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To reach fusion conditions and control the plasma configuration in ITER, the next step in tokamak fusion research, two neutral beam injectors (NBIs) will supply 16.5 MW each, by neutralizing accelerated negative hydrogen or deuterium ions. The requirements of ITER NBIs (40A/1 MeV D- ions for ≤1 h, 46A/870 keV H- ions for ≤1000 s) have never been simultaneously attained. So in the Neutral Beam Test Facility (NBTF, Consorzio RFX, Italy) the operation of the full-scale ITER NBI prototype (MITICA) will be tested and optimised up to full performances, focussing on accelerator (including voltage holding), beam optics, neutralisation, residual ion removal. The NBTF includes also the full-scale prototype of the ITER NBI source with 100 keV particle energy (SPIDER), for early investigation of: negative ion production and extraction, source uniformity, negative ion current density and beam optics. This paper will describe the main results of the first two years of SPIDER operation, devoted to characterizing plasma and beam parameters, including investigation of RF-plasma coupling efficiency and magnetic filter field effectiveness in reducing co-extracted electrons. SPIDER is progressing towards the first caesium injection, which aims at increasing the negative ion density. A major shutdown, planned for 2021, to solve the issues identified during the operation and to carry out programmed modifications, will be outlined. The installation of each MITICA power supply and auxiliary system is completed; in-vessel mechanical components are under procurement by Fusion for Energy (F4E). Integration, commissioning and test of the power supplies, procured by F4E and QST, as the Japanese Domestic Agency (JADA), will be presented. In particular, 1.0MV insulating tests were carried out step-by-step and successfully completed. In 2020 integrated tests of the power supplies on the accelerator dummy load started, including the assessment of their resilience to accelerator grid breakdowns using a short-circuit device located in vacuum. The aggressive programme, to validate the NBI design at NBTF and to meet ITER schedule (requiring NBIs in operation in 2032), will be outlined. Unfortunately, in 2020 the coronavirus disease infection affected the NBTF activities. A solution to proceed with integrated power tests despite the coronavirus is presented.

Heating neutral beam (HNB) injectors, necessary to achieve burning conditions and to control plasma instabilities in ITER, are characterized by such demanding parameters that a neutral beam test facility (NBTF) dedicated to their development and optimization is being realized in Padua (Italy) with direct contributions from the Italian government (through Consorzio RFX as the host entity) and the ITER international organization (with kind contributions from the ITER domestic agencies of Europe, Japan and India) and technical and scientific support from various European laboratories and universities. The NBTF hosts two experiments: SPIDER, devoted to ion source optimization for the required source performance, and MITICA, the full-size prototype of the ITER HNB, with an ion source identical to SPIDER. This paper gives an overview of the progress towards NBTF realization, with particular emphasis on issues discovered during this phase of activity and on solutions adopted to minimize the impact on the schedule and maintain the goals of the facilities. The realization of MITICA is well advanced; operation is expected to start in 2023 due to the long procurement time of the in-vessel mechanical components. The beam source power supplies, operating at 1 MV, are in an advanced phase of realization; all high-voltage components have been installed and the complex insulation test phase began in 2018. At the same time, construction and installation of SPIDER plant systems was successfully completed with their integration into the facility. The mechanical components of the SPIDER ion source were installed inside the vessel and connected to the plants. Integrated commissioning with the control, protection and safety systems ended positively and the first experimental phase has begun. The first results of the SPIDER experiment, with data from operational diagnostics, and the plans for the 1 MV insulation tests on the MITICA high-voltage components are presented.

The neutral beam system for ITER consists of two heating and current drive neutral beam injectors (HNB) and a diagnostic neutral beam (DNB) injector. The proposed physical plant layout allows a possible third HNB injector to be installed later. For the correct operation of the beam, the ion source and the ion path until it is neutralized must operate under a very low magnetic field environment. To prevent the stray ITER field from penetrating inside those mentioned critical areas, a magnetic field reduction system (MFRS) will envelop the beam vessels and the high voltage transmission lines to ion source. This system comprises the passive magnetic shield (PMS), a box like assembly of thick low carbon steel plates, and the Active Correction and Compensation Coils (ACCC), a set of coils carrying a current which depends on the tokamak stray field. This paper describes the magnetic model and analysis results presented at the PMS and ACCC preliminary design review held in ITER organization in April 2013. The paper focuses on the magnetic model description and on the description of the analysis results. The iterative process for obtaining optimized currents in the coils is presented. The set of coils currents chosen among the many possible solutions, the magnetic field results in the interest regions and the fulfillment of the magnetic field requirements are described.

The neutral beam system for ITER consists of two heating and current drive injectors plus a diagnostic neutral beam injector. The proposed physical plant layout allows for a possible third heating injector to be installed later. For correct operation of the beam source, and to avoid deflections of the charged fraction of the beam, the magnetic field along the beam path must be very low. To minimize the stray ITER field in critical areas (ion source, acceleration grids, neutralizer, residual ion dump), a Magnetic Field Reduction System will envelop the beam vessels and the high voltage transmission lines to ion source. This whole system comprises the Passive Magnetic Shield, a set of thick steel plates, and the Active Correction and Compensation Coils, a set of coils carrying currents which depend on the tokamak stray field. This paper describes the status of the coil design, terminals and support structures, as well as a description of the calculations carried out. Most coils are suitable for removal from their final position to be replaced in case of a fault. Conclusions of the chosen design highlight the strategy for the system feasibility.

