P. Languille

Cea Leti, Grenoble, Rhône-Alpes, France

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Publications (18)26.37 Total impact

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    ABSTRACT: In the context of the Tokamak Tore-Supra evolution, the CEA aims at transforming it into a test bench for ITER actively cooled tungsten (ACW) plasma facing components (PFC). This project named WEST (Tungsten Environment in Steady state Tokamak) is especially focused on the divertor target. The modification of the machine, by adding two axisymmetric divertors will make feasible an H-mode with an X-point close to the lower divertor. This environment will allow exposing the divertor ACW components up to 20 MW/m2 heat flux during long pulse. These specifications are well suited to test the ITER-like ACW target elements, respecting the ITER design.
    Fusion Engineering and Design 03/2015; DOI:10.1016/j.fusengdes.2015.02.053 · 1.15 Impact Factor
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    ABSTRACT: The primary goal of the WEST project is to be a test-bed for ITER W divertor components, in terms of manufacturing issues as well as fatigue and lifetime regarding thermal loads during operations. Therefore it is necessary to ensure that the thermal loads available in Tore Supra are of the same magnitude as those expected for ITER. It is also necessary to ensure that the other plasma facing components would not be a limitation to reach this target. On this basis, simulations of the incident heat flux on the main components have been done for different parameters. Those simulations are done with PFCFlux code and include the variation of the toroidal magnetic field called ripple effect. This paper reports the results of those simulations and concludes on the usability of the plasma facing components with the required heat loads.
    Fusion Engineering and Design 01/2015; DOI:10.1016/j.fusengdes.2014.12.024 · 1.15 Impact Factor
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    ABSTRACT: 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.
    Fusion Engineering and Design 01/2015; DOI:10.1016/j.fusengdes.2014.12.017 · 1.15 Impact Factor
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    ABSTRACT: Recently, the surface of carbon fibre composite tiles of the toroidal pump limiter of Tore Supra has been analysed by scanning electron and atomic force microscopies. In regions where fibres are perpendicular to the surface, a specific erosion pattern has been observed. It is constituted of a striation oriented with an angle oblique with respect to the magnetic field. The characteristic wavelength of this structure is micrometric, and similar to the fibre size. Modelling has been undertaken to reproduce this micrometric pattern. It is shown to originate from the carbon composite structure, for which it has been found by measurement using laboratory plasma that the erosion rate of the fibres is different from that of the surrounding matrix. Modelling emphasizes the effect of the impinging flux angle distributions of deuterium ions and carbon impurities that are preliminarily determined from computation of the magnetic sheath. In the case of deuterium the sheath is shown to have little effect on the particle trajectories for the simulation parameters considered here, although when impurities are included the sheath deflection is significant. Furthermore this study shows how the fibre organization in the composite influences the striation direction and points out the importance of the angular dependence of the sputtering yield.
    Nuclear Fusion 11/2014; 54(12):123006. DOI:10.1088/0029-5515/54/12/123006 · 3.24 Impact Factor
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    ABSTRACT: 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.
    Fusion Engineering and Design 10/2014; DOI:10.1016/j.fusengdes.2014.01.050 · 1.15 Impact Factor
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    ABSTRACT: 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 Environment for Steady state Tokamak) 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/m(2).
    Fusion Engineering and Design 10/2013; 88(9-10):1793-1797. DOI:10.1016/j.fusengdes.2013.03.048 · 1.15 Impact Factor
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    ABSTRACT: The laser ablation technique had been already shown to be well suited to clean the top surface of the Toroidal Pumped Limiter of Tore Supra. The next phase was to apply this technique to the deposits observed in the gaps. This paper focuses on the characterization of gap deposits and the optimization of the laser cleaning process for this application. In general most of the deposits are located inside the gap within the first mm. In the toroidal direction most of the deposits are located on the low field side with thicknesses reaching up to 610 μm, while in the poloidal direction deposits thicknesses up to 450 μm are observed on both sides. The influence of the laser angle has been investigated; an angle of 45° is recommended in the toroidal direction while a perpendicular beam is adopted in the poloidal direction. This study concludes that gap cleaning induces significant additional time for castellated surfaces.
    Journal of Nuclear Materials 07/2013; 438:1079-. DOI:10.1016/j.jnucmat.2013.01.237 · 2.02 Impact Factor
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    ABSTRACT: A diagnostic based on confocal microscopy was developed at CEA Cadarache in order to measure erosion on large plasma facing components during shutdown in situ in Tore Supra. This paper describes the diagnostic and presents results obtained on Beryllium and Carbon Fibre Composite (CFC) materials. Erosion in the range of 800 μm was found on one sector of the Toroidal Pumped Limiter (TPL) which provides, by integration to the full limiter a net carbon erosion of about 900 g over the period 2002-2007.
