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Introduction Takotsubo syndrome (TTS) is an important form of acute heart failure with significant risk of acute complications and death. In this analysis we sought to identify predictors for in-hospital clinical outcome in TTS patients by concentrating on routine laboratory parameters at admission. Methods In this analysis from the Austrian national TTS registry, univariable and multivariable analyses were performed to identify significant predictors for severe in-hospital complications requiring immediate invasive treatment or leading to irreversible damage, such as cardiogenic shock, intubation, stroke, arrhythmias and death. Furthermore, the influence of identified predictors with long-term survival was evaluated. Results A total of 338 patients (median age 72 years, 86.9% female) from 6 centres were included. Severe in-hospital complications occurred in 14.5% of patients, including cardiogenic shock (9.8%), death (3.3%) and intubation (1.2%), respectively. Patients with complications during the hospital stay had more prevalent chronic kidney disease (CKD), were less often previous smokers and TTS was less often preceded by an emotional trigger. C-reactive protein and neutrophile lymphocyte ratio (NLR) was higher in patients with complications, and midventricular ballooning and reduced left ventricular ejection fraction (LVEF) were more prevalent. In multivariable analysis, high NLR (OR 1.04 [95% CI 1.02–1.07], p=0.009) and low LVEF (OR 0.92 [0.90–0.95] per %, p Discussion Low LVEF and high NLR at admission were independently associated with increased in-hospital complications and reduced long-term survival in TTS patients. NLR is a new easy-to-measure tool to predict worse short and long-term outcome after TTS. Funding Acknowledgement Type of funding sources: None.

WEST is an upgrade of the Tore Supra tokamak to test ITER like divertor elements over very long pulses. For the studies of MHD instabilities affected by tungsten impurities, two microwave imaging diagnostics have been developed to obtain density and temperature fluctuation maps. The core reflectometer was refurbished and now measures the core density profiles in few microseconds. Measurements of several thousand profiles in sequence with microsecond dead time (burst mode) will be available in 2019 enabling the reconstruction of density maps with unprecedented time resolution. The major issue for the implementation of an electron cyclotron emission imaging diagnostic was the heat radiation during long pulses on the large vacuum window and the first optics. Two metallic in-vessel mirrors inside the personnel access tube solved this issue. The first mirror can withstand the heat radiation without water cooling, the second mirror redirects the beam toward the personnel access port. The oblique and small view angle of this access eases also the thermomechanical constraints on the vacuum window. As the space between the port flange and the tokamak access lobby was tight, a compact optical enclosure with vertical aligned optics was designed. To ensure a good reproducibility of the mirrors and the optical enclosure positions after a maintenance period and thus avoid re-alignment, different mechanical systems are used to manipulate the mirrors and the optical enclosure and to fine tune their position and their orientation. The first channel covers the low field side and will provide 192 pixel images of temperature fluctuations with 1–2 microsecond time resolution. The installation of the in-vessel mirrors is scheduled during March 2019. Diagnostic qualification tests will be done before the restart of the experiments late May 2019.

The loss of calcium oxide CO2 capture capacity and the extent of particle attrition over iterated cycles are relevant to the design of Calcium Looping processes. Thermogravimetric analyzers or single batch fluidized bed reactors are typically used at the lab-scale to evaluate the sorbent performance. One drawback of these devices is that they do not reproduce the thermal history that is actually experienced by sorbent particles in real looping cycles. In this study, a novel experimental device is proposed to overcome this limitation. The test reactor has been used to assess the effect of the thermal history of CaO particles on their CO2 capture capacity and attrition tendency. The results were compared to those previously obtained with the same sorbent under comparable operating conditions in a single bed apparatus. The comparison reveals that the thermal history experienced by the CaO has a non-negligible effect on its performance.

This paper presents the current status of the WEST TWA antenna, its mock-up and a possible extrapolation to DEMO. The updated WEST TWA design has a reduced antenna length and features feeding and mechanical support from a single vessel port. A mock-up of the WEST TWA antenna was designed in 2019, manufactured during 2020 and installed in the TITAN test facility at the beginning of 2021. The results of the mock-up at low and high power, its diagnostic system and the prospects are explained. Extensions towards a TWA antenna for WEST and a possible TWA system for the future DEMO tokamak reactor are briefly discussed.