The ITER Passive Magnetic Shield (PMS) main function is to protect the Neutral Beam Injector (NBI) from the external magnetic field coming from the tokamak, and to shield the NB cell from the radiation coming from all activated components. The shielding from the external magnetic field is performed in association with the Active Compensation Cooled Correction Coils (ACCC). The Bushing and Transmission Line (TL) PMS also provides structural support for HV bushing, allowing its maintenance and providing air sealing function between NBI cell and High Voltage deck room. The paper summarizes the structural analyses performed in order to evaluate the mechanical behaviour of the HNB PMS under operation combined with seismic event. The RCC-MR Code is used to validate the design, assuming creep is negligible, since the structure is expected to be at room temperature. P-type damage is assessed.

The neutral beam (NB) system for ITER is composed of two heating neutral beam injectors (HNBs) and a diagnostic neutral beam injector (DNB). A third HNB can be installed as a future up-grade. This paper will present the design development of the components between the injectors and the tokamak; the so-called ‘front end components’: the drift duct consists of the NB bellows and the drift duct liner, the vacuum vessel pressure suppression system box (VVPSS box), the absolute valve, and the fast shutter. These components represent the key links between the ITER tokamak and the vessels of the NB injectors. The design of these components is demanding due to the different loads that these components will have to stand. The paper will describe the different design solutions which have to be implemented regarding the primary vacuum confinement, the power handling capability and the remote maintenance operations. The sizes of the components are determined by the large cross section of the neutral beam. The power handling capability is driven by the anticipated re-ionization of the neutral beam and the electromagnetic fields in this region. The drift duct bellows (with an inner diameter of 2.5 m) shall guarantee a leak tight vacuum enclosure during the vertical and radial displacements of the ITER vacuum vessel. The conductance of the VVPSS box must be maximized in the available space. The absolute valve remains a challenging development. The total leak rate through the valve must be ≤1 × 10−8 Pa m3/s when the valve is closed. Due to the radiation environment, the seals of the gate valve will be metallic. An R&D program has been launched to develop a suitable metallic seal solution with the required dimensions. The maximum allowed closing time for the fast shutter shall be less than 1 s. For all these components the leak tightness will be guaranteed by a welded lip seal and the mechanical stability by bolted structures.

The realization of the ITER Neutral Beam Test Facility (NBTF) and the start the experimental phase are important tasks of the fusion roadmap, since the target requirements of injecting to the plasma a beam of Deuterium atoms with a power up to 16.5 MW, at 1MeV of energy and with a pulse length up to 3600s have never been reached together before. The ITER NBTF, called PRIMA (Padova Research on ITER Megavolt Accelerator), is hosted in Padova, Italy; it includes two experiments: MITICA, the full-scale prototype of the ITER injector and SPIDER, the full-size negative ion source. The realization promoted by the ITER organization is carried out with the contribution of the European Union, channeled through the Joint Undertaking for ITER (F4E), of the Consorzio RFX which hosts the Test Facility, the Japanese and the Indian ITER Domestic Agencies (JADA and INDA) and several European laboratories, such as IPP-Garching, KIT-Karlsruhe, CCFECulham, CEA-Cadarache. The early start of operation of PRIMA experiments is urgent because sufficient experimental time is necessary to face and solve the issues related to the achievement of the desired performance in time for the ITER operation, requiring NBI since the beginning. Substantial progresses have been recently achieved: the buildings construction, begun in October 2012, has been completed by the end of 2015 and the installation of some components has been started since the end of 2014. The SPIDER realization is well advanced: the installation phase is proceeding in good agreement with the general plan; it is expected to be almost completed by the end of 2016. In parallel, the commissioning of the SPIDER power supply (PS) and auxiliary plant systems is being proceeding. Tests at full power and remote control are planned, including also those addressed to reproduce the grid breakdowns and to test the relevant protections. The design of the MITICA injector components was completed in 2015their procurement is being made through a number of tenders, some of them already launched. The HV Power Supply system of the MITICA 1MV accelerator, provided by JADA, was delivered on site in December 2015. The challenging installation of these components, including the step-up transformers and the SF6 gas insulated transmission line, started soon after and will go on throughout 2016. The present phase, with the PRIMA buildings continuously filled with new components, with the installation activities progressing and with also the commissioning and testing phase starting represents a substantial step forward toward the main target. The paper will describe the main challenges the Project Team has dealt during this phase and the important feedback derived for the ITER HNB systems both from the technical and the organizational standpoints.

