In order to fully validate "ITER-like" actively water cooled tungsten plasma facing units, addressing the issues of long plasma discharges, an axisymmetric divertor structure has been studied and manufactured for the implementation in the WEST (W-Tungsten Environment in Steady state Tokamak) tokamak platform. This assembly, called divertor structure and coils (4 m diameter, 20 t), is composed of two stainless steel casings containing an actively water cooled (up to 180 degrees C, 4 MPa) copper winding pack designed for a conductor current in the range of 12.5kA (up to 1000s). It must sustain harsh environmental conditions in terms of ultra-high vacuum, high temperatures and electrodynamic loads. One major difficulty is the assembly by induction brazing of individual bended conductor sectors inside the vacuum vessel and the consecutive sealing of the casings by TIG welding. Therefor development activities have been carried out on a scale one dummy coil, such as brazing, assembly, thermal cycling and electrical insulation tests (5 kV ground voltage). Whereas the brazing assembly technics and the conductor installation were validated without major difficulties, different technical solutions for the electrical insulation had to be tested. The chosen solution is a resin epoxy impregnated fiber glass fabric layered around the conductors followed by a polymerization procedure. In parallel the manufacturing of divertor structure components started in the second half of 2013 with a total delivery at the end of 2015. The paper will illustrate the technical developments which have been performed in order to fully validate the design. It concerns mainly the dummy coil and the complex conductor installation procedureassisted by virtual reality tools. The manufacturing methods proposed by industry in order to fulfil the technical requirements will be also addressed. Finally the processes and associated tools used in order to implement this large component inside the WEST vacuum vessel will be detailed. (C) 2017 Elsevier B.V. All rights reserved.

Embedded RF contacts are integrated within the ITER ICRH launcher to allow assembling, sliding and to lower the thermo-mechanical stress. They have to withstand a peak RF current up to 2.5 kA at 55 MHz in steady-state conditions, in the vacuum environment of the machine. The contacts have to sustain a temperature up to 250 degrees C during several days in baking operations and have to be reliable during the whole life of the launcher without degradation. The RF contacts are critical components for the launcher performance and intensive R&D is therefore required, since no RF contacts have so far been qualified at these specifications. In order to test and validate the anticipated RF contacts in operational conditions, CEA has prepared a test platform consisting of a steady-state vacuum pumped RF resonator. In collaboration with ITER Organization and the CYCLE consortium (CYclotron CLuster for Europe), an R&D program has been conducted to develop RF contacts that meet the ITER ICRH launcher specifications. A design proposed by CYCLE consortium, using brazed lamellas supported by a spring to improve thermal exchange efficiency while guaranteeing high contact force, was tested successfully in the T-resonator up to 1.7 kA during 1200 s, but failed for larger current values due to a degradation of the contacts. Details concerning the manufacturing of the brazed contacts on its titanium holder, the RF tests results performed on the resonator and the non-destructive tests analysis of the contacts are given in this paper. (C) 2015 Elsevier B.V. All rights reserved.

