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The net erosion and deposition patterns in the inner and outer divertor of WEST were determined after different experimental campaigns (C3 and C4) of the first operational phase using ion beam analyses and scanning electron microscopy techniques. The analyses were performed on four entire tiles from inertially cooled, W-coated divertor units with an additional Mo marker coating covered with a further W coating. Strong erosion occurred at the expected location of the inner and outer strike line area with a campaign-averaged net erosion rate of >0.1 nm s−1. On the high field side of the inner strike line area, thick deposited layers were found (>10 μm; growth rate >1 nm s−1), mainly composed of B, C, O, and W. Additionally, strong arcing was observed in this region. At the end of the C4 campaign, He discharges were performed to study the He-W interaction. Although the conditions for nanotendrils, i.e. fuzz formation were fulfilled around the outer strike line position, neither nanotendrils nor He bubbles (>10 nm) were observed at this area.

In the WEST (W -for tungsten- Environment in Steady-state Tokamak) tokamak, an actively cooled tungsten divertor for testing high heat flux technology is implemented. Beside the presently tested ITER divertor technology (i.e. W-monoblock), the WEST team in collaboration with ASIPP (China) worked on the development of an alternative design for actively cooled W/Cu plasma facing components capable to sustain heat loads close to monoblock design. Within this framework, two small-scale mock-ups based on the W-armoured flat-tile concept were produced by ASIPP and successfully tested in the electron beam facility JUDITH-1 (FZ-Julich, Germany) under successive thermal cycling at 10, 15 then 20 MW/m². These results showed the capacity of this technology to remove heat loads over several hundreds of cycles at 20 MW/m2 without obvious indication of damage impacting the heat exhaust capability. However, the post-mortem analysis showed crack formation perpendicular to the loaded surface proceeding into the pure Cu interlayer for those tiles tested at 20 MW/m². This contribution focuses on the thermal and structural analysis performed with ANSYS for these components. The convective heat transfer coefficients were calculated using the Nukiyama model and the mechanical properties of pure copper at high temperature were extrapolated using the Ramberg-Osgood model. The thermal results are in good agreement with the experiments. The thermo-mechanical simulation explains partially the defects observed on the mocks up with the selected approach, which leads to suggest improvements for further study by modelling kinematic hardening for tungsten and creep for oxygen-free copper.

Measuring the temperature in plasma-facing components (PFCs) provides information both on plasma parameters in the divertor region and on the thermal stress experienced by PFCs. Fiber Bragg gratings (FBGs) are interesting candidates for this application because they are immune to electromagnetic interferences and their ability to be multiplexed allows an extended spatial coverage. Four fibers, each of them including eleven regenerated Bragg gratings, have been embedded in tungsten-coated graphite components and operated up to their signal-collapsing limit at 800 °C. Extending the measurement range towards higher temperatures increases the sensitivity to plasma parameters and allows withstanding higher energy experiments. To overcome thermal limitations, the system is up-graded using femtosecond laser inscribed fibers. In addition to their outstanding thermal stability, femtosecond FBGs benefit from higher signal-to-noise ratios than regenerated FBGs. The paper addresses femtosecond FBGs design and issues relative to their integration inside the actively cooled ITER-like PFCs of the WEST tokamak. The period and length of the gratings are designed to increase the number of measurement spots to fourteen gratings per fiber, regularly distributed over 17 cm, while ensuring robust detection even with strong thermal gradients (no overlapping or deformation of Bragg peaks). The system operates up to 1200 °C with gradients reaching 200 °C/mm perpendicularly and 40 °C/mm in parallel to the fiber. FBGs are inserted in actively cooled ITER-like PFCs through a 2.5 mm deep lateral groove localized at 5 mm beneath the top of bulk tungsten monoblocks. A PFC mock-up machined with a groove has been tested under HHF facility to assess the effect of the groove on monoblocks thermal behavior. The test demonstrates that machined monoblocks behave as expected from simulation and can withstand 20 MW/m2 heat flux (i.e. 1200 °C in the fiber) with 20 % overheating as compared to intact monoblocks.

see also: http://west.cea.fr/WESTteam; International audience; In the WEST tokamak, ITER-like divertor targets consisting of tungsten monoblocks bonded via an OHFC-Cu compliance layer to CuCrZr cooling tubes were exposed to plasma during the 2018 experimental campaign in which modest heating power was available. Up to 2.5 MW/m$^2$ divertor surface heat flux was attained. Inspection of the components after the campaign revealed a wide variety of damage at both leading and trailing monoblock edges, and at the optical hot spots which are the projections along magnetic field lines of the toroidal gaps between monoblocks onto the poloidal leading edges. Cracking, deformation, and melting occurred. Consideration of the large body of past work on high heat flux testing, combined with the expected loading conditions in WEST, suggests that fractures form during the first transient events such as disruptions. Deformation occurs under subsequent exposure to steady state heat loads. Nearly identical damage was observed on reciprocating probes made of W-La(10%) alloy under measured irradiation conditions, lending credence to this hypothesis.