The ITER project requires at least two Neutral Beam Injectors (NBIs), each accelerating to 1MV a 40A beam of negative deuterium ions, to deliver to the plasma a power of about 33 MW for one hour as additional heating.Since these requirements have never been experimentally met, it was recognized necessary to build-up a test facility, named PRIMA (Padova Research on ITER Megavolt Accelerator), in Italy, including both a 100 kV full-size negative ion source (SPIDER - Source for Production of Ion of Deuterium Extracted from Rf plasma) and a prototype of the whole 1 MV ITER injector (MITICA - Megavolt ITER Injector & Concept Advancement).The mission of SPIDER is to increase the understanding of the source operation and to optimize the source performance in terms of extracted current density, current uniformity and duration.The SPIDER experiment needs a dedicated Cooling Plant to remove up to 10 MW heat loads applied during the one hour pulses to the beam source, the beam dump/calorimeter and the power supplies. The total water flow rate necessary to remove the heat and guarantee the optimal control of temperatures is about 150 kg/s at full operational conditions.The plant is widespread over the whole PRIMA site area and is characterized by several physical interfaces with the SPIDER experiment, the related power supplies and control system and finally with PRIMA buildings and civil plants. A well-equipped system of sensors and feed-back control system is also present to guarantee the correct performances of the plant during different operational phases.The paper deals with the installation and testing phases of the plant under execution, with particular emphasis on integration problems, electrical insulation, tests of immunity to electromagnetic interference, vibration issues, monitoring and safety issues for possible activated water and finally the acceptance tests before integrated commissioning.

SPIDER is one of two projects of the ITER Neutral Beam Test Facility under construction in Padova, Italy, at the Consorzio RFX premises. It will have a 100 keV beam source with a full-size prototype of the radiofrequency ion source for the ITER neutral beam injector (NBI) and also, similar to the ITER diagnostic neutral beam, it is designed to operate with a pulse length of up to 3600 s, featuring an ITER-like magnetic filter field configuration (for high extraction of negative ions) and caesium oven (for high production of negative ions) layout as well as a wide set of diagnostics. These features will allow a reproduction of the ion source operation in ITER, which cannot be done in any other existing test facility. SPIDER realization is well advanced and the first operation is expected at the beginning of 2018, with the mission of achieving the ITER heating and diagnostic NBI ion source requirements and of improving its performance in terms of reliability and availability. This paper mainly focuses on the preparation of the first SPIDER operations - integration and testing of SPIDER components, completion and implementation of diagnostics and control and formulation of operation and research plan, based on a staged strategy

This paper presents a proposal of Cooling Plant for two new Neutral Beam experiments called MITICA and SPIDER to be realized in Padova (Italy). A large amount of Power (up to 70 MW) has to be removed from in-vessel components and auxiliary systems belonging to these two experiments. Different experimental scenarios (pulse duration ranging from few seconds up to 3600 s), requirements for operating temperature, coolant quality and voltage holding are taken into account in this conceptual design proposal. To reduce the radiological risks due to possible presence of activated corrosion products (ACP) in some water cooled components suitable design choices have been analysed. This work was carried out by considering carefully a lot of different aspects like operability, standardization of components, maintenance and repair, optimization of the installed power and the overall costs of the plant. Experiment components with similar requirements are grouped in the same primary circuits where fine temperature regulation, water quality monitoring and calorimetric measurements are the main characteristics. Each primary circuit (PC) is connected to secondary circuits which allow thermal dissipation and, in some cases, also component preheating. Secondary circuits are connected to two large basins the water of which is cooled down by active cooling rejection system such as cooling towers and air coolers. In this way the requirement for impulsive heat dissipation is fulfilled by the water basins allowing to install a less powerful active rejection system and so reducing the total costs. A large effort was done to guarantee good plant integration with the Experiment Main Hall (in which MITICA and SPIDER are located) and other technical supplies, buildings and areas. Other special requirements for stand-alone systems like Draining and Drying System, Pressure Test System and Chemical Control System are also part of this work.

A new Passive Active Multijunction (PAM) Lower Hybrid heating and Current Drive (LHCD) launcher, has been successfully manufactured and tested on Tore-Supra (TS). The design and the fabrication of this new actively cooled launcher based on the PAM concept, as the present ITER LHCD design, is a major component of the TS CIMES project (Components for the Injection of Matter and Energy in Steady-state), and will play a key role in the TS near term program. To achieve 1000 s pulses with a power flux of 25 MW/m 2 the PAM launcher has been designed for steady state (CW) operation (active cooling) with the objective of coupling 2.7 MW of LHCD power to the plasma at 3.7 GHz with a parallel index N ∥ = 1.7 ± 0.2. The launcher has achieved its qualifications tests, i.e. low power Radio Frequency measurements, vacuum and hydraulic leak tests, and has been installed on Tore-Supra tokamak in September 2009. It is commissioning on plasma started a month later, quickly achieving its design performance of 2.7 MW on a 35 s pulse. After a technical description of the PAM, this paper presents an overview of the project phases (RF optimization, manufacturing and qualification) and concludes with the first experimental results of the PAM.