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 °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 procedure assisted 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.

An Electron Cyclotron Emission Imaging (ECEI) diagnostic system for the WEST tokamak has been developed under the UNIST-WEST collaboration. This diagnostic system is designed to overcome accessibility and thermomechanical constraints for long pulse operation. The first O-mode channel will be installed in the first trimester of 2019 to probe the low field side (LFS) of the WEST plasma. Two large metallic reflective mirrors are installed inside the duct which is being used for maintenance access. They are suspended on a rail to facilitate mirror manipulation. The ex-vessel optical system (lens, detection array, etc.) is housed in a compact optical enclosure that fits in a tight free space between the port flange and tokamak access lobby. The design emphasized reproducibility of the precise alignment between in-vessel mirrors and optical enclosure since the both elements must be removed during shutdown period for maintenance access. The overall optical system was fully tested at UNIST last year. The test results demonstrated that the imaging optics can full access at any radial position on the LFS. The 2D beam pattern measurements were consistent with the design values.

Fusion Tokamaks are complex machines which require special conditions for their operation, in particular, high vacuum inside the vessel and high temperature of the vessel walls. During plasma phases, the first wall components are highly stressed and a control is necessary in case of doubt about their condition. To be able to make safely such an inspection in a short period of time is a great advantage. The Articulated Inspection Arm (AIA) developed by the CEA provides the capability for fast inspections of the first wall overall surface keeping the vacuum and temperature conditions of the vessel. The robot prototype was validated in Tore Supra in 2008. In the frame of a joint laboratory, CEA/IRFM and ASIPP have decided to upgrade the existing AIA prototype to use it routinely in the EAST and WEST tokamaks. The robot has followed an important upgrade program in 2013 and 2014. The document presents the various upgrades made on the mechanics, the sensors, the electronics, the control station and the integration adaptation for the operation on EAST. From the AIA experience, thoughts for future inspection robots are given.

An articulated inspection arm (AIA) has been developed by CEA for visual inspection between pulses inside the Tore Supra tokamak vacuum vessel without breaking temperature and vacuum conditions. The eight meters length robot is composed of a shuttle and six articulated segments with a video camera at its end. A demonstration prototype has been achieved in 2008 at Tore Supra (Gargiulo, 2007; Houry, 2008; Perrot, 2003). A project to upgrade the AIA into a fully operational robot has been undertaken by IRFM and ASIPP in an Associated Laboratory. It will be in operation first in the EAST machine and afterwards in Tore Supra in its WEST (W/Tungsten Environment in Steady-state Tokamak) configuration where it is of paramount importance to survey possible degradation of W component surface. The control system of the robot has been extensively upgraded. The effort has been focused on three areas: (1) improvement of the arm position accuracy, (2) increase of the operational robustness, (3) use of a powerful graphical user interface including simulation of trajectories and robot deployment capabilities in a 3D viewer environment. The aim of this paper is to detail the architecture of the AIA control system.

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 °C, 4 MPa) in which a total divertor current of up to 16 × 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 °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.

The WEST-project ( W -tungsten E nvironment in S teady-state T okamak) involves equipping Tore Supra with a full tungsten divertor, capable of withstanding heat load of 10 MW/m2 in steady-state conditions, in discharges sustained by Lower Hybrid Current Drive (LHCD). The LHCD generator, recently upgraded to deliver 9.2 MW/1000 s, is equipped with sixteen TH2103C klystrons powering two launchers. The WEST transformation involves reducing the plasma volume, thus moving the launchers ∼10 cm closer to the tokamak centre. The toroidal curvature of the launchers no longer fits the plasma curvature due to the strong magnetic field ripple effect, leading to a degradation of the LH wave coupling, especially with the Full Active Multijunction Launcher (FAM). The toroidal curvature radius of the FAM launcher mouth will therefore be reshaped from 1700 mm to 2300 mm. The machining process is described in this article. In order to improve the coupling of the LH wave, the local gas injection has been modified to help to meet the requirement of 7 MW/1000 s of LH power coupled to the plasma in the WEST scenarios. Finally, the curvature radius of the waveguide septa are rounded to minimize the excitation of suprathermal electrons near the plasma edge, which can induce high power loads on the plasma facing components.