    Journal of Nuclear Materials 07/2013; DOI:10.1016/j.jnucmat.2013.01.269 · 2.02 Impact Factor
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    ABSTRACT: This paper presents an analysis of the carbon–deuterium circulation and the resulting balance in Tore Supra over the period 2002–2007. Carbon balance combines the estimation of carbon gross erosion from spectroscopy, net erosion and deposition using confocal microscopy, lock-in thermography and SEM, and a measure of the amount of deposits collected in the vacuum chamber. Fuel retention is determined from post-mortem (PM) analyses and gas balance (GB) measurements. Special attention was paid to the deuterium outgassed during the nights and weekends of the experimental campaign (vessel under vacuum, Plasma Facing Components at 120 °C) and during vents (vessel at atmospheric pressure, PFCs at room temperature). It is shown that this outgassing is the main process reconciling the PM and GB estimations of fuel retention, closing the coupled carbon–deuterium balance. In particular, it explains why the deuterium concentration in deposits decreases with increasing depth.
    Journal of Nuclear Materials 07/2013; 438(Supplement):S120-S125. DOI:10.1016/j.jnucmat.2013.01.019 · 2.02 Impact Factor
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    ABSTRACT: a b s t r a c t Analyses of erosion and deposition over the toroidal pump limiter of Tore Supra were performed combin-ing scanning electron microscopy, confocal microscopy and lock-in thermography. The consistency between the different methods allows a complete mapping of the eroded and deposited mass of carbon to be performed. $920 g of eroded carbon and $520 g of deposited carbon are found, showing that more than half of the eroded carbon is redeposited on the limiter. The highest deposition zones are close to the eroded zones. The gap deposition contribution is estimated at $23%, mostly from the erosion zones and with a main contribution from the low field side of the tile toroidal gap surfaces. Raman microscopy and transmission electron microscopy in-depth analyses show that the structure of deposits is non-homoge-neous, in agreement with a deuterium impoverishment of the deep layers. Ó 2013 Elsevier B.V. All rights reserved.
    Journal of Nuclear Materials 07/2013; 438. DOI:10.1016/j.jnucmat.2013.01.165 · 2.02 Impact Factor
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    ABSTRACT: As part of the deuterium inventory in the Tore Supra project, a detailed structural analysis of the eroded tiles (carbon fiber composite tiles) of the main plasma-facing component of Tore Supra, the toroidal pump limiter, has been performed. It combines scanning electron microscopy, atomic force microscopy and Raman microscopy. A tile cross-section has been scanned, showing the eroded profile of the tile and allowing the net erosion rate to be determined. Ripples are observed on the surface of the eroded tiles and this pattern gives a clear indication of the ion flux direction. The height profile along the valleys indicates that the fibers are eroded faster than the matrix, the difference in the erosion rates probably originating from differences in microstructure.
    Physica Scripta 12/2011; 2011(T145):014024. DOI:10.1088/0031-8949/2011/T145/014024 · 1.30 Impact Factor
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    ABSTRACT: We performed electron microscopy analysis of carbon tiles dismantled from the toroidal limiter of the Tore Supra tokamak. The tile top surfaces are either eroded, or covered with deposits, depending on ion flux. On the contrary, the tile gap surfaces are covered with deposits whatever the region of the limiter. We analyzed both the topography and the microstructure of gap deposits. Deposits are tip-shaped and deposition is significant only down to ~1 mm, due to the limited penetration of ions. The direction of tips results from the combination of the ion speed along the magnetic lines and the drift associated to the local electric fields. The microstructure of deposits reveals the presence of graphitic species, such as onion-like nanoparticles, and our results therefore show that deposition results from local plasma chemistry similar to what is observed for cold plasmas.
    The European Physical Journal Applied Physics 11/2011; 56(2). DOI:10.1051/epjap/2011110171 · 0.79 Impact Factor
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    ABSTRACT: Nuclear Reaction Analysis (NRA) with a 3He ion beam is a powerful analytical technique for analysis of light elements in thin films. The main motivation for 3He focused beam applications is lateral mapping of deuterium using the nuclear reaction D(3He,p)4He in surfaces exposed to a tokamak plasma, where a lateral resolution in the μm-range provides unique information for fuel retention studies.At the microprobe at the Jožef Stefan Institute typical helium ion currents of 300 pA and beam dimensions of 4 × 4 μm2 can be obtained. This work is focused on micro-NRA studies of plasma-facing materials using a set-up consisting of a silicon partially depleted charge particle detector for NRA spectroscopy applied in parallel with a permanently installed X-ray detector, an RBS detector and a beam chopper for ion dose monitoring. A method for absolute deuterium quantification is described. In addition, plasma-deposited amorphous deuterated carbon thin films (a-C:D) with known D content were used as a reference.The method was used to study deuterium fuel retention in carbon fibre composite materials exposed to a deuterium plasma in the Tore Supra and TEXTOR tokamaks. The high lateral resolution of micro-NRA allowed us to make a detailed study of the influence of topography on the fuel retention process. We demonstrated that the surface topography plays a dominant role in the retention of deuterium. The deep surfaces inside the castellation gaps showed approximately two orders of magnitude lower deuterium concentrations than in areas close to the exposed surface.