International audience; Reliable and repetitive high power and long pulse tokamak operation is strongly dependant of the ability to secure the Plasma Facing Components (PFCs). In Tore Supra, a network of 7 infrared (IR) video cameras is routinely used to prevent PFCs overheating and damage in selected regions. Real time feedback control and offline analysis are essential for basic protection and understanding of abnormal thermal events. One important limitation detected by the IR real time feed-back loop during high power RF operation (injected power of 9.5 MW over 26 s and 12 MW over 10 s have been achieved respectively in 2006 and 2008) is due to the interaction between fast ions which increase the power flux density and flaking of the boron carbide coatings on the Faraday screen box of the ICRH antennas. An IR-based experimental procedure is proposed in order to detect new flakes during plasma operation. The thermal response of the B(4)C coating is studied with and without flaking during plasma operation. The experimental heat flux deposited by fast ion losses on the Faraday screen is calculated for high (3.8T) and low magnetic field (2 T) during high RF power operation (with fundamental hydrogen minority and second harmonic ICRH heating schemes respectively). The paper addresses both thermal science issues applied to machine protection and limitation due to fast ions issues during high RF power, long pulse operation. Safety margin to critical heat flux and number of fatigue cycles under heat load are presented in the paper. (C) 2011 Elsevier B.V. All rights reserved.

The ITER Ion Cyclotron Resonant Heating (ICRH) antenna provides plasma heating at a power of 20 MW. Operation in the ITER environment imposes significant thermal power handling capability, structural integrity, shielding and operations requirements. The design will require a step change over any predecessor in terms of power, scale and complexity. This paper reports the main mechanical design features that address the challenges and often conflicting requirements during the conceptual design phase.

SPIDER is one of the two projects of the ITER Neutral Beam Test Facility (NBTF) under construction in Padova, Italy, at Consorzio RFX premises; it is the full-size prototype of the radiofrequency (RF) ion source for ITER Neutral Beam Injector (NBI), with beam energy of 100keV, designed to operate with pulse length up to 3600 s, featuring ITER-like filter field configuration, caesium oven layout and a wide set of diagnostics. These features will allow reproducing the ion source operation like in ITER as cannot be done in any other existing test facility. SPIDER realization is well advanced and the first operation is expected at the beginning of 2018, with the mission to prove the possibility of achieving the ITER NBI ion source target requirements and of improving its performance in terms of reliability and availability. This paper mainly focuses on the preparation of the first SPIDER operations: integration and testing of SPIDER components, completion and implementation of diagnostics and control and workout of operation and research plan, based on a staged strategy.

The suppression of ELMs with a resonant magnetic perturbation (RMP), as demonstrated in DIIID experiments [T. Evans, et al., Suppression of large edge localized modes with edge resonant magnetic fields in high confinement DIII-D plasmas, Nucl. Fusion 45 (2005) 595–607 [1] ], and widely published [K.H. Burrell, et al., Plasma Phys. Control., Fusion 47 (2005) B37–B52 [2] ; M. Becoulet, et al., Edge localized modes control by stochastic magnetic field, Nucl. Fusion 45 (2005) 1284–1292 [3] ], is an attractive solution to the problem of divertor target erosion in ITER. Ideally, the RMP should be strong at the plasma edge and weak in the core, which requires the installation of the RMP coils as close as possible to the plasma. In both ITER and JET, an installation of the RMP coils within the vacuum vessel would subject the coils to neutron irradiation, tritium contamination and all the problems of furnishing cooling and electrical services, across many interfaces. So for JET, it has been proposed to install an “ITER prototype” RMP system located ex-vessel.

The project of an ITER-like (IL) ion cyclotron (IC) prototype antenna was initiated in mid 2002 in Cadarache, to assess the operational characteristics of the IL scheme in a Tore Supra (TS)-sized array. The prototype was developed by modifying an existing Tore Supra resonant double loop (RDL) antenna. This strategy, chosen to reduce cost and manufacturing time, allowed the completion of the array construction at the end of 2003. The array was installed on TS in January 2004, at the beginning of the new experimental campaign. After a few comments on the design, the paper reports results on low-power characterisation, high-power commissioning and preliminary tests on TS plasmas, prior to the array dis-assembly for inspection.

The 118 GHz electron cyclotron heating and current drive (ECRH/ECCD) system under development in Cadarache, France, for use on the Tore Supra tokamak (Pain M. et al 1994 Proc. 18th SOFT (Karlsruhe) pp 481–4: Darbos C. et al 2000 Proc. 21st SOFT (Madrid) pp 605–9), is designed to launch 2.4 MW of power for up to 10 min into the plasma. At present two out of six gyrotrons are installed and available for injection of up to 800 kW. This paper concentrates on the generation and transmission of the ECRH/ECCD power for very long pulse operation. The power is injected into the plasma as Gaussian beams by an antenna which, using actively cooled mirrors inside the Tore Supra vacuum vessel, allows extensive control of both the poloidal and toroidal injection angles. The toroidal field on Tore Supra is normally in the range of 3.8–4 T, which for 118 GHz gives almost central deposition at the fundamental electron cyclotron resonance. A pair of actively cooled corrugated mirrors is installed in each matching optics unit at the output of each gyrotron allowing complete control of the polarization of the wave transmitted to the antenna, with the result that pure O-mode—or pure X-mode—power injection can be achieved for all injection angles. In tokamak experiments, a world record energy of 17.8 MJ has been injected into the plasma. New upgraded gyrotrons specified to produce 400 kW for up to 10 min will be introduced over the next 3–4 years.