The fusion machines under development or construction (ITER, W7X) use several hundreds of actively cooled plasma facing components (ACPFC). They are submitted to leak tightness requirements in order to get an appropriate vacuum level in the vessel to create the plasma. During the ACPFC manufacturing and before their installation in the machine, their leak tightness performance must be measured to check that they fulfill the vacuum requirements. A relevant procedure is needed which allows to segregate potential defects. It must also be optimized in terms of test duration and costs.Tore Supra, as an actively cooled Tokamak, experienced several leaks on ACPFCs during the commissioning and during the operation of the machine. A test procedure was then defined and several test facilities were set-up. Since 1990 the tightness of all the new ACPFCs is systematically tested before their installation in Tore Supra. During the qualification test, the component is set up in a vacuum test tank, and its cooling circuits are pressurized with helium. It is submitted to 3 temperature cycles from room temperature up to the baking temperature level in Tore Supra (200 degrees C) and two pressurization tests are performed (6 MPa at room temperature and 4 MPa at 200 degrees C) at each stage. At the end of the last cycle when the ACPFC is at room temperature and pressurized with helium at 6 MPa, the measured leak rate must be lower than 5 x 10(-11) Pa m(3) s(-1), the pressure in the test tank being <5 x 10(-5) Pa. A large experience has been gained on ACPFCs with carbon parts on stainless steel and CuCrZr structures (toroidal pumped limiter (TPL) composed of 12 sectors of 48 fingers) but also on all metallic ones (ICRH antennas and LHH launchers, vacuum vessel protection walls similar to 60 m(2)). During the tests, some leaks were found and the components had to be repaired. The analysis of the results database allows defining guidelines to optimize the presently used procedure for the new ACPFCs generation which will be set-up in WEST or ITER (tungsten (W) elements, high mass components).This paper presents the experience of the Tore Supra teams in the implementation of ACPFC with regard to the tightness issues including the tests procedures which were developed and used since 1990. The prospects for the new generation actively cooled machines are also discussed. (C) 2015 Elsevier B.V. All rights reserved.

In order to fully validate "ITER-like" actively water cooled tungsten plasma facing units, the implementation of an axisymmetric divertor structure in the Tokamak Tore-Supra has been studied. With this major upgrade, the so-called WEST (Tungsten Environment in Steady state Tokamak), Tore-Supra will be able to address the issues of long plasma discharges using a tungsten divertor based on monoblock targets.The divertor structure and coils assembly are made up of two stainless steel casings containing a copper winding pack cooled by a pressurized hot water circuit (up to 180 degrees C, 4 MPa) in which a total divertor current of up to 16 x 13 kA is circulating in steady state. The conductor is electrically insulated and wedged inside the casing in order to be mechanically protected.The divertor which is designed to perform steady state plasma operation (up to 1000 s), must sustain harsh environmental conditions in terms of ultra light vacuum conditions, electromagnetical loads and electrical insulation (5 kV ground voltage) under high temperature (180 degrees C).Therefore, a feasibility study of such a complex structure has been performed. It implied activities on a scale one dummy coil, such as installation, assembly issues and representative tests (electric, thermal and hydraulic).The manufacturing of the divertor structure, which is a large assembly of 4-m diameter representing a total weight of around 20 tonnes, started in the second half of 2013 and is expected to be delivered by the end of 2014.The paper will illustrate the technical developments and tests performed during 2013 and beginning of 2014 in order to fully validate the design concept before the industrial phase. The manufacturing methods proposed by the contractor in order to fulfil the technical requirements will be also addressed. (C) 2015 Elsevier B.V. All rights reserved.

In parallel to the direct contribution to the procurement phase of ITER and Broader Approach, CEA has initiated research & development programmes, accompanied by experiments together with a significant modelling effort, aimed at ensuring robust operation, plasma performance, as well as mitigating the risks of the procurement phase. This overview reports the latest progress in both fusion science and technology including many areas, namely the mitigation of superconducting magnet quenches, disruption-generated runaway electrons, edge-localized modes (ELMs), the development of imaging surveillance, and heating and current drive systems for steady-state operation. The WEST (W Environment for Steady-state Tokamaks) project, turning Tore Supra into an actively cooled W-divertor platform open to the ITER partners and industries, is presented.

Actively cooled tungsten plasma facing components will be used in the ITER divertor. In order to fully validate such a technology (industrial manufacturing, operation with long plasma duration), the implementation of a tungsten axis symmetric divertor in the tokamak Tore-Supra is studied. With this major upgrade, so called WEST (Tungsten Environment in Steady state), Tore-Supra will be the only European tokamak able to address the problematic of long plasma discharges with an actively cooled metallic divertor.To do so, it is planned to install two symmetric divertor coils inside the vacuum vessel. This assembly, called divertor structure, is made up of two stainless steel casings containing a copper winding pack cooled by hot pressurized water (200°C, 4MPa). These two casings are located at the top and bottom of the vacuum vessel in order to create two magnetic X-point areas, which are protected by W-PFCs (Tungsten Plasma Facing Components) in order to extract the thermal loads. The two casing are robustly maintained together by 18 brackets in order to constitute a rigid assembly attached thanks to 12 legs (one per lower vertical port) outside the Tore_Supra vacuum vessel.The paper will illustrate the technical developments performed during 2011 in order to produce a preliminary design of the Tore-Supra WEST divertor structure with a particular focus on: the mechanical design of this major component and its integration in the Tokamak, the manufacturing issues and the technical results of the feasibility studies done with industry as well as the design of a scale one coil mock up.

