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This paper summarizes the emissivity measurements performed on the plasma-facing units (PFU) of the WEST lower divertor during the first phase of WEST running with a mix of actively cooled ITER-like PFUs made of bulk tungsten (W) and inertially cooled PFUs made of graphite with a coating of tungsten. In situ assessments of the emissivity and laboratory measurements after removing the W-coated graphite and ITER-grade PFUs from the WEST device are shown. The measurements exhibit a complex pattern with strong emissivity variation as a function of space and time mainly explained with the variation of magnetic equilibrium (strike point location) as well as the plasma performances during the experimental campaigns. The exposed ITER-grade PFU exhibits sharp spatial variation of the emissivity from 0.05 to 0.85 at a monoblock scale (12 mm) at the transition of the erosion (strike point location) and deposition (next to the strike point location) areas on the high-field side. On the low-field side, the emissivity varies from 0.12 at the strike point location, to 0.2 a few cm away in the low-field-side direction. This emissivity range after exposure is much higher than the emissivity variation of unexposed PFU with emissivity from 0.09 to 0.15. In situ observation performed on the W-coated graphite PFU shows a rapid evolution, typically a few pulses, of the emissivity in the inner and outer strike point location. The whole spatial distribution is discussed as well as its variation due to the plasma operation from the start-up of WEST to the removal of the W-coated graphite components.

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International audience; During the 2019 experimental campaign, the WEST tokamak was partially equipped with ITER-like plasma facing units (PFUs) made of discrete monoblocks (MBs) in one of the twelve lower divertor sectors. The magnetic field lines can enter the gaps between two MBs and strike their leading edges (LEs) with near normal incidence, leading to high localized heat flux, temperature and thermomechanical stress during both stead-state operation and transients. Exposed leading edges are a particular matter of concern because of potential crack formation, recrystallization or even melting of tungsten. During the 2019 experimental campaign in WEST, five of them were misaligned (vertical misalignment h = 0.30 ± 0.1 mm) with their poloidal leading edges exposed to the plasma heat flux. A medium wavelength IR filter (MWIR: 3.9 ± 0.1 µm) was installed in the very high resolution infrared system, featuring a submillimeter spatial resolution (~0.1mm/pixel). This system has a temperature detection threshold of T threshold, BB ≈ 250°C. In this paper, thermal analysis will be presented with a specific focus on overheating of poloidal and toroidal edges using post-mortem measured emissivity maps. The study reveals an unexpected hot spot in the top LE corner of misaligned PFUs at the outer strike point (OSP). Photonic calculations were performed with different emissivity on top and lateral surfaces, in order to consider the complex reflectance characteristics in the toroidal gaps. During this experiment, the Larmor radius (0.35 mm) is comparable to the gap size. Thus, the helical trajectory of the particles may significantly affect the heat load distribution in the edges vicinity. Heat load simulations were performed with ion orbit modeling that consider the Larmor radius, to study its impact on the thermal distribution of the MB. The photonic simulations showed that it was possible to observe a false hot spot at the top LE corner, with both optical approximation (OA) and ion orbit (IO) modeling, due to reflections and different emissivity values.

International audience; A real time Wall Monitoring System (WMS) is used on the WEST tokamak during the C4 experimental campaign. The WMS uses the wall surface temperatures from 6 fields of view of the Infrared viewing system. It extracts the raw digital data from selected areas, converts it to temperatures using the calibration and write it on the shared memory network being used by the Plasma Control System (PCS). The PCS feeds back to actuators, namely the injected power from 5 antennae's of the lower hybrid and ion cyclotron resonance radiofrequency (RF) heating systems. WMS activates feed back control 63 times during C4, which is 14% of the plasma discharges. It activates mainly as the result of a direct RF loss to the upper divertor pipes. The feedback control maintains the wall temperature within the operation envelope during 97% of the occurrences, while enabling plasma discharge continuation. The false positive rate establishes at 0.2%. WMS significantly facilitated the operation path to high power operation during C4, by managing the technical risks to critical wall components.