Significant spin-off from magnetic fusion research in Euratom-CEA association, over the last 40 years, has been induced and developed through a continuous process of exchange of scientific, technology and managerial expertise between the fusion scientists and manufacturing engineers. The growth in shared expertise, associated innovative applications and cooperative efforts with industry can be clearly identified (i) in the frame of the European Fusion Development Agreement (EFDA) and underlying technology programme, (ii) by the industrial applications induced from Tore Supra programme and the associated joint development of large test bed facilities for control and acceptance test, (iii) by the appreciation of the expected impacts of ITER from the companies involved in the Tore Supra construction.

Ten years after the first ion cyclotron range of frequency (ICRF) heated Tore Supra (TS) plasma shot, and as TS facility, now equipped with the new ‘CIEL’ high power handing in-vessel components (This SOFT), is progressing towards long pulse performances, the main results and experience gained from the ICRF system are reviewed. An overview of the present status of TS ICRF facility is given, with highlights on the main modifications made to the initial plant and associated results. TS ICRF system achievements are then reported from a large ICRF heated pulses database covering ten experimental campaigns (1991–2001). The main technological issues encountered, especially those relevant to next step ICRF system design, are also discussed. Finally, in view of extended pulse length performances on TS, the present limitations of the ICRF facility and future plans are outlined.

The high technology components used in fusion devices, especially high heat flux (HHF) plasma facing components, require high reliability. This can be only guaranteed by a very high level of quality control (QC) through rigorous acceptance inspection procedure. Qualification of the power exhaust capabilities of HHF components in normal operation requires control of their thermal and mechanical integrity. During the last 4 years, the Tore Supra QC team improved acceptance inspection procedures for HHF components, which have been 100% controlled before their installation in the Tokamak. Innovative investigations of non-destructive techniques, such as pressure drop test, helium leak testing procedure, infrared thermography, were performed and it was shown that these techniques are essential tools for QC of critical parts.

ITER-like IC antennas, fulfilling or possibly exceeding ITER requirements and designed for long pulse operation (1000 s) are currently being developed at Cadarache. As part of these developments, a project of an IC antenna prototype obtained by modifying a tore supra (TS) resonant double loop (RDL) antenna is proposed, to assess in real working conditions the performance of this scheme. After a brief comparison between the new RF circuit version and the conventional RDL circuit, which has been tested in TS for 10 years, the paper focuses on the design and manufacturing aspects of the prototype.

The ITER project requires additional heating by two neutral beam injectors, each accelerating to 1 MV a 40 A beam of negative deuterium ions, to deliver to the plasma a power of about 17 MW for one hour. As these requirements have never been experimentally met, it was recognized as necessary to setup a test facility, PRIMA (Padova Research on ITER Megavolt Accelerator), in Italy, including a full-size negative ion source, SPIDER, and a prototype of the whole ITER injector, MITICA, aiming to develop the heating injectors to be installed in ITER. This realization is made with the main contribution of the European Union, through the Joint Undertaking for ITER (F4E), the ITER Organization and Consorzio RFX which hosts the Test Facility. The Japanese and the Indian ITER Domestic Agencies (JADA and INDA) participate in the PRIMA enterprise; European laboratories, such as IPP-Garching, KIT-Karlsruhe, CCFE-Culham, CEA-Cadarache and others are also cooperating. Presently, the assembly of SPIDER is on-going and the MITICA design is being completed. The paper gives a general overview of the test facility and of the status of development of the MITICA and SPIDER main components at this important stage of the overall development; then it focuses on the latest and most critical issues, regarding both physics and technology, describing the identified solutions.

The paper reports on latest developments for the ITER Ion Cyclotron Heating and Current Drive system: imminent acceptance tests of a prototype power supply at full power; successful factory acceptance of candidate RF amplifier tubes which will be tested on dedicated facilities; further design integration and experimental validation of transmission line components under 6MW hour-long pulses. The antenna Faraday shield thermal design has been validated above requirements by cyclic high heat flux tests. R&D on ceramic brazing is under way for the RF vacuum windows. The antenna port plug RF design is stable but major evolution of the mechanical design is in preparation to achieve compliance with the load specification, warrant manufacturability and incorporate late interface change requests. The antenna power coupling capability predictions have been strengthened by showing that, if the plasma scrape-off layer turns out to be steep and the edge density low, the reference burning plasma can realistically be displaced to improve the coupling.

A systematic study of the thermal behaviour of ion cyclotron range of frequencies (ICRF) antenna front faces was undertaken on the Tore Supra tokamak over two years of plasma operation, by means of infrared (IR) and visible light cameras. Among the variety of edge-antenna interaction phenomena observed experimentally, the present paper focuses on the most deleterious effect for ICRH operation, a non-resonant radio-frequency (RF) process causing hot spots on the Faraday screen and bursts of metallic impurities. The hot spots phenomenology (time history, location) is presented. Their magnitude is characterized by semi-quantitative indicators, defined from the IR and visible light films. This makes possible a parametric study of the antenna–plasma interaction, over a wide range of experimental configurations. The roles of the local RF electric field and of the edge density regimes in the spurious process are outlined. The observed hot spot behaviour is compatible with the build-up of RF sheaths. Some ways of reducing the problems are suggested.