After the installation and commissioning of the TITAN [1] , [2] , test facility, a key element – the T resonator – was assembled in order to facilitate testing components at high RF voltages and currents. This work is within the CEA roadmap for testing ITER ICRH components in a relevant environment. Several components of the future ITER ICRH antenna have been targeted. The embedded RF contact within the ITER ICRH antenna appeared as a critical component for antenna performance, requiring extensive R&D. Therefore, CEA has proposed and subsequently prepared a platform to test and validate the anticipated RF contact. A steady state resonator with active water cooling has been manufactured and assembled within the TITAN facility, including a hot pressurized water loop. The program consists of testing the contact at 2.25 kA and 62 MHz in steady state conditions. Sliding tests are also performed at high temperature and vacuum to understand component aging, including wear. The equipment installed is consistent with that required to test an ITER ICRH extensively.

In order to fully validate actively cooled tungsten plasma facing components (industrial fabrication, operation with long plasma duration), the implementation of a tungsten axisymmetric divertor structure in the tokamak Tore-Supra is studied. With this major upgrade, so-called WEST (Tungsten Environment in Steady state Tokamak), Tore-Supra will be able to address the problematic of long plasma discharges with a metallic divertor target.To do so, it is planned to install two symmetric divertor coils inside the vacuum vessel. This assembly, called divertor structure, is made up of two stainless steel casings containing a copper winding pack cooled by a pressurized hot water circuit (up to 180 C, 4 MPa) and is designed to perform steady state plasma operation (up to 1000 s).The divertor structure will be a complex assembly ring of 4 m diameter representing a total weight of around 20 tons. The technical challenge of this component will be the implementation of angular sectors inside the vacuum vessel environment (TIG welding of the coil casing, induction brazing and electrical insulation of the copper winding). Moreover, this complex assembly must sustain harsh environmental conditions in terms of ultra high vacuum conditions, electromagnetical loads and electrical isolation (13 kV ground voltage) under high temperature.In order to fully validate the assembly and the performance of this complex component, the production of a scale one dummy coil is in progress.The paper will illustrate, the technical developments performed in order to finalize the design for the call for tender for fabrication. The progress and the first results of the simplified dummy coils will be also addressed. (C) 2013 Published by Elsevier B.V.

The design and overall dimensions of a 5 GHz TE10-TE30 mode converter are presented. This mode converter is a RF element of a 20 MW CW lower hybrid system proposed for ITER. A low power mock-up of this device has been manufactured at CEA/IRFM and measured at low power. RF measurements indicate a return loss of 40 dB and a transmission loss of 4.78 dB +/- 0.03 dB for the three outputs. The forward conversion efficiency from TE10 mode to TE30 has been measured from electric field probing to 99.9%. The good RF performances obtained validate the RF design of this element. (C) 2012 Elsevier B.V. All rights reserved.

Tore Supra (TS) has been designed to operate using technologies that allow long plasma operation (a few minutes), by means of superconducting magnets and actively-cooled high heat flux plasma facing components (PFCs). Actively cooled tungsten PFC will be used in the baffle area of the first ITER divertor. In order to validate such a technology fully (industrial manufacturing, operation with long plasma duration), the implementation of a tungsten axi-symmetric divertor in the tokamak Tore Supra has been studied [1]. With this second major upgrade, Tore Supra should be able to address the problematic of long plasma discharges with a metallic divertor.The proposed divertor is made up of two stainless steel casings containing a copper coil winding located at the top and bottom area of the vacuum vessel. These casings are firmly maintained by connection beams and protected by PFC. This paper describes the mechanical design of this major component and its integration in TS, the associated electromagnetic and thermomechanical analysis, the manufacturing issues and finally the integration of ITER representative PFCs. (C) 2011 Elsevier B.V. All rights reserved.

