally from the design utilized in the MAST (Mega Amp Spherical Tokamak) TS system located in Oxfordshire, United Kingdom. This design was implemented on TCV with an overall layout of optical fibers and spectrometers to achieve an overall increase in the spatial resolution, specifically a resolution of approximately 1% of the minor radius within the plasma pedestal region. These spectrometers also enhance the diagnostic spectral resolution, especially within the plasma edge, due to the low Te measurement capabilities. These additional spectrometers allow for a much greater diagnostic flexibility, allowing for quality full Thomson profiles in 75% of TCV plasma configurations.

The ITER Electron Cyclotron Heating Upper Launcher (ECHUL) will be used to drive current locally inside magnetic islands located at the q = 2 (or smaller) rational surfaces in order to stabilize neoclassical tearing modes (NTMs). Each of the fours antennas consists of eight beam lines, totaling 12.8 m that are designed for the transmission of up to 1.5 MW of mm-wave power at 170 GHz, with at least 90% of the power in the main HE11 mode. Each transmission line consists of a Z shaped set of straight corrugated, aluminum alloy waveguides connected by miter bends with a nominal inner diameter of 50 mm. The FCS system also encompasses intrinsic components such as: closure plate, mm-waveguide taper, isolation valve, diamond window and an EU-US interface waveguide. All FCS components are connected via a dedicated flange coupling, with a bolted connection, comprising of two concentric metallic seals in order to satisfy the SIC-1 requirements. The coupling shall be capable of resisting the applied external loads and displacements, including thermal cycles due to ohmic losses in the transmission line, and also adhere to the matting of different materials (EN AW-6082, CuCrZr, SS 316L) while maintaining ultra-high vacuum tightness (10(-8) mbar). The present study will use a dedicated experimental apparatus for the characterization of the mechanical and vacuum properties of a single metallic seal under compression, with applied heat flow. In order to ensure reproducibility of the components mechanical characteristics, randomly chosen metallic seals of each diameter have been tested from two independent manufactures (A and B). (C) 2017 Elsevier B.V. All rights reserved.

•Presentation of two variants of a novel high-power waveguide cooling design.•Fluid dynamic simulation at nominal ECH power transmission.•Transient analysis of a typical ITER thermal cycle.•Manufacturing and assembly concept for the waveguide cooling channels.

The four electron cyclotron (EC) upper port antennas (or "upper launchers" -UL) will be used to drive current locally inside magnetic islands located at the q= 2 (or smaller) rational surfaces in order to stabilize neoclassical tearing modes (NTMs), as well as heat inside of p of about 0.4. Each antenna consists of eight beam lines that are designed for the transmission of 1.5 MW of mm-wave power at 170 GHz. The First Confinement System (FCS) is formed by the ex-vessel mm-wave waveguide components, for which SIC-1 classification requirements apply.The beam lines in the FCS comprise a Z shaped set of straight corrugated waveguides with a nominal diameter of 50 mm connected by miter bends. This system is subjected to imposed displacements coming mainly from the thermal expansion of the vacuum vessel, seismic events and/or plasma disruption events. In absence of suitable SIC-1 waveguide bellows, the FCS waveguides must provide the necessary mechanical functional compliance. This has required the development of a dedicated, flange type coupling system with double metallic seals, capable of resisting the generated external loads while maintaining vacuum tightness and alignment. This paper presents the results of the design, analysis and pre-qualification experimental work done on the waveguides and the integrated SIC-1 compliant coupling system. (C) 2016 Elsevier B.V. All rights reserved.

The electron cyclotron (EC) system of ITER for the initial configuration is designed to provide 20MW of RF power into the plasma during 3600s and a duty cycle of up to 25% for heating and (co and counter) non-inductive current drive, also used to control the MHD plasma instabilities. The EC system is being procured by 5 domestic agencies plus the ITER Organization (IO). F4E has the largest fraction of the EC procurements, which includes 8 high voltage power supplies (HVPS), 6 gyrotrons, the ex-vessel waveguides (includes isolation valves and diamond windows) for all launchers, 4 upper launchers and the main control system. F4E is working with IO to improve the overall design of the EC system by integrating consolidated technological advances, simplifying the interfaces, and doing global engineering analysis and assessments of EC heating and current drive physics and technology capabilities. Examples are the optimization of the HVPS and gyrotron requirements and performance relative to power modulation for MHD control, common qualification programs for diamond window procurements, assessment of the EC grounding system, and the optimization of the launcher steering angles for improved EC access. Here we provide an update on the status of Europe’s contribution to the ITER EC system, and a summary of the global activities underway by F4E in collaboration with IO for the optimization of the subsystems.