The WEST project (Tungsten (W) Environment in Steady state Tokamak) is an upgrade of the existing fusion machine, Tore Supra. The goal is to equip the tokamak with a fully cooled tungsten divertor and to transform the machine in a test platform open to all ITER partners. The main assembly challenge of this project consists of an implementation of two magnet systems, called divertors, with an accuracy of 1 mm. Indeed, each divertor has about 4 m as diameter and has a heavy weight of 10 tons; also it introduces piece by piece in the original vessel through tight ports then assembled inside. To ensure a perfect fitting between these new components and a very constrained environment, it is necessary to use the latest CAD technologies available. Beyond conventional CAD tools, the virtual reality (VR) room of the institute provides several useful tools. Thanks to the 185″ stereoscopic 3D screen and a force feedback arm linked to clash detection software developed by the CEA LIST, a new way to carry out design and assembly studies was performed. In order to improve VR results, metrology data (3D scan) enhance simulations. Therefore, it becomes possible to be aware of the real size of a component and future difficulties in assembling it. At last, performance of such simulations is evaluated and compared to physical mockup in order to bring enhancement to the VR tools, before to be compared to the real operations on Tore Supra. The aim is to build a design tool that helps the designer since early stage of the design of complex systems, taking into consideration integration, assembly and maintenance aspects while reducing costs and schedule of a project.

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 °C) and two pressurization tests are performed (6 MPa at room temperature and 4 MPa at 200 °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 × 10 −11 Pa m 3 s −1 , the pressure in the test tank being −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 ∼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.

On WEST new components will be implemented in an existing environment, emphasis has to be put on the metrology to optimize the design and the assembly. Hence, at a particular stage of the project, several components have to coexist in the limited vessel. Therefore, all the difficulty consists in validating the mechanical interfaces between existing components and new one; minimize the risk of the assembling and to maximize the plasma volume. The CEA/IRFM takes the opportunity of the ambitious project to sign a partnership with an industrial specialized in multipurpose metrology domains. To optimize the assembly procedure, the IRFM Assembly group works in strong collaboration with its industrial, to define and plan the campaigns of metrology. The paper will illustrate the organization, methods and results of the dedicated metrology campaigns have been defined and carried out in the WEST dis/assembly phase. To conclude, the future needs of metrology at CEA/IRFM will be exposed to define the next steps.

The Tore Supra tokamak is being transformed in an X-point divertor fusion device in the frame of the WEST project, launched in support to the Iter tungsten divertor strategy. The WEST Baffle aims to evacuate particles by creating a toroidal pumping throat. It also protects the lower divertor cooling pipes and the passive stabilization plate from heat fluxes. The Baffle is made of actively cooled plasma facing components (PFCs), with underneath a passive stabilization plate and a support beam. The Baffle design is presented in a first part. In a second time the different ANSYS simulations are described: electromagnetic, thermal and mechanical results are presented. The design choices are justified for the different results obtained thanks to the simulation. In a final part, some aspects of the PFC manufacturing are presented.

This work explores the effect of biomass templating on fragmentation of calcium-aluminate based pellets. Three different types of pellets have been tested: one with calcium aluminate cement (LC), another with flour (LF) and one with both cement and flour (LCF). Fragmentation has been investigated by two different techniques, namely with a pressurized heated strip reactor (PHSR) and a bubbling fluidized bed (BFB), under different conditions. Both sets of experiments showed that that the addition of biomass enhances the propensity to undergo fragmentation. The addition of cement partially counteracts this effect.

The anodic oxidation of caffeine and theophylline has been carried out at a controlled potential in a one compartment cell using glassy-carbon as the anode, platinum as the cathode and glacial acetic acid/sodium acetate as solvent-supporting-electrolyte (SSE). The oxidation of caffeine yields 1,3,8-trimethyl-1,3,6,8-tetrazaspiro[4.4]nonane-2,4,7,9-tetrone and (4-acetoxy-1,3,7-trimethyl-2,6,8-trioxo-9H-purin-5-yl) acetate. Despite of the different structure, the oxidation of theophylline leads only to the same spiro compound 1,3,8-trimethyl-1,3,6,8-tetrazaspiro[4.4]nonane-2,4,7,9-tetrone as in the case of caffeine. The structure of both compounds is described and the reaction mechanism is discussed.

