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Congenital mesoblastic nephroma (CMN) is the most frequent renal tumor in newborns and infants. under 3 months. The clinical case of a patient younger than 3 months, with prenatal diagnosis, referred in a timely manner for evaluation and management by the Institution is presented. CMN is a low incidence tumor. Early diagnosis as well as complete excision of the tumor are predictors of good prognosis, as in the case of our patient.

International audience; Limiter objects are facing high loads, even if they are not as close to the plasma as divertor for example. Loads come fromplasma interaction during the standard situation (heat loads) and during abnormal situation like disruption (electromagneticloads, runaway electrons). In WEST, the initial guess was to use CFC tiles with a tungsten coating for the outer bumperlimiter. After two campaigns, it appears that the power and quantity of runaway beams had been underestimated, leadingto early damages of the coating on significant surfaces. A second attempt for the tiles is to use bulk tungsten. This shouldensure a high resistance the runaway beams. The design is validated for the given specifications. The tiles aremanufactured in early 2019, and used during the WEST operations later this year. The first observations show a goodresistance to the runaway beam, as expected.

Five endoscopes equipped with infrared cameras working in the medium infrared range (3–5 μ m) are installed on the controlled thermonuclear fusion research device Tore-Supra. These endoscopes aim at monitoring the plasma facing components surface temperature to prevent their overheating. Signals delivered by infrared cameras through endoscopes are analysed and used on the one hand through a real time feedback control loop acting on the heating systems of the plasma to decrease plasma facing components surface temperatures when necessary, on the other hand for physics studies such as determination of the incoming heat flux . To ensure these two roles a very accurate knowledge of the absolute surface temperatures is mandatory. Consequently the infrared endoscopes must be calibrated through a very careful procedure. This means determining their transmission coefficients which is a delicate operation. Methods to calibrate infrared endoscopes during the shutdown period of the Tore-Supra machine will be presented. As they do not allow determining the possible transmittances evolution during operation an in-situ method is presented. It permits the validation of the calibration performed in laboratory as well as the monitoring of their evolution during machine operation. This is possible by the use of the endoscope shutter and a dedicated plasma scenario developed to heat it. Possible improvements of this method are briefly evoked.

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.

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 (W – for tungsten – Environment in Steady-state Tokamak) project is based on an upgrade of the Tore Supra tokamak, into a double X-point divertor device, while taking advantage of its long discharge capability. Therefore components with tungsten (W) as plasma-facing material will be used. This paper presents the upper divertor of the WEST project (design constraints, thermal performance). This component, with a total surface of 8 m2 is designed to exhaust 4 MW of conducted power in steady state with a maximum local heat load of 8 MW m−2. The actively cooled heat sink of this upper divertor target is made of CuCrZr. It is covered with a W coating up to 30 μm thickness. The suited thermal behaviour of the component was checked using finite element modelling. Moreover, under high heat flux tests this component had the expected thermal exhaust capability and survived without visible damage up to 10.5 MW m−2 in steady-state.

The microstructures and mechanical properties of electron beam welded CuCrZr/Inconel/316 L junctions for WEST project have been investigated. As the welding of ending tubes is performed after the hot isostatic pressing (HIP) for the real components, in this study the CuCrZr tubes were annealed with identical parameters of HIP before welding in order to simulate the real situations of component manufacturing. The junctions were free of defects such as micro-cracks or incomplete penetration. The mechanisms of macro-segregation features, such as beaches, peninsulas, and irregular islands, and of micro-segregation features, such as small spheres, dendrites, Cu-rich particles, and Nb/Mo-rich networks have been discussed. Mechanical testing shows that postweld aging nearly does not improve the tensile strength of the junctions because it just shifts the weakest point of CuCrZr slightly away from the Inconel/CuCrZr weld, but it is beneficial for the tensile elongation of the junctions at 200 °C; meanwhile, postweld aging decreases the hardness of CuCrZr in regions out of the heat affected zone. Cold helium testing performed after rotary fatigue testing indicated that postweld aging does not change the fatigue resistance of the junctions, at least under the current testing conditions. Significant increments in hardness of irregular Inconel-rich islands induced by aging may be attributable to the formation of Nb/Mo-rich networks. This study may be of significance for the manufacturing of ITER divertor components.

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.