Tore Supra is involved in flat tile carbon armoured plasma facing components (PFCs) since 1985. In 1997, a third generation of components, based on the original concept developed with Plansee Company, called active metal casting (AMC®), has been launched. Since 1998, 660 elementary components for the toroidal pump limiter (TPL) are in production. The route of the manufacture is rather complex and many controls were requested all along the fabrication to insure a high reliability of the elements. One of the main controls is the final infrared (IR) test allowing to determine the quality of the bonding between the carbon fibre composite (CFC) tiles and the heat sink made of copper–chromium–zirconium alloy (CuCrZr). Although results for the first batch of elements were as expected (less than 5% rejected at the final test), unexpected defects appeared with the followings batches. Investigations on the fabrication processes underlined the importance of having a better heat treatment of the pieces in copper alloy (CuCrZr), however this was not sufficient to completely explain the observed defects.

As an alternative to commercial capacitors which have experienced reliability problems and are irrelevant to the next step, a new all-metal impedance concept has been elaborated for Tore Supra ICRF Resonant Double Loop (RDL) antenna. Two detailed design versions are presented, both based on the principle of a parallel LC transformer inserted in a coaxial structure. First low power measurement results on a full-scale mock-up are discussed.

A new actively cooled advanced launcher is being built for Tore Supra LHCD to inject 4 MW during 1000 s at 3.7 GHz, at a power density of 25 MW/m 2 (a conservative value observed in Tore Supra experiments). It is made from an array of 6×48 active and 6×9 passive waveguides. The design uses technologies which are relevant for a next step machine such that it can: (i) withstand a plasma radiated flux of 0.15 MW/m 2 ; (ii) radiate power with spectra having peak N// values of 2.02±0.35; (iii) withstand a total torque of 8.6 10 4 N m during disruptions; (iv) allow an antenna 20 cm radial stroke adjustable in real time, (v) withstand a convected power flux of 10 MW/m 2 on its guard limiter made of CFC tiles. A prototype of each new component of this antenna has been tested successfully at the nominal power with a pulse length of 1000 s.

During the 1997 and 1998 Tore Supra shutdown, a first set of new antenna guard limiters was installed on one of the three ion cycloton resonance heating (ICRH) antennas of Tore Supra. This limiter, which was one of the main technological studies of the 1998 campaign, was widely experimented in real plasma conditions, thus allowing the validation in situ, for the first time, of the technology of active metal casting (AMC) for plasma facing components. The huge improvement in the thermal response of the new limiter generation, compared to the old one, is shown on plasma pulses made identical in terms of antenna position and injected RF power profile. By using the infrared cameras installed inside Tore Supra and viewing the antennas front, the power density fluxes received by the carbon fibre composite (CFC) surface of the limiter were evaluated by correlation with the heat load tests made on the electrons beam facility of CEA/Framatome.

The Vacuum vessel of the ITER Heating Neutral Beam injector is a key component of the neutral beam system. It supports the injector components and is also an extension of the ITER Vacuum Vessel. This paper presents the status of the design of the vessel. In the first part of this paper we present the main design requirements. Besides the pressure and thermal loads, the various interfaces with other components are important constraints for the design. Some have specific requirements concerning displacement, or require specific features for their maintenance. Indeed all maintenance involving opening of the vessel needs to be performed using remote handling systems. The second part aims at presenting the present design of the NB vessel. Various concepts were studied before coming to the present design. Structures were optimised to satisfy the chosen design code (RCC-MR) under the various loading scenarios applied to the vessel, both for normal and accidental cases. In addition, the vessel incorporates two sets of long lip seals which tolerate limited displacements. Limiting the displacement is a real challenge considering the size of the flange and the forces involved. An innovative solution has been developed to improve the mechanical connection between the lids and the vessel walls in order to fulfil the requirements. The vessel displacement is also important for the alignment of the Beam Line Components with respect to the neutral beam and for the cryopumps interfaces. The geometry of the vessel was optimised to reduce the electric field stresses and to increase the voltage holding of the Beam Source, which is at 1MV potential. The present design satisfies all the requirements. The main issues have been solved, and the proposed design is now being optimised to facilitate manufacturing.

The Vacuum vessel of the ITER Heating Neutral Beam injector is a key component of the neutral beam system. It is host to the injector components and is also an extension of the ITER Vacuum Vessel. This article presents the s tatus of the design of the vessel. The first part of this article is dedicated to the presentation of the main design requirements. Besides the pressure and thermal load s, the various interfaces with other components are important constraints for the design . Some have specific requirements concerning displacement, or require specific featur es for their maintenance. Indeed all maintenance involving an opening of the vessel need s to be performed using remote handling systems. The second part aims at presenting the present desi gn of the NB vessel. Various concepts were studied before coming to the present day design. Structures were optimised to satisfy the RCC-MR code under the various loadin g scenarios applied to the vessel, both for normal and accidental cases. In addition, the vessel incorporates two sets of long lip seals which tolerate limited displacements. Lim iting the displacement is a real challenge considering the size of the flange and th e forces involved. An innovative solution has been developed to improve the mechanic al connection between the lids and the vessel walls in order to fulfil the requirement s. The vessel displacement is also important for the alignment of the Beam Line Compon ents with respect to the neutral beam and for the cryopumps interfaces. The geometry of the vessel was optimised to reduce the electric field stresses and to increase the voltage holding of the Beam Source, which is at 1MV. The present design satisfies all the requirements. The main issues have been solved, and the proposed design is now being optimised to f acilitate manufacturing.