In preparation of the production of the actively cooled tungsten targets for the lower divertor of WEST (W -for tungsten- Environment in Steady-state Tokamak), the qualification of suppliers is done by exposing small-scale mock-ups to cyclic high heat flux tests in the electron beam facility JUDITH 1 at Forschungszentrum Julich. Thereby, the small-scale mock-ups are based on the ITER design, consisting of 7 monoblocks mounted on a CuCrZr cooling tube and connected via a soft Cu-interlayer. The results presented herein focus on the high heat flux test results of two mock-ups produced by AT&M in China. The testing was performed by thermal cycling at 10 MW/m2 and 20 MW/m2 with a maximum of 1200 cycles at 10 MW/m2 followed by 500 cycles at 20 MW/m2 without obvious damage formation. The qualification is supported by subsequent hardness tests and microstructural analyses. The latter focus on local variation of tungsten recrystallization and macro-crack formation and deformation induced changes in the heat sink materials.

La conception des composants face au plasma est un enjeu majeur pour les machines de fusion de prochaine generation. Cet article fait le point sur les avancees significatives dans ce domaine, depuis les premieres generations de composants des annees 1980 jusqu’aux composants actuellement en cours de construction pour ITER. Apres un rappel des conditions extremes auxquelles sont soumis les composants, les materiaux retenus sont passes en revue, ainsi que les differents concepts de composants et les moyens de qualification associes.

The development of a divertor target for DEMO is of great importance, being able to sustain the harsh environment that is imposed on this component. To fulfill the loading requirements, different concepts were developed within the EUROfusion WPDIV project. The baseline concept is based on the ITER divertor target W-monoblock design. It is made of tungsten as armour material, CuCrZr as structural material and Cu-OFHC as compliant layer. One of the proposed alternative concepts aims to minimize the stress at interfaces by replacing the thick copper interlayer by W/Cu functionally graded material (FGM). In this study, the FGM interlayer, with a thickness of 500 μm, is composed of stacked elementary layers with the following three compositions: 25 vol.%W + 75 vol.%Cu, 50 vol.%W + 50 vol.%Cu, 75 vol.%W + 25 vol.%Cu. Several FGM interlayers were studied. In total three monoblock type mock-ups were manufactured. This paper describes the steps needed to manufacture mock-ups and characterization of elementary layers (composition, porosity, Young’s modulus). HHF tests applying up to 1000 cycles at 20 MW/m² and sub-sequent post-mortem examinations were performed to qualify the concept performance.

We analyzed data from ultrasonic testing and infrared thermography non-destructive examinations performed on a subset of 25 small scale water-cooled target mockups of four target designs developed in the 2nd R&D phase of WPDIV. Examinations were performed before high heat flux tests for the 25 mockups and after high heat flux tests for 18 of them. The detected manufacturing defects are various in size and location. Widths of defects range from 0.2 to 12 mm, which is the entire block width. Angular sizes of defects range from few degrees to 360°. Defects having an angular size less than 50° or a width less than 4 mm are likely to be missed by infrared thermography examinations. After high heat flux tests at 20 MW m−2 up to 500 cycles, we noticed no significant evolution of pre-existing defects.

The consequences of tungsten (W) melting on divertor lifetime and plasma operation are high priority issues for ITER. Sustained and controlled W-melting experiment has been achieved for the first time in WEST on a poloidal sharp leading edge of an actively cooled ITER-like plasma facing unit (PFU). A series of dedicated high power steady state plasma discharges were performed to reach the melting point of tungsten. The leading edge was exposed to a parallel heat flux of about 100 MW.m−2 for up to 5 s providing a melt phase of about 2 s without noticeable impact of melting on plasma operation (radiated power and tungsten impurity content remained stable at constant input power) and no melt ejection were observed. The surface temperature of the MB was monitored by a high spatial resolution (0.1 mm/pixel) infrared camera viewing the melt zone from the top of the machine. The melting discharge was repeated three times resulting in about 6 s accumulated melting duration leading to material displacement from three similar pools. Cumulated on the overall sustained melting periods, this leads to excavation depth of about 230 μm followed by a re-solidified tungsten bump of 200 μm in the JxB direction.