The WEST (W – for tungsten – Environment in Steady-state Tokamak) project is an upgrade of Tore Supra from a limiter based tokamak with carbon PFCs into an X-point divertor tokamak with full-tungsten armour while keeping its long discharge capability. The WEST project will primarily offer the key capability of testing for the first time the ITER technology in real plasma environment. In particular, the main divertor (i.e. the lower divertor) of the WEST project will be based on actively cooled tungsten monoblock components and will follow as closely as possible the design and the assembling technology, foreseen for the ITER divertor units. The current design of WEST ITER-like tungsten divertor has now reached a mature stage following the 2013 WEST Final Design Review. This paper presents the key elements of the design, reports the technological requirements and reviews the main design and integration issues.

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.

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, 4 MPa). 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.

Among major issues for PFCs design, the impact of leading edges (exposed surface) which would be directly intersected by particles following magnetic field lines at glancing incident angles in the high heat flux areas is much discussed. This paper presents the key outcome of a thermal analysis performed on different shaping solutions for the ITER-like W monoblocks occurred for the components of the WEST ( W E nvironment for S teady state T okamak) divertor which could shadow any direct leading edge and to counteract a potential misalignment due to assembly tolerance. The results, in terms of surface temperature rise and wall heat flux into the cooling channel, are discussed for magnetic field lines incident at glancing angles expected in the higher heat flux regions of divertor (i.e. close to the strike point regions) and for perpendicular incident heat flux up to 20 MW/m2.

One goal of the WEST project (W Environment in Steady-state Tokamak) is the manufacturing, quality assessment and operation of ITER-like actively water-cooled divertor plasma facing components made of tungsten. Six W monoblock plasma facing units (PFUs) from different suppliers have been successfully evaluated in the high heat flux test facility GLADIS at IPP. Each PFU is equipped with 35 W monoblocks of an ITER-like geometry. However, the W blocks are made of different tungsten grades and the suppliers applied different bonding techniques between tungsten and the inserted Cu-alloy cooling tubes. The intention of the HHF test campaign was to assess the manufacturing quality of the PFUs on the basis of a statistical analysis of the surface temperature evolution of the individual W monoblocks during thermal loading with 100 cycles at 10 MW m−2. These tests confirm the non-destructive examinations performed by the manufacturer and CEA prior to the installation of the WEST platform, and no defects of the components were detected.

For the fabrication of 600 actively cooled finger elements for the Tore Supra pump limiter in operation since 2001 it was necessary to rely on two different batches of the CFC N11 grade (Carbon Fibre reinforced Composite) namely so-called SEP N11-92 (fabricated in 1992) and N11-98 (fabricated in 1998). It came out during the incoming inspection of the fingers that the bonding quality was degraded for the 98-batch so that an important number of tiles had to be repaired. Due to the coming upgrade of the Tore Supra heating system, two high heat flux test campaigns were performed on the neutral beam GLADIS facility (IPP Garching, Germany) including micro-structural analyses in order to evaluate, compare and understand the fatigue behaviour of 92- and 98-batch finger elements.

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.

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 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.

To fulfill the Tore Supra mission (the realization and study of high-performance long-duration discharges), the development of reliable actively cooled plasma-facing components is mandatory. This w...

Deuterium trapping in carbon fiber composite N11 and pyrolitic graphite PG99 irradiated with plasma ions and electrons was examined with thermal desorption spectrometry. It has been found that the deuterium trapping takes place even at ion and electron energies of about 10 eV. For equal ion fluences, the deuterium retention and probability of CD4 formation are higher for ion irradiation at lower ion flux. Peculiarities of the deuterium retention and CD4 formation are explained; driving forces and mechanisms of the D trapping are discussed.

Lock-in thermography has been used to control plasma facing components in Tore-Supra (TS) fusion device, during maintenance shutdown. The aim was to perform an in-situ health monitoring of high heat flux components after several years of plasma campaigns with a Non-Destructive Examination (NDE) method, to evaluate the components damage rate. This technique has been tested successfully in laboratory conditions, and provides a qualitative method to detect major defects on refractory armour tile/heat sink interface. It has been finally applied in-situ on the whole TS Toroidal Pump Limiter (TPL). It appears that a few elementary tiles are potentially damaged, which represent less than 1/1000 of the surface. The paper presents the method of this original health monitoring, and discusses the experimental results.