The Electron Cyclotron (EC) system for the ITER tokamak is designed to inject >= 20 MW RF power into the plasma for Heating and Current Drive (H&CD) applications. The EC system consists of up to 26 gyrotrons (between 1 and 2 MW each), the associated power supplies, 24 transmission lines and 5 launchers. The EC system has a diverse range of applications including central heating and current drive, current profile tailoring and control of plasma magneto-hydrodynamic (MUD) instabilities such as the sawtooth and neoclassical tearing modes (NTMs). This diverse range of applications requires the launchers to be capable of depositing the EC power across nearly the entire plasma cross section. This is achieved by two types of antennas: an equatorial port launcher (capable of injecting up to 20 MW from the plasma axis to mid-radius) and four upper port launchers providing access from inside of mid radius to near the plasma edge. The equatorial launcher design is optimized for central heating, current drive and profile tailoring, while the upper launcher should provide a very focused and peaked current density profile to control the plasma instabilities.The overall EC system has been modified during the past 3 years taking into account the issues identified in the ITER design review from 2007 and 2008 as well as integrating new technologies. This paper will review the principal objectives of the EC system, modifications made during the past 2 years and how the design is compliant with the principal objectives. (C) 2011 ITER Organization. Published by Elsevier B.V. All rights reserved.

A 24MW CW Electron Cyclotron Heating and Current Drive (EC H&CD) system operating at 170GHz is to be installed for the ITER tokamak. The EC system will represent a large step forward in the use of microwave systems for plasma heating for fusion applications; present day systems are operating in relatively short pulses (<= 10s) and installed power levels of <= 4.5MW. The magnitude of the ITER system necessitates a worldwide collaboration. This is also reflected in the EC system that is comprised of the power supplies, sources, transmission line and launchers. A partnership between Europe, India, Japan, Russia, United States and the ITER organization is formed to collaborate on design and R&D activities leading to the procurement, installation, commissioning and operation of this system.

The ITER Heating and Current Drive Upper Launcher (H & CD EC UL) uses a pneumomechanical steering-mirror assembly (SMA) to steer the RF beams for their deposition in the appropriate location in the plasma to control magnetohydrodynamic activity (neoclassical tearing modes (NTMs) and sawtooth oscillations). For NTM stabilization, the mirror rotation needs to be controlled to an accuracy that is better than 0.1°. A 10°·s-1 mirror steering speed is also required. To assess the performance of the two SMA prototypes that have been manufactured, a test stand that reproduces the expected pneumatic configuration of the UL has been built. So far, only the first SMA prototype has been tested, and tests on the second prototype are foreseen in the 2009-2010 period. The steering angle of the mirror will be deduced from the pressure applied to the mechanism since there is no in situ angle measurement at present. An ¿off-the-shelf¿ commercial servo valve with a proportional-integral-derivative controller has been used to control the pressure with good results for the switching cycle. These tests show that a more advanced controller will be required to attain the desired accuracy and speed for the modulation cycles.

The ITER ECRH upper launcher (UL) uses a steering mechanism assembly (SMA) to steer the beams in the appropriate location for controlling magnetohydrodynamic (MILD) activity in the plasma (NTMs and sawtooth oscillations).Switching and modulation rotation cycles with an accuracy of 0.1 degrees are requirements for Neoclassical Tearing Modes (NTMs) stabilisation in ITER.Two SMA prototypes have been manufactured. A test stand that reproduces the expected pneumatic configuration of the UL has been built to perform control tests on both SMA prototypes. The first SMA prototype has been tested, and tests on the second prototype are foreseen on the second half of the present year.An "off-the-shelf" commercial servo-valve with a HD controller has been used to control the pressure with good results for the switching cycle. These tests have shown that a more sophisticated control strategy will be needed to attain the desired accuracy for the modulation cycles, since the present controller presents a permanent tracking error for ramp commands.