International audience; Plasma Facing Components (PFCs) temperature measurement is required to ensure safe high power for long pulse tokamak operation and for physics studies. A set of twenty thermocouples (TCs) and four optical fiber temperature sensing probes, each of them including eleven wavelengthmultiplexed fiber Bragg gratings (FBGs) written with UV radiation, have been integrated and deployed in the WEST lower divertor. To avoid any thermal erasure occurring above 400°C, the gratings have been regenerated using a high temperature annealing process. The FBG probes are embedded in tungsten-coated graphite components with no active cooling. The diagnostic was working correctly from the first plasma breakdown achieved in WEST (in 2017) until the first dismantling of the lower divertor three years later (2020). Collected data are sorted by temperature range to investigate the ageing of both PFCs and diagnostics. Measurements obtained with FBGs and TCs are consistent, no long-term drift is reported during the overall period of use. The maximum temperature reported by FBGs in the divertor is 830 °C obtained with 90 MJ of cumulated plasma energy in one pulse. A short but reversible collapse of the Bragg peak (from 5 down to 2 dB) has been observed under high temperature due to the strong temperature gradients measured along the fiber (about 20°C/mm, for an FBG size of 3 mm). This feedback demonstrates that regenerated FBGs are suitable for long-term temperature monitoring in tokamak environment with high heat load, high temperature and gradient, strong magnetic field and vacuum constraint.

The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the ITER tungsten divertor strategy. The WEST project aims to test W monoblock Plasma Facing Units (PFU) under long plasma discharge (up to 1000s), with thermal loads of the same magnitude as those expected for ITER. Therefore the divertor is a key component of the WEST project, and so is its support structure, which has to handle strong mechanical loads. The WEST upper and lower divertor are made of 12 30° sectors, each one composed of 38 PFU that can be made of tungsten, CuCrZr or graphite. A generic 316L stainless steel 30° conic support plate is used to hold the 38 PFU together, regardless of their material. The PFUs are fixed on the support plate thanks to 152 Xm19 stainless steel fixing elements (4 per PFU), and in each of this fixing element an Aluminium-Nickel-Bronze alloy (Al-Ni-Br) pin is engaged in a slotted hole, in order to allow thermal expansion in the length direction of the PFU. The support plate is fixed on the divertor coil casing thanks to 10 M10 screws. Mechanicals loads which act on the PFUs are transmitted to the support plate through the fixing elements. These loads are due to Vertical Displacement Event (VDE), disruptions and thermal expansion of the PFU. First the different load cases, PFU configurations and scenario are presented. Then an ANSYS plastic mechanical simulation is performed in order to validate the number of cycles of the support plate for each scenario: 30 000 cycles in steady-state and 3000 cycles in VDE. Finally reactions forces from the previous ANSYS simulation are used in order to calculate the stress in the M10 screws.

International audience; 4 probes with Fiber Bragg Gratings (FBGs) will be integrated in Plasma Facing Component (PFC) of the WEST lower divertor in order to measure the bulk temperature distribution in the poloidal direction (11 locations equally spaced). A 2D nonlinear unsteady calculation combined with the Conjugate Gradient Method (CGM) and the adjoint state is used in order to estimate the space and time evolution of the surface heat flux based on temperature measurements. Synthetic measurements are generated assuming optical projection of the heat flux taking into account physical effects such as heat flux decay length ( λq ) and power spreading factor ( S ) in the private flux region and geometrical effects such as surface chamfering and round leading edges of the component. The inverse method is applied on different synthetic measurements in order to evaluate the ability of using FBG measurements to characterize the heat flux poloidal profile.

The present paper deals with the surface heat flux estimation with thermocouples (TC) and fiber Bragg grating (FBG) embedded in the plasma facing components (PFC) of the WEST tokamak. A 2D heat transfer model combined with the conjugate gradient method (CGM) and the adjoint state is used to estimate the plasma heat flux deposited on the PFC. The plasma heat flux is characterized by the time evolution of its amplitude and spatial shape on the target (heat flux decay length $\lambda^t_q$, power spreading in the private flux region $S^t$ and the strike point location $x_0$). As a first step, five ohmic pulses have been investigated with different magnetic configuration and divertor X-point height varying from 44 to 68 mm from the surface. Despite an outboard shift, the relative displacements of the outer strike point as well as the heat flux decay length derived from the TC/FBG systems are consistent with the magnetic equilibrium reconstruction., Comment: Fusion Engineering and Design, Elsevier, 2019

The lifetime of tungsten coatings used for Plasma Facing Components (PFCs) in fusion devices is mainly limited by two degradation mechanisms occurring during the operation of the tokamak: erosion resulting from exposure to long plasma discharges (long term) and local damages caused by transient events (runaway electrons,..). An alternative solution to the current replacement of damaged PFCs would be to repair their surface by stripping the damaged tungsten coating and depositing a new one. In this paper, the process of removing tungsten was investigated on CuCrZr samples coated with a 15 μm layer of W using four techniques: plasma etching, sandblasting, chemical attack and laser ablation. For each technique, etching properties were assessed by means of microscopic observations, topographical surface analyses and chemical analysis. The results presented here also highlight the issues related with controlling the etching depth and size of the treated surface.