The WEST project recently launched at Cadarache consists in transforming Tore Supra in an X-point divertor configuration while extending its long pulse capability, in order to test the ITER divertor technology. The implementation of a full tungsten actively cooled divertor with plasma facing unit representative of ITER divertor targets will allow addressing risks both in terms of industrial-scale manufacturing and operation of such components. Relevant plasma scenarios are foreseen for extensive testing under high heat load in the 10–20 MW/m2 range and ITER-like fluences (1000 s pulses). Plasma facing unit monitoring and development of protection strategies will be key elements of the WEST program. WEST is scheduled to enter into operation in 2016, and will provide a key facility to prepare and be prepared for ITER.

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.

Fusion machine operation requires high-vacuum conditions and does not tolerate water or gas leak in the vacuum vessels, even if they are micrometric. Tore Supra, as a fully actively cooled tokamak, has got a large leak management experience; 34 water leaks occurred since the beginning of its operation in 1988. To handle this issue, after preliminary machine protection phases, the current process for leak localization is based on water or helium pressurization network by network. It generally allows the identification of a set of components where the leakage element is located. However, the unique background of CEA-IRFM laboratory points needs of accuracy and promptness out in the leak localization process. Moreover, in-vessel interventions have to be performed trying to minimize time and risks for the persons. They are linked to access conditions, radioactivity, tracer gas high pressure and vessel conditioning. Remote operation will be one of the ways to improve these points on future fusion machines. In this case, leak sensors would have to be light weight devices in order to be integrated on a carrier or to be located outside with a sniffing process set up. A leak localization program is on-going at CEA-IRFM Laboratory with the first goal of identifying and characterizing relevant concepts to localize helium or water leaks on ITER. In the same time, CEA has developed robotic carrier for effective in-vessel intervention in a hostile environment. Three major tests campaigns with the goal to identify leak sensors have been achieved on several CEA test-beds since 2010. Very promising results have been obtained: relevant scenario of leak localization performed, concepts tested in a high volume test-bed called TITAN, and, in several conditions of pressure and temperature (ultrahigh vacuum to atmospheric pressure and 20–120 °C), “ppm” traces of helium or water have been measured. To improve these first promising results, a new test-bed called ROVE (remote operation vacuum equipment) has been set up in 2012 to qualify leak sensors and remote operation. A list of concepts has been set out from contacts with manufacturers and laboratories. For sniffing solutions it includes laser absorption measurement, mass spectrometer analyses and vibrating quartz. For embeddable concepts thermal conductivity and capacitive sensors are considered. Choice of concepts to detect water or helium traces is based on major criteria, such as weight, high sensibility and fast response time and recovery time. They also have to be operational under primary vacuum conditions or atmospheric pressure and from ambient temperature to 60 °C.

Understanding the mobilization of dust in tokamaks is essential for proper machine operation and occupational worker exposure during maintenance operations. It is also a key issue for the safety assessment of ITER in Loss of Vacuum Accident (LOVA) scenario. Dust mobilization by airflow is strongly dependent on particle size and shape, and these characteristics are related to the specific process of erosion of the plasma facing components. Accordingly, it is important to develop techniques for in situ analysis of wall dustiness and potential dust mobilization by airflow. We propose a device, called DUSTER BOX, allowing to study dustiness of Tore Supra surfaces and to collect dust sample without mechanical stresses. This device consists in a simple airflow channel box and allows injecting a filtered airflow tangentially to the surface. The flow velocity parallel to the plasma facing components is 10 m/s and the friction velocity is 0.7 m/s which is the value needed to tear off particles greater than few microns from surfaces. After being resuspended, carbon dusts are collected on a membrane filter for weight analysis, size distribution analysis and electron microscope imaging. DUSTER BOX has been successfully used to realize a mapping of particles deposit in Tore Supra and to collect samples for size distribution analysis.