International audience; Infra-red (IR) thermography is a widely used tool in fusion devices to monitor and to protect the plasma-facing component (PFC) from excessive heat loads. However, with the use of all-metal walls in fusion devices, deriving surface temperature from IR measurements has become more challenging. In this paper, an overview of infra-red measurements in the metallic tokamaks WEST and ASDEX Upgrade (AUG) is reported and the techniques carried out in the modeling and experimental fields to deal with this radiative and fully reflective environment are presented. Experimental characterizations of metallic samples in laboratory and experiments in WEST and AUG reveal that the behavior of both the emission and the reflectance can vary significantly with surface roughness, machining process and as the plasma operation progress. In parallel, the development of a synthetic IR diagnostic has allowed for a better interpretation of the IR images by assessing the reflection patterns and their origin. This has also proven that small-scale change in the emission pattern of beveled PFC can be confused with abnormal thermal events. Numerical solutions to evaluate the contribution of the reflections associated with a variable emissivity in a fully reflective and radiative environment are finally presented.

In this paper, a method was developed to estimate in-situ the emissivity of the W-coated graphite divertor tiles in the WEST tokamak. This method is based on the double heating method and take advantages of the divertor temperature increase after successive plasma experiments due to the inertial behavior of the plasma facing components. Photonic calculations have been used to disentangle the emitted and the reflected parts in the measured radiances from the infrared system. The uncertainty as well as the robustness of the method have been investigated thanks to the wide IR and thermocouple coverage in the WEST divertor. The results show strong variation of the emissivity along the divertor W surfaces with a factor 4 variation after the experimental campaigns including 18.3 GJ and about 21 000 s of cumulated injected energy and duration, respectively. Finally, the implication of a non-uniform emissivity on heat flux estimation from IR measurements is discussed, showing that non-uniform emissivity must be considered to obtain an accurate heat flux decay width.

International audience; Twelve ITER-like plasma-facing units made of tungsten were exposed in the WEST tokamak divertor, with three PFUs significantly overexposed to plasma heat flux: one sharp-edged PFU (vertical misalignment h = 0.8 mm) and two chamfered PFUs (h = 0.6 mm and 0.3 mm, respectively). This paper describes the first temperature analysis obtained with a very high spatial resolution infrared camera (pixel size ~ 0.1 mm) on the misaligned PFU edges and shows the consistency obtained on the parallel heat flux derived from these measurements. The analysis is focused on the hottest areas of the PFU misaligned leading edges, since the temperature detection threshold of the VHR camera is high (Tthreshold,BB ≈ 370°C). The heat flux parallel to the magnetic field lines is assessed by matching the toroidal temperature profile in the vicinity of the leading edge with 3D finite element modelling. The tungsten emissivity assumed in this study is 0.6, consistent with laboratory measurements for damaged PFUs. For the three PFUs studied (with different vertical misalignments, incident angles, geometries), the derived parallel heat flux is similar, and consistent with independent measurements by a Fiber Bragg grating embedded in a graphite PFU at another toroidal location, giving confidence in future experiments using the same settings.

The WEST (Tungsten Environment for Steady State Tokamak) platform aims at testing ITER like divertor targets in an integrated tokamak environment, to minimize risks for procurement and operation. After 4 years of construction, WEST is now operational, with the first plasma breakdown achieved in December 2016. To operate long plasma discharge in WEST, infrared thermography is a key device, which enables a safe operation by means of a real time surface temperature monitoring, while providing essential data for various physics studies. For WEST, the IR thermography system has been deeply renewed, to match with the new tokamak configuration, and to fulfil the new measurement requirements. It consists of a set of 5 different diagnostics: 1) 7 endoscopes located in upper ports viewing the whole lower divertor and the 5 heating devices 2) a tangential endoscope located in a median port, providing a wide angle view of the vacuum vessel, and in particular of the upper divertor and first wall protections 3) a submillimetre resolution view on the divertor strike points from an upper port endoscope 4) 2 additional views of the divertor and of the heating devices. The paper describes the new IR diagnostics configuration (spatial covering, IR cameras, optics, data processing hardware and software), the main design options, manufacturing and installation issues, and some measurement performances.

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

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.