Tore supra (TS) in vessel components represent a unique combination of metals in the hydraulic circuit. Different materials, e.g. stainless steel, copper alloys, nickel, etc., were joined together by fusion welding, brazing and friction. Since the operation and baking temperature of all in vessel components has been defined to be set at 230 °C/40 bars a special water chemistry of the cooling water loop was suggested in order to prevent eventual water leaks due to corrosion at relative high temperatures and pressures in tubes, bellows, coils and coolant plant ancillary equipments. Following experiences with water chemistry in Pressurised Water Reactors, an all volatile chemical treatment (AVT) has been defined for the cooling water quality of TS. Since then, a simplified static (no fluid circulation) corrosion test program at relatively high temperature and pressure has been performed using capsule-type samples made of above mentioned multi-metal assemblies.

In this paper the results of post-mortem analysis of mirror samples (molybdenum, stainless steel and copper) exposed for the 1 year experimental campaign inside the Tore Supra tokamak are presented. The mirrors were not protected during the experiment and therefore were exposed both to glow discharge conditioning of the vessel walls and to working discharges. After exposure, all samples were found to be eroded and at the same time were coated with a thin contaminating deposit. We pay attention to the fact, that the observed erosion cannot be described using the published sputtering yields. In particular, the difference between the erosion of stainless steel and copper exceeded a factor of 10 and the difference between the erosion of stainless steel and molybdenum was less than a factor of 2 in comparison with the corresponding factors of approximately 2.5 and apprximately 10 known from the literature. An attempt is made to correlate these observations with the different sticking coefficients of carbon on ...

The primary goal of the WEST project is to be a test bed to characterize the fatigue and lifetime of ITER-like W divertor components subjected to relevant thermal loads. During the first phase of exploitation (S2 2016), these components (W monoblock plasma facing unit—W-PFU) will be installed in conjunction with graphite components (G-PFU). Since the G-PFU will not be actively cooled, it is necessary to ensure the expected pulse duration allows the W-PFU to reach its steady state without overheating the G-PFU assembly structure or the embedded stainless-steel diagnostics. High heat flux tests were performed at the GLADIS facility to assess the thermal behavior of the G-PFU. Some operational limits based on plasma parameters were determined. It was found that it is possible to operate at an injected power such that the maximal incident heat flux on the lower divertor is 10 MW m−2 for the required pulse length.

A dramatic exothermic reaction has been observed in Tore Supra with molten aluminium enclosed in a thermally loaded 2D carbon fibre reinforced carbon composite (CFC) tile, exposed after component removal to atmospheric moisture. This led to an important local destruction of the graphite matrix and fibre texture. Beryllium, as aluminium, a carbide forming metal, is foreseen for the ITER first wall armour. Since it can be expected that sputtered, molten and/or evaporated beryllium material will be transported to the hot CFC based divertor surface during normal and off-normal plasma operation, a preliminary reaction experiment has been performed in order to investigate to which extent the Be contaminated graphite matrix and fibre texture of 3D-CFC SEPCARB ® NB31 could be locally damaged (“erosion”) following a subsequent long lasting air venting or a supposed in-vessel ingress of coolant (steam) event (ICE).

On Tore Supra, five far infrared (FIR) retro-reflected channels have been installed to double the existing traversing ones, in order to increase the spatial resolution of the electron and current density measured by the interfero-polarimeter diagnostic. The new channel geometries are similar to those of the intended ITER polarimeter and have been designed for long pulse plasma operation. Inside the tokamak, the beams are reflected on edge-cooled retro-reflectors. During the 2003–2004 TS campaign, a prototype and reference plan mirrors had been exposed to plasma. After their removals, quantitative SIMS deposit analysis estimate the thickness of the carbon up to 700 nm and FIR measurements confirmed that the surface changes were too small to modify the reflectivity of the mirrors. The transport of the FIR beams by mirrors to the detectors is then discussed. Ray tracings through plasma and mirrors have been done to rightly design and position the windows, the detectors and the separation grids and minimize the misalignment effects. First results of both interferometric and polarimetric reflected channel signals show a good consistency with the traversing ones and allow resolving more accurately the profiles.

The uniform layer with thickness at least 50–100 μm was found on the CFC tiles from the inboard midplane after more than four years of tokamak operation. The upper part of the uniform layer was amorphous, but at the depth of ∼5 μm a structure consisting of micro-size regions with aromatic chains located parallel to the surface was found. Gradual transition from uniform layer to underlying CFC structure was observed. The reciprocating material probe was used for installation of CFC samples in the Tore Supra deuterium plasma. The thermal desorptional spectra of these samples are compared with the spectra of the samples irradiated in the laboratory stand and with the spectra of hydrogenated carbon film. The peculiarities of hydrogen isotope trapping under plasma irradiation and at the atmosphere are presented and discussed.