A one year-long major upgrade of the 3.7 GHz Lower Hybrid Current Drive (LHCD) generator for the Tore Supra (TS) tokamak has been performed. It consisted in installing a first series of eight Thales Electron Devices (TED) 700 kW CW klystrons, new CW components and auxiliaries, and in modifying the transmitter control and protection software. Modifications and calibration of the sensors and the RF subsystems were completed as well. Finally, the RF power available in the generator has been increased by 35% and the pulse duration could reach 1000 s. A complete validation and optimization of the klystrons have been performed in 2010 on matched load before the generator could enter into operation. The eight klystrons connected with the Full Active Multijunction (FAM) antenna delivered 3.5 MW/50s in December 2010. The upgrade of the generator and the steps to validate the modifications are described.

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 design and overall dimensions of a RF window and a TE10—TE30 mode converter of a 5 GHz Lower Hybrid system which should be able to deliver 20 MW CW on ITER are presented. A low power mock‐up of the mode converter has been manufactured and measured. The good RF performances obtained validates the RF design of this element.

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: 2.75MW – 78s, power density of 25MW/m2 in active waveguides, steady-state apparent surface temperatures <350°C; 10cm long distance coupling.

The design and the fabrication of a new Lower Hybrid (LH) actively cooled antenna based on the passive active concept is a part of the CIMES project (Components for the Injection of Mater and Energy in Steady-state). The major objectives of Tore-Supra program is to achieve 1000 s pulses with this LH launcher, by coupling routinely > 3 MW of LH wave at 3.7 GHz to the plasma with a parallel index n(//)=1.7(+/- 0.2). The launcher is on its way to achieve its validation tests - low power Radio Frequency (R-F) measurements, vacuum and hydraulic leak tests - and will be installed and commissioned on plasma during the fall of 2009.

This work concerns the development of an ITER relevant advanced robotic systems for fusion reactor. The feasibility demonstration will be performed on the thermonuclear experimental Tokamak: Tore Supra, located in Cadarache facilities. The manipulator developed by the interactive robotics unit of CEA-LIST will be used for close inspection intervention tasks in a Tokamak. The robot must meet severe specifications: small diameter, long reach with a 9.5 m cantilever length, high dexterity to move in a Torus shape environment and able to carry a 10 kg payload on its end effector. As the robot must be introduced without breaking the machine conditioning, it has to cope with the vacuum and temperature constraints: Ultra High Vacuum (10 -6 Pa) and 120degC in use, 200degC during baking phase. This long reach multi-link carrier has 10 DOF (9 rotary joints and 1 prismatic joint). It is composed of 5 modules and a precise guiding and pushing system (the deployer) for a total weight of about 300 kg. The gravity effect in the manipulator is largely compensated by a special mechanical structure (the parallelogram) that helps reducing the size of the actuators. The severe operating conditions impose a selection of several vacuum and temperature technologies that have been tested, qualified or optimized to cope with the requirements.In September 2007, the Remote Handling Equipment deployment under atmospheric conditions in a real Tokamak environment (Tore Supra) enabled to qualify the entire prototype. The scenario has to be rehearsed under vacuum and temperature to achieve the complete feasibility demonstration of a Tokamak close inspection by means of a robotic equipment under real operating conditions. The demonstration is an important step in the project but further developments are or could be necessary to cope with 100% of the requirements toward a reliable industrial prototype. In particular, due to its size and weight, this large robot manipulator holds lots of elastic and geometric deformations. Thus its accuracy is a challenge. A mechanical model could be developed to take into consideration the flexibilities of the structure and by means of calibration, the model parameters could be identified and then integrated in a real time controller. Further advanced developments on the on line monitoring system are also necessary for operator assistance and fault detection.