The anodic oxidation of caffeine and theophylline has been carried out at a controlled potential in a one compartment cell using glassy-carbon as the anode, platinum as the cathode and glacial acetic acid/sodium acetate as solvent-supporting-electrolyte (SSE). The oxidation of caffeine yields 1,3,8-trimethyl-1,3,6,8-tetrazaspiro[4.4]nonane-2,4,7,9-tetrone and (4-acetoxy-1,3,7-trimethyl-2,6,8-trioxo-9H-purin-5-yl) acetate. Despite of the different structure, the oxidation of theophylline leads only to the same spiro compound 1,3,8-trimethyl-1,3,6,8-tetrazaspiro[4.4]nonane-2,4,7,9-tetrone as in the case of caffeine. The structure of both compounds is described and the reaction mechanism is discussed.

The Interactive Robotics Laboratory of CEA LIST is in charge of the development of remote handling technologies to meet energy industry requirements. This paper reports the research and development activities in advanced robotics systems for inspection or light intervention in hazardous environments with limited access such as blind hot cells in the nuclear industry or the thermonuclear experimental Tokamak fusion reactor. A long-reach carrier robot called the articulated inspection arm (AIA) and diagnostics and tools for inspection or intervention are described. Finally experimental field tests are presented and actual challenges in modeling the robot's flexibilities are discussed. © 2011 Wiley Periodicals, Inc.

ITER will greatly rely on remote-handling operations to accomplish its scientific missions. Robotic systems will also be required to operate inside vacuum vessels in order to limit or replace human access, to intervene quickly between experimental sessions for in-vessel inspections and measurements, and to preserve the machine conditioning and thus improve machine availability. In this prospect, a multipurpose carrier prototype called Articulated Inspection Arm (AIA) was developed by CEA laboratories within the European work program. With an embedded camera, it successfully demonstrated close inspection feasibility inside Tore Supra tokamak. The AIA robot was designed for mini-invasive operations with interchangeable diagnostics to be plugged at its head. This covers various applications for the safety, the operation and the scientific mission (in-vessel inspection, plasma diagnostics calibrations or inner components analysis and treatments). This paper presents recent analysis and results obtain with diagnostics developed by CEA for in-vessel remote-handling intervention.

In order to reduce the risks for ITER Plasma Facing Components (PFCs), it is proposed to equip Tore Supra with a full tungsten divertor, benefitting from the unique long pulse capabilities, the high installed RF power and the long experience with actively cooled high heat flux components of the Tore Supra platform. The transformation from the current circular limiter geometry to the required X-point configuration will be achieved by installing a set of copper poloidal coils inside the vacuum vessel. The new configuration will allow for H-mode access, providing relevant plasma conditions for PFC technology validation. Furthermore, attractive steady-state regimes are expected to be achievable. The lower divertor target design will be closely based on that currently envisaged for ITER (W monoblocks), while the upper divertor region will be used to qualify the main first wall heat sink technology adopted for the ITER blanket modules (CuCrZr copper/stainless steel) with a tungsten coating (in place of the Be tiles which ITER will use). Extended plasma exposure will provide access to ITER critical issues such as PFC lifetime (melting, cracking, etc.), tokamak operation on damaged metallic surfaces, real time heat flux control through PFC monitoring, fuel retention and dust production.

A multipurpose carrier prototype called Articulated Inspection Arm (AIA) has been developed and demonstrated by CEA laboratories to operate embedded diagnostics or tools into a tokamak vacuum-vessel. The AIA robot with a vision diagnostic has been deployed inside the vacuum-vessel of Tore Supra tokamak and has realized closed inspections of plasma facing components’ under relevant conditions (high vacuum and temperature). CEA is engaged in development of interchangeable diagnostics or tools to be plugged on the front head of the arm. This covers applications such as in-vessel inspections, water leak localization, Plasma Facing Components (PFCs) characterization and treatment. This paper describes R&D activities performed in CEA laboratories to develop diagnostics and tools for in-vessel interventions. The AIA multipurpose carrier to carry such devices is also presented. The first demonstration of a quick intervention for plasma facing components performed in the Tore Supra Facility is described as well.