A number of designs for external or in-vessel coils generating Resonant Magnetic Perturbations (RMPs) for type I ELMs control in ITER are analyzed. The RMPs generated by the I-coils in the successful DIII-D ELMs control experiments are taken as a reference. The three ITER characteristic scenarios (H-mode, hybrid and steady-state) are studied. In a second part of the paper, the first results of a self-consistent non-linear MHD modelling of the plasma response to the RMPs with the JOREK code, for a DIII-D case, are presented.

Metal mirrors will be implemented in about half of the ITER diagnostics. Mirrors in ITER will have to withstand radiation loads, erosion by charge-exchange neutrals, deposition of impurities, particle implantation and neutron irradiation. It is believed that the optical properties of diagnostic mirrors will be primarily influenced by erosion and deposition. A solution is needed for optimal performance of mirrors in ITER throughout the entire lifetime of the machine. A multi-machine research on diagnostic mirrors is currently underway in fusion facilities at several institutions and laboratories worldwide. Among others, dedicated investigations of ITER-candidate mirror materials are ongoing in Tore-Supra, TEXTOR, DIII-D, TCV, T-10 and JET. Laboratory studies are underway at IPP Kharkov (Ukraine), Kurchatov Institute (Russia) and the University of Basel (Switzerland). An overview of current research on diagnostic mirrors along with an outlook on future investigations is the subject of this paper.

Objectives-The aim of this study was to determine the intra- and inter-rater reliability of sonographic measurements of the median nerve cross-sectional area in individuals with carpal tunnel syndrome and healthy control participants.Methods-The median nerve cross-sectional area was evaluated by sonography in 18 participants with carpal tunnel syndrome (18 upper extremities) and 9 control participants (18 upper extremities) at 2 visits 1 week apart. Two examiners, both blinded to the presence or absence of carpal tunnel syndrome, captured independent sonograms of the median nerve at the levels of the carpal tunnel inlet, pronator quadratus, and mid-forearm. The cross-sectional area was later measured by each examiner independently. Each also traced images that were captured by the other examiner.Results-Both the intra- and inter-rater reliability rates were highest for images taken at the carpal tunnel inlet (radiologist, r = 0.86; sonographer, r = 0.87; inter-rater, r = 0.95; all P < .0001), whereas they was lowest for the pronator quadratus (r = 0.49, 0.29, and 0.72, respectively; all P < .0001). At the mid-forearm, the intra-rater reliability was lower for both the radiologist and sonographer, whereas the inter-rater reliability was relatively high (r = 0.54, 0.55, and 0.81; all P < .0001). Tracing of captured images by different examiners showed high concordance for the median cross-sectional area at the carpal tunnel inlet (r = 0.96-0.98; P < .0001).Conclusions-The highest intra- and inter-rater reliability was found at the carpal tunnel inlet. The results also demonstrate that tracing of the median nerve cross-sectional area from captured images by different examiners does not contribute significantly to measurement variability.

21st International Conference on Plasma-Surface Interactions in Controlled Fusion Devices (PSI), Natl Inst Fus Sci, Kanazawa, JAPAN, MAY 26-30, 2014; International audience; Understanding heat flux deposition processes is essential for the design of the plasma facing component allowing reliable high power steady state plasma operations. Misalignments up to delta = 0.2 mm between two adjacent CFC tiles have been reported on the Toroidal Pump Limiter of Tore Supra. Heat flux impinging the top and leading edge of the protruding tile are characterized with both IR thermographic system and numerical modelling using 2D particle-in-cell simulations that accounts for the Larmor radius smoothing effect of incident ions. Numerical heat loads are coupled with a 2D thermal model of the tile and with a specific sensor correction to simulate spatial-resolution related effects (necessary here since the tile misalignment is smaller than the spatial resolution of the IR system). In the experiment depicted here, with a misalignment smaller than an ion gyro-radius, the Larmor radius smoothing effect is maximum and overheating of the leading edge is reduced by a factor of two. (C) 2014 Elsevier B.V. All rights reserved.