The ongoing design of the ITER Ion Cyclotron Heating and Current Drive system (20 MW, 40-55 MHz) is rendered challenging by the wide spectrum of requirements and interface constraints to which it is subject, several of which are conflicting and/or still in a high state of flux. These requirements include operation over a broad range of plasma scenarios and magnetic fields (which prompts usage of wide-band phased antenna arrays), high radio-frequency (RF) power density at the first wall (and associated operation close to voltage and current limits), resilience to ELM-induced load variations, intense thermal and mechanical loads, long pulse operation, high system availability, efficient nuclear shielding, high density of antenna services, remote-handling ability, tight installation tolerances, and nuclear safety function as tritium confinement barrier. R&D activities are ongoing or in preparation to validate critical antenna components (plasma-facing Faraday screen, RF sliding contacts, RF vacuum windows), as well as to qualify the RF power sources and the transmission and matching components. Intensive numerical modeling and experimental studies on antenna mock-ups have been conducted to validate and optimize the RF design. The paper highlights progress and outstanding issues for the various system components. (C) 2013 ITER Organization. Published by Elsevier B.V. All rights reserved.

Activities towards the establishment of the NB Test Facility (NBTF) in Padua-Italy and those related to the procurement of the heating neutral beams for ITER have recently reached a good level ofprogress thanks to the finalization of the agreements on theNBTF between F4E (the EU Domestic Agency for ITER), Consorzio RFX (the host of the NB test facility) and the ITER organization. This paper presents the status of the design of the various components within the EU scope of procurement, with a focus on the modificationsimplemented in the last years as a result of intense R&D activity undertaken in EU.

A new concept of multijunction-type antenna has been developed, the Passive Active Multijunction, which improves the cooling of the waveguides and the damping of the neutron energy (for ITER) compared to Full Active Multijunction. Due to the complexity of the structures, prototypes of the mode converters and of the Passive-Active-Multijunction launcher were fabricated and tested, in order to validate the different manufacturing processes and the manufacturer's capability to face this challenging project. This paper describes the manufacturing process, the tests of the various prototypes and the construction of the final Passive-Active-Multijunction launcher, which entered into operation in October 2009. It has been commissioned and is fully operational on the Tore-Supra tokamak, since design objectives were reached in March 2010: 2.75 MW - 78 s, power density of 25MW/m2 in active waveguides, steady-state apparent surface temperatures < 350 °C; 10 cm long distance coupling.

The Ion Cyclotron Heating and Current Drive (IC H & CD) system for ITER will provide Radio frequency (RF) heating and current drive to the ITER plasma. This system is designed to provide 20 MW into the plasma using frequencies in the range of 40 MHz to 55 MHz, which could be upgraded to 40 MW in future. This paper presents the recent development of the IC H&CD system configuration in the ITER Tokamak complex & auxiliary buildings with developed interfaces.

This paper reviews the on-going design, R&D and procurement activities, mostly conducted within the ITER framework, on-going in Europe under the co-ordination of Fusion for Energy (F4E), in co-operation with the European Fusion Associations and aimed at the establishment of the ITER Heating Neutral Beam (HNB) system.

The ITER neutral beam (NB) system presents several challenges and a robust program is necessary in order to achieve the requirements within the tight constraints due to the ITER construction plan. The establishment of full scale NB test facilities (NBTF) has therefore become a centre piece of the international NB development strategy. This paper describes the status of the design activities that have been undertaken in Europe to develop the components for the heating NB injectors along with the main plans and results of the R&D activities. A description of the programme towards the establishment of the test facilities and the planned activities is also reported.

RF‐sheath induced heat loads are identified from infrared thermography measurements on Tore Supra ITER‐like prototype and JET A2 antennae, and are quantified by fitting thermal calculations. Using a simple scaling law assessed experimentally, the estimated heat fluxes are then extrapolated to the ITER ICRF launcher delivering 20 MW RF power for several plasma scenarios. Parallel heat fluxes up to 6.7 MW/m2 are expected very locally on ITER antenna front face. The role of edge density on operation is stressed as a trade‐off between easy RF coupling and reasonable heat loads. Sources of uncertainty on the results are identified.

The antenna design proposed here is based on the resonant double loop concept with conjugate T matching to make the circuit resilient to strong plasma load variations as ELMs. The antenna is constituted of two main parts; the in‐vessel launcher which is inside the primary torus vacuum and the Compact Vacuum Tuners (CVT) that is located after the first barrier in a private vacuum. This CVT allows to match at the strap location, the antenna impedance with the plasma load, over the 45 to 55 MHz frequency range. It has been designed to ease its repair and maintenance, and can be easily removed from the rear without breaking the primary vacuum. Apart from the Faradays screens fit to shape the plasma edge, the in‐vessel launcher and CVT are made out of 6 identical modules, to allow the best economical approach for the manufacture, the assembly and the maintenance of the antenna.