In order to demonstrate the feasibility of a mini-invasive inspection arm in ITER, the next step in fusion energy research machines, and to match the need for such a device in the French tokamak Tore Supra, a hardened long-reach robot has been developed by the CEA-LIST robotics team. The AIA (Articulated Inspection Arm), a 8 m long, highly articulated dexterous carrier with a 10 kg payload has been developed to be deployed anywhere in the Tore Supra vacuum vessel from one single O250 mm entry point, while the vessel is kept at the operating temperature (120) and pressure . Since its successful demonstration in Tore Supra in September 2008, the robot has been identified by different teams of scientists working on the tokamak for various potential tasks, from vision to diagnostic calibration or component analysis. In order to make such a device even more available, the CEALIST team is now working on the magnetic field constraint on robots. The goal being to demonstrate capabilities of intervention in ITER conditions, the considered field is 4 to 8 T. Technologies, command schemes and design strategies available for an inspection robot tolerant to high magnetic field have been studied. This paper first details the different technologies chosen for the AIA. How such an inspection device has been integrated on a multidisciplinary research machine such as Tore Supra is then detailed in a second part. The paper finally presents various concepts and experimentations to be used in a magnetic field tolerant robot for inspection of a large superconductor machine.

In 2008, the Articulated Inspection Arm (AIA) performed its first deployment in Tore Supra tokamak vessel, under real vacuum and temperature conditions (10−6 Pa and 120 °C) after a conditioning phase to avoid pollution of the chamber. This full demonstration is a turning point in the project, allowing a second phase to start toward the objective to make use of the Remote Handling Equipment as an inspection routine tool. This feasibility demonstration will also be performed on Tore Supra. At this stage, the system requires enhancements of overall technologies and also developments of the Remote Handling control system. Lessons learned in the first phase of the project and experience collected for more than 20 years by Robotics Units of CEA-LIST teams on teleoperated systems enable to identify three main fields to explore to reach the objectives: • Development of a reliable control based on a real time generic command control system. • Development of a graphical supervisor for intuitive steering. • Development of a generic flexible model for realistic computations of the robot localization in the environment. This paper is dedicated to the description of the further developments that shall be done for routine Remote Handling operations. It presents the R&D project roadmap for the next years.

Robotic operations are one of the major maintenance challenges for ITER and future fusion reactors. CEA has developed a multipurpose carrier able to realize deployments in the plasma vessel without breaking the Ultra High Vacuum (UHV) and temperature conditioning. A 6 years RD its geometry allows a complete close inspection of Plasma Facing Components (PFCs) of the Tore Supra vessel. Different tools are being developed by CEA to be plugged at the front head of the carrier. The diagnostic presently in operation consists in a viewing system offering accurate visual inspection of PFCs. Leak detection of first wall based on helium sniffing and laser compact system for carbon co-deposited layers characterizations or treatments are also considered for demonstration. In April 2008, the AIA robot equipped with its vision diagnostic has realized a complete deployment into Tore Supra and the first closed inspection of the vessel under UHV conditions. During the upcoming experimental campaign, the same operation will be performed under relevant conditions (10 −6 Pa and 120 °C) after a conditioning phase at 200 °C to avoid outgassing pollution of the chamber. This paper describes the different steps of the project development, robot capabilities with the present operations conducted on Tore Supra and future requirements for making the robot a tool for tokamak routine operation.

During Tokamak operation, Plasma Facing Components (PFCs) are subjected to severe interaction with plasma. As a consequence and independently of the PFCs composition, materials eroded and then re-deposited in the form of layers on the surfaces, can flake and produce dusts. These fragile structures are able to trap part of the hydrogenated species (tritium for example) in vessel inventory. For safety reasons, it is mandatory to measure and to control vessel dust and tritium inventory. Up to now, laser techniques are a part of the most promising methods able to solve these ITER open issues. Of special interest are laser systems loaded on a miniature tool that can be attached to a Multi Purpose Deployer (MPD) and used for laser treatments (detritiation and other), for PFCs chemical analysis as well as for micro particles recovery of dust produced during laser ablation. Such a system (Laser Ablation System Kit: LASK) is currently under development at IRFM and the following presentation will describe the current achievements of this project and the perspectives. In this paper, we will present an innovative compact system, which, loaded on a Multi Purpose Deployer, could allow operation in a harsh environment (pressure range from atmospheric to Ultra High Vacuum and temperature up to 120 °C). According to the process conditions, different treatments can be performed: at low laser fluence, PFCs thermal treatment will be expected, while at high laser fluence material will be ablated allowing Dust (and T) recovery as well as chemical analysis of material. This “in-line” chemical analysis based on Laser Induced Breakdown Spectroscopy (LIBS) enables the ablation process to be controlled and preserves the substrate integrity. The paper will be focussed on the methodology followed during the LASK development and the method used to determine a laser process window able to remove co-deposited film without damaging the bulk material and taking into account external parameter variation (Multi Purpose Deployer vibrations for example). The first design of the system is proposed that complies with the process requirements and the external constraints. Special emphasis will be given on limitations, and alternatives to these limitations will be proposed.