Based on a monoblock concept (e.g. a tube-in-tile concept), each elementary tungsten plasma facing component (called Plasma-Facing Unit PFU) of the WEST lower divertor follows as closely as possible the same monoblock geometry, materials and bonding technology that is envisaged for ITER. A fatigue simulation of W PFU was used to validate its specific integration into WEST. The complex design, the material heterogeneities and the usage outside operational load design envelope are all possible causes of fatigue failure. This paper shows how the ITER like monoblocks and its U-shaped attachments technology are integrated into the WEST divertor by performing finite element analysis. The WEST lower divertor is designed to withstand 15 MW steady-state of injected power, with peaked heat fluxes up to 20 MW/m 2 . The integration and the design choices of a WEST ITER Like Plasma Facing Unit inside the WEST vacuum chamber is valid for an “expected life time” of repeated inter ELMs thermal steady state (>10 s) cycles and for 300 off-normal vertical displacement events.

International audience; Plasma Facing Components (PFC) temperature measurement is mandatory to ensure safe high power and long pulse tokamak operation. IR thermography systems which are widely used in magnetic fusions devices become challenged with the choice of tungsten as a PFC material in the I l'ER tokamak, mainly due to emissivity uncertainties and reflection issues in a hot environment. Embedded temperature measurements are foreseen to cross-check the IR thermography measurements. Fiber Bragg grating sensors are potentially of great interest for this application because they are immune to electromagnetic interference and allow the measurement of a large number of temperature spots on a single fiber. Four optical fiber temperature sensing probes, each of them including 11 regenerated fiber Bragg gratings equally spaced by 12.5 mm (equivalent to one ITER-like tungsten monoblock), have been specifically designed and manufactured for the WEST project (W-tungsten Environment and Steady State Tokamak). The four probes are embedded in W-coated graphite components at two different distances from the surface, 3.5 mm and 7 mm, to cover a wide range of temperatures up to 900 degrees C. This paper addresses the design and integration issues and the qualification and performance assessment performed in the laboratory. It also shows the first measurements of this new diagnostic achieved in a tokamak environment during baking of the machine and during early diverted plasma exposure.

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.

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. One challenge in such machine evolution is to integrate components in an existing vacuum vessel in order to obtain the best achievable performance. This paper deals with the design integration of ITER ACW target elements into the WEST environment considering magnetic, electric, thermal and mechanical loads. The feasibility of installation and maintenance has to be strongly considered as these PFC could be replaced several times. The ports size allows entering a 30° sector of pre-installed tungsten targets which will be plugged as quickly and easily as possible. The main feature of steady state operation is the active cooling, which leads to have many embedded cooling channels and bulky pipes on the PFC module including many connections and sealings between vacuum and water channels. The 30° sector design is now finalized regarding the ITER ACW elements specifications. No major modifications are expected.

The paper discusses a potential approach of infrared (IR)-analysis for high heat flux quality assessment of divertor tungsten monoblock components. The aim of these activities is the development of a statistics based method for the general application to the high heat flux testing of industrially manufactured W monoblock plasma-facing components. We have performed tests with 100 cycles at 10 MW/m 2 on 13 components or mock-ups manufactured for ITER, for the DEMO divertor development and for the WEST project. For eight W monoblock components with grinded surfaces, the statistical high heat flux (HHF) quality assessment method originally developed for the series manufacturing of carbon fibre reinforced carbon flat tile divertor components for Wendelstein 7-X was applied. The large variability of the emissivity of W as well as the monoblock geometry require a careful emissivity correction of IR camera data. The emissivity is strongly dependent on the W surface quality and final machining and influences the apparent temperature measured by an IR camera. Therefore we have applied a pragmatic correction algorithm based on two-colour pyrometry. Finally we propose a possible HHF testing strategy of a large number of industrially manufactured W monoblock components.