The purpose of the ITER electron cyclotron resonance heating (ECRH) upper launcher (UL), or antennae will be to provide localised current drive by accurately directing mm-wave beams up to 2MW, out of the four allocated upper port plugs, at chosen rational magnetic flux surfaces in order to stabilise neoclassical tearing modes (NTMs). This paper will present an overview of the UL, with emphasis on the mm-wave components. The mm-wave layout includes corrugated waveguide sections and a quasi-optical path with both focusing mirrors and plane steering mirrors. One of the essential components of the UL is the Steering Mechanism Assembly (SMA), providing variable poloidal injection angles fulfilling high deposition accuracy requirements at the plasma location. The Actuator principle and rotor bearings are frictionless and backlash free, avoiding tribological difficulties such as stickslip and seizure. The underlying working principle is the use of mechanically compliant structures. Validation and proof testing of the steering principle is achieved with an uncooled first prototype demonstrator. A second prototype is currently being manufactured, comprising the functionalities needed for the ITER compatible system such as water cooling and high power mm-wave compatibility. In order to perform the fatigue tests of the actuator bellows, a test facility has been built, under ITER-like vacuum and temperature working conditions. Results of the cyclic fatigue tests are compared to the various manufacturer standards and codes, combining stress and strain controlled material fatigue properties.

Four of the 16 ITER upper port plugs will be devoted to electron cyclotron resonance heating (ECRH) in order to control the magneto-hydrodynamic (MHD) instabilities.In order to achieve the stabilisation of the neoclassical tearing modes (NTM) and sawtooth oscillation, a deposition of a very localized, narrow and peaked current density profile over a broad poloidal steering range at the ITER rational surfaces q=1 to q=3/2 and q=2 is required [1]. The quasi-optical configuration consists of eight mm-wave beams entering each of the four upper launchers (UL) through waveguides into the vacuum vessel. Each beam is then directed to the plasma using a serial arrangement of four mirrors, the first set of 2 (mirrors M1 and M2) zigzag the beam line to avoid stray radiation, while the beam-waist locations and beam-shaping properties in the vacuum vessel are defined by the last two mirrors, a static focusing mirror (M3) and a flat poloidally-steerable mirror (M4). These last mirrors (M3 and M4) reflect a group of four beams coming of four individual M1/M2. The plane mirror M4 is steered using a backlash and friction free mechanism that is pneumatically actuated using helium gas. A first prototype of this has been manufactured as a proof of principle and tested for controller design purposes. A second prototype, ITER compatible in terms of the materials used and cooling circuits, will be subject to similar tests throughout the last half of 2009 period.The UL mirrors (static and steerable) absorb heat generated essentially by three sources: the ohmic loss of the RF beam reflected at the mirror surfaces (which is mirror-temperature dependant) and by nuclear (volumetric heating) and thermal radiation (surface heating) coming from the plasma. While the average heat load is calculated to be 2MW beam compatible with reasonable engineering limits, three more important constraints conditioned the actual mirror design, the peak ohmic heat load (the electromagnetic forces generated in ver- tical disruption events (VDE), and the ITER cooling water requirements. This paper provides an overview of the different upper port-plug mirror designs and stresses several different prototype manufacturing methods for the steering mirror, outlining lessons learned.

The main objective of the ITER ECRH upper launcher (UL) is to control magnetohydrodynamic activity, in particular neoclassical tearing modes (NTMs), by driving several MW of EC current near the q=1, 3/2, 2 flux surfaces, where NTMs are expected to occur.

The front steering (FS) launcher is one of two concepts that have been considered for the ITER electron cyclotron heating upper launcher by the European Union. During the development of a detailed conceptual design, the team involved with the FS launcher project listed all of the critical issues associated with installing an FS launcher in the ITER upper port, and then work was concentrated on providing a solution to each of the critical design issues. A similar procedure was Performed for the alternative launcher option (remote steering launcher). These actions helped the ITER International Team evaluate the two systems and then choose a final optimum launcher. This evaluation occurred at the end of 2005, with both systems having equivalent reliability, but the FS offered significant enhancement in the physics performance. These differences led ITER-IT to select the FS launcher as the reference design. The goal of this paper is to provide a generalized review of the critical design issues and their solutions as they pertain to the FS launcher. In addition, the overall design and performance of the FS launcher is given along with a brief description of an extended performance launcher design that relaxes the engineering constraints, while increasing the physics capabilities.