This paper presents our experience with precipitation hardened CuCrZr in the fabrication of actively cooled components for Tore Supra (toroidal pump limiter fingers, guard limiter heat sink hollow-profiles, ripple protection tubes, endoscope heads) and JET high heat flux elements for the beam-lines. The supposed interplay phenomenon between CuCrZr procurement, component geometry and welding procedure is discussed. The associated mechanical characterisation of materials from different suppliers and component joinings, focused especially on CuCrZr–CuCrZr fusion weldings using electron beams (EB), is presented.

International audience; The primary goal of an irradiation facility for fusion applications will be to generate a material irradiation database for the design, construction, licensing and safe operation of a fusion demonstration power station (e.g., DEMO). This will be achieved through testing and qualifying material performance under neutron irradiation that simulates service up to the full lifetime anticipated in the power plant. Preliminary investigations of 14 MeV neutron effects on different kinds of fusion material could be assessed by the SPIRAL 2 Project at GANIL (Caen, France), aiming at rare isotope beams production for nuclear physics research with first beams expected by 2009. In SPIRAL 2, a deuteron beam of 5 mA and 40 MeV interacts with a rotating carbon disk producing high-energy neutrons (in the range between 1 and 40 MeV) via C (d, xn) reactions. Then, the facility could be used for 3-4 months $y^−1$ for material irradiation purposes. This would correspond to damage rates in the order of 1-2 dpa $y^−1$ (in Fe) in a volume of $\sim$10 $cm^3$. Therefore, the use of miniaturized specimens will be essential in order to effectively utilize the available irradiation volume in SPIRAL 2. Sample package irradiation temperature would be in the range of 250-1000 °C. The irradiation level of 1-2 dpa $y^−1$ with 14 MeV neutrons (average energy) may be interesting for micro-structural and metallurgical investigations (e.g., mini-traction, small punch tests, etc.) and possibly for the understanding of specimen size/geometric effects of critical material properties. Due to the small test cell volume, sample in situ experiments are not foreseen. However, sample packages would be, if required, available each month after transfer in a special hot cell on-site.

Tore-Supra Tokamak ( I p = 1.5 MA, B t = 4 T) has been constructed with a steady-state magnetic field using super-conducting magnets and water-cooled plasma facing components (PFCs) for high-performance long pulse plasma discharges. When not actively cooled, plasma facing components can only accumulate a limited amount of energy since the temperature increases continuously during the discharge until radiation cooling equals the incoming heat flux. Such an environment is found in the JET Tokamak [JET Team, IAEA-CN-60/A1–3, Seville, 1994] and on TRIAM [M. Sakamoto, H. Nakashima, S. Kawasaki, A. Iyomasa, S.V. Kulkarni, M. Hasegawa, E. Jotaki, H. Zushi, K. Nakamura, K. Hanada, S. Itoh, Static and dynamic properties of wall recycling in TRIAM-1M, J. Nucl. Mater. 313–316 (2003) 519–523] [Y. Kamada, et al., Nucl. Fusion 3 (1999) 1845]. In Tore-Supra, the surface temperature of the actively cooled plasma facing components reach steady state within a second. We present here the Tore-Supra thermographic system, made of seven endoscope bodies equipped so far with eight infrared (IR) cameras. It has to be noted that this diagnostic is the first diagnostic to be actively cooled, as required for steady state. The main purpose of such a diagnostic is to prevent the plasma to damage the actively cooled plasma facing components (ACPFCs), which consist of the toroidal pumped limiter (TPL), 7 m 2 , and of five radio-frequency antennae, 1.5 m 2 each.

Tore-Supra is a Tokamak (Ip=1.5MA, Bt=4T) aiming at doing researches in the field of controlled nuclear fusion. It has been constructed with a steady-state magnetic field using super-conducting magnets and water-cooled plasma facing components for high performances long pulse plasma discharges. When not actively cooled, plasma-facing components can only accumulate a limited amount of energy since the temperature increase continuously (T˜sqrt(time)) during the discharge until radiation cooling is equal to the incoming heat flux (T>1800K). Such an environment is found in the most today Tokamaks. In the present paper we report the recent results of Tore-Supra, especially the design of the new generation of infrared endoscopes to measure the surface temperature of the plasma facing components. The Tore-Supra system is described in details, the new JET infrared thermography system is presented and some insight of the ITER set of visible/infrared endoscope is given.