During steady state plasma operation in fusion devices, leading edges of the actively cooled plasma-facing components can be impacted by plasma flux with nearly normal angle of incidence, causing local overheating. The overheating can be a critical issue in high-power machines, especially in the presence of mechanical misalignments. Due to heat diffusion through the material, the edge power overload also leads to a local increase of temperature on the top part of the tile that can be detected by the infrared imaging system (viewed from the top of the machine). In the Tore Supra tokamak, heat flux impinging on the top and the leading edge of the carbon fibre composite (CFC) flat tiles are characterized with both an infrared (IR) thermographic system and 2D thermal modelling of the tile. A specific sensor correction based on a laboratory blackbody-slit experiment has been developed to simulate the spatial resolution related effects (necessary here since the temperature gradient near the leading edge is smaller than the pixel size of the IR system). The transfer function of the IR system is characterized by a Gaussian distribution function. The standard deviation is found to be σ = 1.75 mm for a pixel size of 3.1 mm. The heat flux calculation is applied to CFC flat tiles and, after being processed with the transfer function, compared to experimental IR data for two geometrical situations: one with 0.2 mm misalignment between two adjacent tiles and the other without misalignment (well-aligned tiles). The heat flux ratio between the leading edge and top is found to be ~25 in the case of the protruding tile, which is lower than the expected ratio using the guiding-centre ballistic approximation with no cross-field heat flux (57).

The Tore Supra tokamak is being transformed in an x-point divertor fusion device within the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the ITER tungsten divertor strategy. The WEST project aims at testing W monoblock Plasma Facing Units (PFU) under long plasma discharge, with thermal loads of the same magnitude as those expected for ITER. The others Plasma Facing Components (PFC) will also be modified and coated with W to transform Tore Supra into a fully metallic environment. Different coating techniques have been selected, taking into account the specifications of the various PFC: heat loads, complex geometries (length up to 1m) and different substrates (CuCrZr for actively cooled PFC, graphite and CFC for other components). This paper gives an overview on the different processes used and the associated validation program and concludes on the adequacy of the W coating with the WEST experimental program requirements.

The divertor is the key in-vessel plasma-facing component being in charge of power exhaust and removal of impurity particles. In DEMO, divertor target must survive an environment of high heat fluxes (∼up to 20 MW/m2 during transient) and neutron irradiation. This paper presents the first part of the development for the realization of an advanced actively cooled concept aiming at improving the quality of material joining. It is based on the use of a W/Cu functionally graded material (FGM) interlayer instead of the copper interlayer. For this preliminary development, the interlayer (∼25 μm) is made with PVD (Physical Vapor deposition) and composed of continuous gradient of W from 100% at the interior part of the tile to 0% at the external part while Cu varies from 0% to 100%. Hot isotactic pressing is realized to join the CuCrZr tube to W tile equipped with graded material. With infrared thermography (SATIR facility) and US non-destructive examinations, manufactured mock-ups show good interface quality.

In the field of fusion experiments using a tokamak, the plasma facing components (PFC) are the closest object to the hot plasma. Due to the plasma-wall interaction, the material composing the PFC may enter the plasma and disturb the experiments. In the past, the main material for PFC was carbon (CFC, graphite), while the future reactors like ITER will be fully metallic, in particular tungsten. The Tore Supra tokamak has been transformed in an x-point divertor fusion device within the frame of the WEST (W (tungsten) Environment in Steady-state Tokamak) project in order to have plasma conditions close to those expected in ITER. The PFC other than the divertor has been coated with W to transform Tore Supra into a fully metallic environment. Different coating techniques have been selected for different kind of PFC. This paper gives an overview on the coating process used for the antennae protection limiter, the associated validation programme and concludes on the adequacy of the W coating with the WEST experimental programme requirements and gives perspectives on the development to be pursued.

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.

This paper presents the technical solutions and methodologies that are used and under development for managing the design lifecycle of the WEST project (W – for tungsten – Environment in Steady-state Tokamak, upgrade of Tore Supra's with actively cooled tungsten plasma facing components) fusion components and explains the interfaces that are implemented or in construction to connect together the different tools like documents management system, CAD modeler, or simulation codes around the data management backbone. It describes the methodologies used on the WEST project to optimize the design process by managing the engineering data workflow and ensuring the consistency between the different 3D representations for design or analysis as well as the specification or interfaces documents. Finally it explains how this platform contributes to reach the project targets in terms of performance, cost and schedule.

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.