    Nuclear Instruments and Methods in Physics Research Section B Beam Interactions with Materials and Atoms 10/2011; 269(20):2317-2321. DOI:10.1016/j.nimb.2011.02.049 · 1.19 Impact Factor
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    ABSTRACT: Scanning and transmission electron microscopy analyses have been performed for tiles extracted from the toroidal pump limiter of Tore Supra for erosion- and deposition-dominated zones. Deposit thicknesses have been estimated for the plasma-facing top and the gap side lateral surfaces. Deposit thickness profiles have been measured inside gaps, showing that deposition mainly occurs in the first millimetre and that both poloidal and toroidal gap deposition is asymmetric. Quantitative information on the deposit volume and on D-retention are thus obtained from these measurements. Carbon probed at the tile top surfaces is mainly amorphous carbon, due either to the amorphization induced by ion bombardment in the erosion dominated zone, or to deposit formation processes in the deposition-dominated zones. Deposits are tip-shaped and are oriented, which should give information on transport processes.
    Journal of Nuclear Materials 08/2011; DOI:10.1016/j.jnucmat.2010.11.006 · 2.02 Impact Factor
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    ABSTRACT: A laser ablation technique has been adapted for removal of the deposited layers on carbon plasma facing components from Tore Supra. This paper describes in detail the experiments performed to adjust the scan parameters and optimize the laser ablation technique. The goal was to reduce the process duration in order to be able to treat large surfaces. To remove layers up to 50 mu m thick, with an Ytterbium fiber laser of 1 mJ, the process duration limit is 20 h/m(2). The conditions to reach 1 h/m(2) are presented. Confocal Microscopy (CM) as well as Nuclear Reaction Analysis (NRA) and Thermal Desorption Spectroscopy (TDS), were used to assess the thickness of the layers removed and the Deuterium content before and after ablation. The efficiency of the laser ablation technique to remove the D content is established.
    Journal of Nuclear Materials 08/2011; 415(1):S797-S800. DOI:10.1016/j.jnucmat.2010.08.033 · 2.02 Impact Factor
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    ABSTRACT: A dedicated study on fuel retention has been launched in Tore Supra, which includes a D wall-loading campaign and the dismantling of the main limiter (Deuterium Inventory in Tore Supra, DITS project). This paper presents new results from a second post-mortem analysis campaign on 40 tiles with special emphasis on the D retention in the gaps. SIMS analysis reveals that only 1/3 of the thickness of deposits in the plasma shadowed zones are due to the DITS wall-loading campaign. As pre-DITS deposits contain less D than DITS deposits, the contribution of DITS to the D inventory is about 30-50%. The new estimate for the total amount of D retained in the Tore Supra limiter is 1.7 × 1024 atoms, close to the previous estimate, with the gap surfaces contributing about 33%. NRA measurements show a stepped decrease of D along the gap with strong asymmetries between different gap orientations.
    Journal of Nuclear Materials 07/2011; 415:757. DOI:10.1016/j.jnucmat.2010.11.075 · 2.02 Impact Factor
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    ABSTRACT: In order to benchmark predictions for the in vessel tritium inventory in ITER, a survey of fuel retention measured in 4 carbon dominated tokamaks (TEXTOR, ASDEX Upgrade in the 2002–2003 carbon configuration, Tore Supra and JET) was performed, showing retention rates from ∼1g D/h in TEXTOR (L mode, limiter machine) up to ∼6–12g D/h in AUG (H mode, divertor machine). A simple scaling used for ITER predictions is applied for comparison with experimental values: (1) estimate of wall fluxes, (2) estimate of the gross carbon erosion, (3) estimate of the net erosion/redeposition assuming a redeposition fraction and (4) estimate of the retention rate using D/C ratio scalings. The validity of each step is discussed, showing that this approach yields the right order of magnitude, but tends to underestimate the experimental values unless a high wall flux, a low local redeposition fraction and/or a high D/C ratio are used.