The aim of the project is to demonstrate on Tore Supra the reliability of a multi-purpose in-vessel remote handling inspection system using a long reach, limited payload carrier. The robot prototype is fully representative of the deployment carrier system that could be required on ITER. The demonstration on Tore Supra will help in the understanding of operation issues that could occur in the tokamak vacuum vessel equipped of actively cooled components. The viewing process that is currently under development will allow close inspection of the Tore Supra plasma facing components that are representative of the ITER divertor targets in terms of confined environment and identification of possible tiles failure of CFC carbon tiles. One of the other potential inspection processes that is foreseen to be tested using the AIA carrier in Tore Supra is the laser ablation system of the CFC armour. It could be fully relevant for the ITER wall detritiation issues. Such process can be simulated on Tore Supra through the deuterium inventory under long-time plasma discharges. The in situ leakage localisation of a damaged plasma facing component is also one of the major ITER maintenance challenges that could use remote handling inspection tools.

The Tore Supra tokamak is a fusion device which has been in operation since 1988. Its main features are its superconducting toroidal magnet and its actively cooled first wall. Tore Supra is devoted to the study of physics and technology of long plasma discharges. Its actively cooled first wall is able to remove the energy deposited from the plasma during a very long pulse regime (several hundred seconds). A major upgrade of the Tore Supra in-vessel components, the CIEL project (French Acronym of Composants Internes Et Limiteur), has been implemented and is in operation since 2001. This paper describes the experience gained from the manufacturing of cast stainless steel hydraulic structures supporting the toroidal pump limiter, the main component of the CIEL project, as well as the technical description of this structure, the associated specifications, the difficulties encountered during manufacture and the repair possibilities.

Closed photobioreactors have to be optimized in terms of light utilization and overall photosynthesis rate. A simple model coupling the hydrodynamics and the photosynthesis kinetics has been proposed to analyze the photosynthesis dynamics due to the continuous shuttle of microalgae between dark and lighted zones of the photobioreactor. Microalgal motion has been described according to a stochastic Lagrangian approach adopting the turbulence model suitable for the photobioreactor configuration (single vs. two-phase flows). Effects of light path, biomass concentration, turbulence level and irradiance have been reported in terms of overall photosynthesis rate. Different irradiation strategies (internal, lateral and rounding) and several photobioreactor configurations (flat, tubular, bubble column, airlift) have been investigated. Photobioreactor configurations and the operating conditions to maximize the photosynthesis rate have been pointed out. Results confirmed and explained the common experimental observation that high concentrated cultures are not photoinhibited at high irradiance level.

In the frame of the Tore Supra Composants d’Injection de Matiere et d’Energie Stationnaire (CIMES) Project, consisting in particle fuelling via a pellet injector and in energy powering using Lower-Hybrid Current Drive (LHCD) for pulses lasting 1000 s, complemented by the transmitter refurbishing, a new launcher, based on the Passive-Active Multijunction (PAM) concept already proposed for ITER, has been developed and is currently under realisation. It was designed for a power injection capability of 2.7 MW CW at 3.7 GHz, a power density of 25 MW/m 2 , to radiate a power spectrum peaked at N // = 1.7 with a maximum power directivity near the electron cut-off density and with very good coupling properties on a wide range of electron densities. In this paper, an overall description of the antenna as well as the foreseen manufacturing and assembling processes are given, followed by the results from studies and optimisations of its RF components, by a stress analysis of its thermo-mechanical behaviour, taking into account the plasma radiated flux at the mouth and the RF losses, and complemented with additional mechanical stresses due to the combined action of the eddy current induced in the launcher by disruptions and the residual toroidal magnetic field.