The W -for tungsten- environment in steady-state tokamak (WEST) project consists in transforming the French tokamak Tore Supra into an X-point divertor one and aims at testing the ITER technology of actively cooled high heat flux tungsten components during ITER-relevant long pulses. In addition to the integration in the vacuum vessel of a lower divertor and an upper one, this metamorphosis has a strong impact on diagnostics initially present in Tore Supra. Indeed, the introduction of a divertor in the upper part of the machine leads to block the lines of sight (LOS) of diagnostics previously installed in the upper ports. Therefore, new diagnostics playing a paramount role for the safe tokamak operation need to be implemented, such as an infrared (IR) thermography diagnostic for the protection of plasma-facing components (PFCs) and heating antennas, and a visible spectroscopy system for the protection of tungsten PFCs. Besides, several existing diagnostics need to be modified to adapt their LOS to the new divertor targets and plasma shape. This paper presents the main issues linked to the integration of diagnostics deriving from the WEST project. First, the changes involved by the upgrade on the set of diagnostics currently implemented in Tore Supra are described. Then, a particular focus is put on the integration of the two new optical systems, namely the IR system and the visible spectroscopy one. These two systems being located within the vacuum vessel, a special care must be paid to their design, to allow them to withstand their harsh environmental conditions. Other new or updated fundamental diagnostics are also briefly presented. Finally, the resulting new implementation of diagnostics envisaged to reach the scientific and technological goals of the WEST project is described.

Lower hybrid current drive (LHCD) and ion cyclotron resonance heating (ICRH) are recognized as important auxiliary heating and current drive methods for present and next step fusion devices. However, these radio frequency (RF) systems generate a heat flux up to several MW/m 2 on the RF antennas during plasma operation. This paper focuses on the determination of the heat flux deposited on the lateral protections of the RF antennas in Tore Supra. The heat flux was calculated by finite element method (FEM) using a model of the lateral protection. The FEM calculation was based on surface temperature measurements using infrared cameras monitoring the RF antennas. The heat flux related to the acceleration of electrons in front of the LHCD grills (LHCD active) and to the acceleration of ions in RF-rectified sheath potentials (ICRH active) were calculated. Complementary results on the heat flux related to fast ions (ICRH active with a relatively low magnetic field) are also reported in this paper.

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.

Plasma facing components (PFCs) with tungsten (W) armor materials for DEMO divertor require a high heat flux removal capability (at least 10 MW/m2 in steady-state conditions). The reference divertor PFC concept is a finger with a tungsten tile as a protection and sacrificial layer brazed to a thimble made of tungsten alloy W – 1% La2O3 (WL10). Defects may be located at the W thimble to W tile interface. As the number of fingers is considerable (>250,000), it is then a major issue to develop a reliable control procedure in order to control with a non-destructive examination the fabrication processes. The feasibility for detecting defect with infrared thermography SATIR test bed is presented. SATIR is based on the heat transient method and is used as an inspection tool in order to assess component heat transfer capability. SATIR tests were performed on fingers integrating or not the complex He cooling system (steel cartridge with jet holes). Millimeter size artificial defects were manufactured and their detectability was evaluated. Results of this study demonstrate that the SATIR method can be considered as a relevant non-destructive technique examination for the defect detection of DEMO divertor fingers.

For steady state (magnetic) thermonuclear fusion devices which need large power exhaust capability and have to withstand heat fluxes in the range 10–20 MW m−2, advanced Plasma Facing Components (PFCs) have been developed. The importance of PFCs for operating tokamaks requests to verify their manufacturing quality before mounting. SATIR is an IR test bed validated and recognized as a reliable and suitable tool to detect cooling defaults on PFCs with CFC armour material. Current tokamak developments implement metallic armour materials for first wall and divertor; their low emissivity causes several difficulties for infrared thermography control. We present SATIR infrared thermography test bed improvements for W monoblocks components without defect and with calibrated defects. These results are compared to ultrasonic inspection. This study demonstrates that SATIR method is fully usable for PFCs with low emissivity armour material.

In this paper water-cooled divertor concepts based on tungsten monoblock design identified in previous studies as candidate for fusion power plant have been reviewed to assess their potential and limits as possible candidates for a DEMO concept deliverable in a short to medium term ("conservative baseline design"). The rationale and technology development assumptions that have led to their selection are revisited taking into account present factual information on reactor parameters, materials properties and manufacturing technologies. For that purpose, main parameters impacting the divertor design are identified and their relevance discussed. The state of the art knowledge on materials and relevant manufacturing techniques is reviewed. Particular attention is paid to material properties change after irradiation; phenomenon thresholds (if any) and possible operating ranges are identified (in terms of temperature and damage dose). The suitability of various proposed heat sink/structural and sacrificial layer materials, as proposed in the past, are re-assessed (e.g. with regard to the possibility of reducing peak heat flux and/or neutron radiation damages). As a result, potential and limits of various proposed concepts are highlighted, ranges in which they could operate (if any) defined and possible improvements are proposed. Identified missing point in materials database and/or manufacturing techniques knowledge that should be uppermost investigated in future R&D activities are reported. This work has been carried out in the frame of EFDA PPPT Work Programme activities. © 2013 Elsevier B.V.