Unexpected hot spots were observed around the edges of vertical ports on Tore-Supra, caused by fast electrons, accelerated by the lower hybrid waves used to drive the current, and trapped in the local ripple wells. Trajectory calculations, with the magnetic fields alone, show that no electrons should reach such locations. However, electrostatic potentials, at a kilovolt level, can induce a toroidal pinch of the trajectories, allowing particle deposition in these normally shadowed areas. Their origin is attributed first to a strong increase of the sheath potential at the port entrance, due to the fast electron flux itself, and second to its reduction within the port, when the sheath width becomes larger than the port width, due to the too low value of the local density. An increase of the capability of the cooling panels within the port, to cope with these additional fluxes, has been implemented.

The implementation of actively cooled high heat flux plasma facing components (PFCs) is one of the major ingredients required for operating the Tore Supra tokamak with very long pulses. A pioneering activity has been developed in this field from the very beginning of the device operation that is today culminating with the routine operation of an actively cooled toroidal pumped limiter (TPL) capable to sustain up to 10 MW/m 2 of nominal convected heat flux. Technical information is drawn from the whole development up to the industrialisation and focuses on a number of critical issues, such as bonding technology analysis, manufacture processes, repair processes, destructive and non-destructive testing. The actual experience in Tore Supra allows to address the question of D retention on carbon walls. Redeposition on surfaces without plasma flux is suspected to cause the final ‘burial’ of about half of the injected gas during long discharges.

A pioneering activity has been developed by CEA and the European industry in the field of actively cooled high heat flux plasma facing components in Tore Supra operation, which is today culminating with the routine operation of an actively cooled toroidal pumped limiter (TPL) capable of sustaining up to 10 MW/m2 of nominal convected heat flux. This success is the result of a long lead development and industrialization program (about 10 years) marked out with a number of technical and managerial challenges that were taken up and has allowed us to build up a unique experience feedback database, which is displayed in the paper.

The copper chromium zirconium (CuCrZr) alloy is extensively used in the Tokamak Tore Supra as a structural material for high heat flux components (HHFC). This precipitation-hardened material has also been selected for International Thermonuclear Experimental Reactor (ITER) for its good thermomechanical properties at elevated temperature. During a loss of coolant occurred in Tore Supra, a brittleness rupture occurred in CuCrZr high heat flux tube structure at high temperature. The scope of the analysis presented in this paper is to describe the rupture scenario and fracture mode.

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.

Actively cooled plasma facing components (PFCs) for Tore Supra (TS) have been designed basically for heat exhaust of ‘normal’ (convected and radiated) plasma power. However, in some cases, fast particles have been observed, which locally increased the power flux density, leading to damage of these PFCs and other inner vessel components. Three different examples for irreversible component damage, such as component melting and water leaks, are described involving runaway and supra-thermal particle strikes. In view of the capability for TS to handle larger input powers and to control the particles over long pulse durations, inner vessel components have been completely redesigned. The improved design concepts retained for the CIEL upgrade and preliminary results in the new configuration are presented.

The next generation of fusion experiments will operate for much longer pulse lengths within a context of high power density, introducing several new requirements for diagnostics compared with existing experiments. In this context, the upgrade of Tore Supra (CIEL Project) foresees high power and high radiating plasma scenario during very long pulse operation. This long plasma operation imposes the improvement of diagnostic systems and the design of thermally resistant in-vessel components with reliable window assemblies. Hence, the specific requirements and the methodology adopted to design Tore Supra/CIEL (TS) high thermal loaded diagnostic windows are summarised in this paper. Thermal and mechanical analyses are reported for different window materials and assembling methods.

A short review of diagnostics and techniques for active power exhaust control is given from the point of view of experiences gained at Tore Supra. The physical and technological aspects are presented with the aim of illustrating strengths, difficulties and potential for future long pulse and steady state fusion devices such as W7X and ITER-FEAT. One of the main features of the Tore Supra approach to power exhaust control is the reliance on infrared (IR) thermography to survey nearly 100% of the high flux target zones. This is applied to the largest elements, the toroidal pump limiter and the heating antennae, by a number of IR endoscopes. For high flux target areas inaccessible to direct view, optical fibres in the IR and the near-IR range are used. The large wall panels, which evacuate the rest of the power, are controlled by calorimetry. Bolometry is used to achieve a complete power balance. These measurements can be inserted in control loops acting typically on the injected heating power or the radiation cooling.