This paper is part of the remote handling (RH) activities for the future fusion reactor ITER. The aim of the R&D program performed under the European Fusion Development Agreement (EFDA) work program is to demonstrate the feasibility of close inspection tasks such as viewing or leak testing of the Divertor cassettes and the Vacuum Vessel (VV) first wall of ITER. It is assumed that a long reach, limited payload carrier penetrates the ITER chamber through the openings evenly distributed around the machine such as In-Vessel Viewing System (IVVS) access or through upper port plugs. To perform an intervention a short time after plasma shut down, the operation of the robot should be realised under ITER conditioning i.e. under high vacuum and temperature conditions (120 °C). The feasibility analysis drove the design of the so-called articulated inspection arm (AIA) which is a 8.2 m long robot made of five modules with a 11 actuated joints kinematics. A single module prototype was designed in detail and manufactured to be tested under ITER realistic conditions at CEA-Cadarache test facility. As well as demonstrating the potential for the application of an AIA type device in ITER, this program is also dedicated to explore the necessary robotic technologies required to ITER's IVVS deployment system. This paper presents the whole AIA robot concept, the first results of the test campaign on the prototype vacuum and temperature demonstrator module.

ITER will be the most important machine equipped with actively cooled plasma facing components (PFCs). In case of abnormal events during a discharge, the PFC will be submitted to localized transient phenomena (high power densities, run away electrons, etc.), leading, in the worst case, to the degradation of the PFC wall and possibly to a water leak. In any case, a leak will have important consequences for the PFCs and equipment located in the vacuum vessel or connected to the ports such as seals, pumping systems or diagnostics. Considerable experience of these events has been gained at Tore Supra over a period of more than 10 years [J.J. Cordier, Ten years of maintenance on Tore Supra actively cooled components, in: Proceedings of the 21th Symp. of Fusion Technology (SOFT), Madrid, Spain, September, 2000.], which will be useful for the next step machines. This paper describes for each leak size type the procedures for maintaining save conditions in the vacuum vessel. It also presents the methods used at Tore Supra to drain-off the primary loop circuits and to identify the leaky PFC.

Coarse fragments influence many soil physical properties. Available studies that address the controversial role of rock fragments in soil functioning are often oriented to assess effects of their presence in relation to specific applicative problems like soil erosion or their influence in agricultural practices. Experimental research is mostly based on the evaluation of the effect of rock fragments cover on soil erosion and of coarse fragment content on water infiltration. Much less is known on the role of stones on the development of soil surface cracks, which in turn influence many other soil properties including thermal conductivity and preferential flows. Surface cracks are especially important in Vertisols where cracks occurrence, width and depth constitute diagnostic criteria for Vertisol identification and classification (e.g. WRB, Soil Taxonomy). This is because soil cracks largely affect land management. In this work we designed a lab experiment in order to evaluate the effect of rock fragments on surface cracks development of a Vertisol. Two concentrations of three different sizes of rock fragments (basaltic type) were mixed with the soil material after sieving and put in flowerpots having diameter of 15cm; then such samples have undergone nine wet/dry cycles. At the end digital images of sample surface were acquired and examined by image analysis. In particular the cracking density was measured by a skeletonization process and the mean width of cracks was determined through a mathematical morphology approach. Results showed that the surface cracking density increased with rock fragments concentration and decreased with the rock fragment size while the mean width of fractures increased with rock fragment size and decreased with rock fragments concentration.

Tore Supra is the only superconducting Tokamak equipped to actively cooled plasma facing components. The CIEL project (composants internes et limiteurs) consists of a complete major upgrade of all the existing internal components (Fusion Engineering and Design 56–57 (2001) 117). This new device should enable removing 15 MW of convected flux on the toroidal pump limiter (TPL) and should sustain high radiated plasma power losses of at least 10 MW on the first wall during pulse length up to 1000 s. The associated primary pressurized water loop (3 MPa at 150 °C) has also been upgraded in order to be able to exhaust the injected power in a steady state configuration. The paper describes assembly operations during the fabrication and setting-up of plasma facing components as part of the CIEL project during the 2000–2002 shutdowns.