International audience; The present paper deals with the surface heat flux estimation with embedded thermocouples (TC) in a Plasma Facing Component (PFC) of the WEST Tokamak. A 2D nonlinear unsteady calculation combined with the Conjugate Gradient Method (CGM) and the adjoint state is achieved in order to estimate the time evolution of the heat flux amplitude and decay length lambda_q. The method is applied on different synthetic measurements in order to evaluate the accuracy of the method. The synthetic measurements are generated with realistic values of lambda_q and magnitudes as those expected for ITER.

15th International Conference on Plasma-Facing Materials and Components for Fusion Applications (PFMC), Aix en Provence, FRANCE, MAY, 2015; International audience; In the next fusion devices, all the plasma facing components will consist of bulk tungsten or tungsten coating on carbon. This paper focuses on the behaviour of tungsten coated on carbon fibre composite designed for the WEST project (Bucalossi et al 2011 Fusion Eng. Des. 86 684-688) under intensive thermal cycling delivered by an electron beam. The use of scanning electron microscope has allowed in particular, the observation of several pore lines inside the coating. These pore lines have different aspects depending on the observed zone according to the localisation of the electron beam, accentuated lines with more numerous enlarged pores in zone exposed to the electron beam. An analogous trend is also observed for JET tungsten-coated samples under similar thermal cycles despite their different properties due to an alternative manufacturing method of the substrate. A systematic and attentive comparison on the coating changes after the application of the electron beam heating is presented. The observed comportments as the formation of the pore lines or the pore shapes are assumed to be inherent to simultaneous diffusion processes. In association with the pore line formation, a migration of the carbon substrate towards the surface is presumed and discussed.

15th International Conference on Plasma-Facing Materials and Components for Fusion Applications (PFMC), Aix en Provence, FRANCE, MAY, 2015; International audience; A thermal wall model is developed for the SolEdge2D-EIRENE edge transport code for calculating the surface temperature of the actively-cooled vessel components in interaction with the plasma. This is a first step towards a self-consistent evaluation of the recycling of particles, which depends on the wall surface temperature. The proposed thermal model is built to match both steady-state temperature and time constant of actively-cooled plasma facing components. A benchmark between this model and the Finite Element Modelling code CAST3M is performed in the case of an ITER-like monoblock. An example of application is presented for a SolEdge2D-EIRENE simulation of a medium-power discharge in the WEST tokamak, showing the steady-state wall temperature distribution and the temperature cycling due to an imposed Edge Localised Mode-like event.

International audience; The peak power flux densities from plasma ion bombardment on tokamak armor surfaces can deviate from geometric-optical projection when the ion thermal Larmor radius becomes comparable to the characteristic size of the surface relief (gaps, misalignments, shaping). We present a fast and reliable tool to estimate the effect of finite ion Larmor radius on heat load deposition for complex 3D elements. The two possible geometrical design options foreseen for WEST tungsten monoblocks are asymmetric chamfer shaping and no shaping on top of monoblocks. This paper presents an evaluation of peak power flux densities for a WEST ITER-like tungsten monoblock plasma facing component which is vertically misaligned with respect to the other components of a maximum expected value of +0.3 mm, for steady state convected plasma power with ion temperature ranging from 10 to 500 eV. With the +0.5 mm asymmetric chamfer, leading edges between plasma facing components are shadowed and there is no significant difference between the optical calculations and the full ion orbit model. For an unshaped and +0.3 mm misaligned monoblock, when considering the full ion orbit model, the heat load on the exposed edge is reduced by a factor of 3-5 compared to the geometric-optical projection. Energy is found to be deposited over several millimetres on top of the monoblock just after the leading edge which prevents the tungsten armor from melting. (C) 2016 Elsevier B.V. All rights reserved.

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.

In order to evaluate the option to start the ITER operation with a full tungsten (W) divertor, the EU-DA launched an extensive R&D program. It consisted in its initial phase in the high heat flux (HHF) testing of W mock-ups and medium scale prototypes up to 20 MW/m2 in the AREVA FE 200 facility (F). Critical heat flux (CHF) experiments were carried out on the items which survived the above thermal fatigue testing. After 1000 cycles at 10 MW/m2, the full W Plasma Facing Components (PFCs) mock-ups successfully sustained either 1000 cycles at 15 MW/m2 or 500 cycles at 20 MW/m2. However, some significant surface melting, as well as the complete melting of a few monoblocks, occurred during the HHF thermal fatigue testing program representative of the present ITER requirements for the strike point region, namely 1000 cycles at 10 MW/m2 followed by 1000 cycles at 20 MW/m2. The results of the CHF experiments were also rather encouraging, since the tested items sustained heat fluxes in the range of 30 MW/m2 in steady-state conditions.