Since the release of the ITER ICRH system reference design report [ITER Final Design Report: DDD 5.1 –Ion Cyclotron and Current Drive System, July 2001], further design studies have been conducted. If the base of the reference design [Final Report on EFDA contract 04/1129, ITER ICRF antenna and Matching system design (Internalmatching), April 2005] is kept unchanged, several significant modifications have been proposed for a better efficiency and reliability. The increase of the poloidal order of the array and strong modifications of the matching system concept are the main changes. Technical aspects insufficiently covered in previous studies are also now worked out in detail, like the integration on a mid-plane port satisfying the constraints of the ITER environment.

In the framework of the ICRH development led at CEA Cadarache, a project of a prototype antenna based on the electrical layout foreseen for the ITER IC launcher has been initiated. First experiments have been performed on Tore Supra in 2004. Pulses of 500 kW lasting up to 6s have been achieved. The importance of the coupling effect between straps has been revealed during these experiments, and thus the need for an active control system of the current phasing between them. Despite the short experimental campaign terminated by the failure of the matching components due to undetected arcs, the load tolerant properties of such a circuit have been observed.Since, the prototype antenna has been repaired, reassembled and enhanced accordingly to the fruitful lessons harvested. The RF circuit was improved, corona rings were added on the capacitor poles and localized “sharp” edges were rounded to increase the voltage standoff capability. To reduce the toroidal coupling between adjacent straps, a thick poloidal sept...

The ECRH experiment on TORE SUPRA is designed to inject up to 3 MW of power at 118 GHz, using an antenna consisting of six fixed spherical mirrors and three mobile steering minors. The position of the mobile mirrors can be varied in real time using two stepper motors for each mobile mirror. In addition to controlling the injection angle, the position of the mobile mirrors also affects the polarisation of the injected wave. Accurate formulae to computer in real time, the stepper motor positions required to obtain the desired beam injection angles have been derived. Formulae to determine the effect on the wave polarisation, of the actual mobile mirror positions have also been determined. These formulae have been verified by precise laser measurements and by comparison of power deposition calculations and experimental results. Experiments with different values of polarisation and injection angles have been carried out in the 2001 campaign.

Although it is generally recognized that the current IC technology is applicable to next step devices, the capability of current IC antennas to efficiently couple power to next step plasma is sometimes questioned. The ITER design effort has identified requirements for an acceptable performance. ITER-like IC antennas, fulfilling or possibly exceeding ITER requirements and designed for long pulse operation (1000 s) are currently being developed at Cadarache. In the paper, their essential features are described. Applications to existing and next step devices am discussed.

The JET-EP antenna is designed to verify the design principles of that proposed for ITER, in relevant plasma conditions. It is intended to launch 7.2 MW of RF power to the plasma of the JET machine during its 'Enhanced Performance' program, at a power density of around 8 MW/m/sup 2/. The antenna comprises two poloidal current straps, each subdivided into four electrically short straps complete with in-vessel capacitors. The antenna is supported via a cantilever support box to the external support structure. This plug-in, through-port type assembly is clamped to the vacuum vessel externally. The antenna itself is fitted remotely. External rails and bellows allow radial positioning of the assembly. Eight capacitors are situated just behind the short antenna current straps. Plug-in replacement of the capacitors is performed through the vacuum vessel port.

The paper reports on the status of the JET ITER‐Like ICRF Antenna project and highlights main challenges that have come up during its design phase.

The Tore Supra ICRH antennas, built in collaboration with ORNL according to the Resonant Double Loop concept, have been used for many years in various experimental schemes. In the frequency range between 35 to 80 MHg they have contributed to meet the initial aim of the Tore Supra tokamak: to operate at a plasma current up to 1.7 MA, over 30 seconds or with a heating power higher than 10 MW. Based on the experience gained, the next step, aiming at high power operation up to 25 MW with long pulse discharges (1000 s), is scheduled on Tore Supra.

In the framework of the JET Enhanced Performance project, an Ion Cyclotron Range of Frequency in-port ITER-FEAT like launcher is being designed with the aim of coupling 8 MW to Elmy H-mode JET plasmas. A full scale flat geometry modular mock-up of one poloidal half of the Array was build in Cadarache to be submitted to an extensive low power measurement campaign in order to check its electrical features against the design model. Detailed results on the effect of top-bottom strap asymmetries, geometrical clearances, grounding, and mutual coupling are reported in the paper. Consequences on the launcher design are finally addressed.

Whereas the essential components for power generation and transmission on an ITER-sized ICRF plant are readily available, or only require modest extrapolation of present day ICRF systems, the IC launching structure itself is regarded as needing development, validation and convincing demonstration. During the ITER-EDA, the design requirements of the Next Step device have resulted in increasingly higher values for this power density, going from less than 1 MW/m2 with the launcher integrated in the blanket, to presently about 10 MW/m2 for the present ITER, where the launcher is required to fit inside an equatorial port. The presently proposed JET-Enhanced Performance project ITER-like ICRF launcher, due for installation during the 2004 shutdown, mainly features a high power density of about 8 MW/m2 at a relevant coupling of 4 Ohm/m and 1C heating scenarios at frequencies from 30 to 55 MHz for pulses up to 20 seconds with a tolerance to load variations due to ELMs.