For ITER divertor Plasma Facing Components (PFCs), tens of thousands of armor/heat sink interfaces will be produced by the industry. Statistically, there is a probability that interfaces with defects be delivered. The defect detection with Non Destructive Techniques (NDT) is then a major challenge. NDT should provide a detectability threshold below the critical defect size. For a defect located all along the axial length of a component, the critical defect size at interface is about 50° for W monoblock (resp. 6 mm for W flat tile). It is defined with thermo-mechanical fatigue behaviour under 10 MW m−2 for W monoblock (resp. 5 MW m−2 for W flat tile). The purpose of this paper is to study the armor/heat sink defect detection of tungsten components (flat tile and monoblock geometries) with SATIR test bed (Infrared thermography NDT). We demonstrate that SATIR is a relevant NDT to detect defect of W components.

Thermal fatigue behaviour of repaired monoblocks was assessed from High Heat Flux (HHF) tests up to 20 MW m−2 on 11 components. Among these components, 8 monoblocks were repaired (2 CFC and 6 tungsten). These components were manufactured by two EU industries: ANSALDO Ricerche and PLANSEE. Non destructive examination was performed on SATIR thermography test bed before and after HHF tests. SATIR results show that repaired monoblocks have a good thermal exhaust capability before HHF tests. For all monoblocks, no degradation of thermal properties was noticed during cycles at 10 MW m−2. After hundreds of cycles at 20 MW m−2, two W repaired monoblock melted. Post-HHF SATIR examination revealed a degradation of thermal properties which is systematic for W melted monoblocks and non-systematic for W repaired ones. For CFC repaired monoblocks, no damage was observed up to 20 MW m−2. For the first ITER divertor set, specifications for the pre-qualification are that CFC (Resp. W) components have to sustain in steady state 1000 cycles at 10 MW m−2 (Resp. 3 MW m−2) followed by 1000 cycles at 20 MW m−2 (Resp. 5 MW m−2). For the first ITER divertor set, the repair process is validated for CFC and W monoblocks.

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.

An extensive development programme has been carried out in the EU on high heat flux components within the ITER project. In this framework, a full-scale vertical target (VTFS) prototype was manufactured with all the main features of the corresponding ITER divertor design. The fatigue cycling campaign on CFC and W armoured regions, proved the capability of such a component to meet the ITER requirements in terms of heat flux performances for the vertical target. This paper discusses metallographic observations performed on both CFC and W part after this intensive thermal fatigue testing campaign for a better understanding of thermally induced mechanical stress within the component, especially close to the armour–heat sink interface.

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.

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.

The commissioning of plasma-facing component fields needs advanced non-destructive methods to detect in a reliable way the defects, which can impair the component performances and/or integrity during operation. Within this framework, CEA developed a dedicated non-destructive examination method based on active infrared thermography (SATIR facility) to inspect the bonding between armour material and metallic heat sink. Used with successful in the commissioning of the toroidal pump limiter of Tore Supra, this technique was applied in the frame of the pre-series activities of the Wendelstein 7-X high heat flux divertor elements to assess the bonding quality of the delivered components. This paper presents the methodology adopted to define an acceptance criterion based on SATIR test bed possibly applied for a serial inspection of the Wendelstein 7-X elements. Using the well-tried acceptance test based on the DTref_max parameter, the new method includes advanced data post-processing techniques from thermo-signal SATIR and a data merging method to help the decision-making and to optimise the reliability of the binary response expected for a final decision in terms of acceptance test.

The components used in fusion devices, especially high heat flux Plasma Facing Components (PFC), have to withstand heat fluxes in the range of 10–20 MW/m2. So, they require high reliability which can be only guaranteed by accurate Non Destructive Examinations (NDE). The SATIR test bed operating at Commissariat a l’Energie Atomique (CEA) Cadarache performs NDE using transient infrared thermography sequence which compares the thermal response of a tested element to a Reference element assumed to be defect free. The control parameter is called DTref max. In this paper, we present two innovative image processing techniques of the SATIR signal allowing the qualification of a component without any Reference element. The first method is based on a spatial image autocorrelation and the second on the resolution of an Inverse Heat Conduction Problem (IHCP) using a BEM (Boundary Element Method) technique. After a validation step performed on numerical data, these two methods have been applied to SATIR experimental data. The results show that these two techniques allow accurate defect detection, without using a Reference tile. They can be used in addition to the DTref max, for the qualification of plasma facing components.