For the long-pulse high-confinement discharges in tokamaks, the equilibrium of plasma requires a contact with the first wall materials. The heat flux resulting from this interaction is of the order of 10 MW/m 2 for steady state conditions and up to 20 MW/m 2 for transient phases. The monitoring on surface temperatures of the plasma facing components (PFCs) is a major concern to ensure safe operation and to optimize performances of experimental operations on large fusion facilities. Furthermore, this measurement is also required to study the physics associated to the plasma material interactions and the heat flux deposition process. In tokamaks, infrared (IR) thermography systems are routinely used to monitor the surface temperature of the PFCs. This measurement requires an accurate knowledge of the surface emissivity. However, and particularly for metallic materials such as tungsten, this emissivity value can vary over a wide range with both the surface condition and the temperature itself, which makes instantaneous measurement challenging. In this context, the multi-spectral infrared method appears as a very promising alternative solution. Indeed, the system has the advantage to carry out a non-intrusive measurement on thermal radiation while evaluating surface temperature without requiring a mandatory surface emissivity measurement. In this paper, a conceptual design for the multi-spectral infrared thermography is proposed. The numerical study of the multi-channel system based on the Levenberg-Marquardt (LM) nonlinear curve fitting is applied. The numerical results presented in this paper demonstrate the design allows for measurements over a large temperature range with a relative error of less than 10%. Furthermore, laboratory experiments have been performed from 200 °C to 740 °C to confirm the feasibility for temperature measurements on stainless steel and tungsten. In these experiments, the unfolding results from the multi-channel detection provide good performance on temperature measurement, which supports our numerical evaluation and demonstrates the potential feasibility for metallic surface high temperature measurement with this method.

The aim of WEST [1] [2] is to master long pulses (> 30s and up to 1000s) and submit ITER like tungsten divertor tiles to power fluxes of 10 MW/m2. To reach these goals, the Tore Supra Lower Hybrid Current Drive system (LHCD) constituted by two launchers with an injection power capability of 7MW/1000s has been upgraded to meet WEST new environment (diverted plasma, new launchers positions). In particular, the LH1 Full Active Multi-junction launcher (FAM) has been toroidally reshaped to match the new magnetic configuration. The poloidal curvature of the plasma has also been optimized to guarantee good coupling properties of the waves. During the last experimental campaign (October-December 2018), we obtained 30s repetitive pulses with a LH power of 2.7 MW (more than 2 hours of plasma during the experimental campaign) and 43 discharges with a total LH power from 4-5.3 MW. This paper describes the results obtained in term of coupling for both launchers. Thanks to Infra-Red (IR) monitoring it also assesses the reshaping of LH1 in terms of thermal loads due to supra-thermal electrons on the launcher edges. During the upcoming experimental campaign, the IR coverage will further improve allowing assessing the LH2 Passive Active Multi-junction launcher (PAM) thermal loads.

International audience; The WEST platform aims at testing ITER like W divertor targets in an integrated tokamak environment. To operate long plasma discharges, IR thermography is required to monitor the main plasma facing components by means of real time surface temperature measurements, while providing essential data for various physics studies. To monitor the new divertor targets, the WEST IR thermography protection system has been deeply renewed, to match the new tokamak configuration. It consists of 7 endoscopes located in upper ports viewing the whole lower divertor and the 5 heating devices. Electronic devices and computers allow data storage of ≈3 Gb/s IR images and real time video frames processing at 50 Hz rate, to ensure the protection of the main plasma facing components during plasma discharges by a feedback control of the power injected by the heating systems. Each endoscope provides 2 views covering 2 divertor sectors of 30°(toroidally) and 1 view of a heating antenna. Each optical line is composed of a tight entrance window followed by a head objective which forms an image transported through the endoscope by a series of 4 optical relays and mirrors, up to a camera objective. Finally, 12 IR cameras specially developed for WEST environment capture the thermographic data, at the wavelength of 3.9 μm, with a 640 × 512 pixels frame size. The paper describes the design constraints and diagnostic technologies: optics, mechanics, electronics, hard & software, cameras. Tvhe laboratory characterization procedures (Modulation Transfer Function, slit response, calibration), and the measurement performance results are given (spatial resolution, temperature threshold). Finally, first results obtained during experimental campaigns in WEST are presented.

International audience; ITER (https://www.iter.org) will take advantage of tungsten (W) actively cooled Plasma Facing Units (PFU) in the lower divertor. These PFUs will receive a steady state plasma heat flux of 10 MW/m² and up to 20 MW/m² in slow transients, pushing these components to their limit. For machine protection reasons and for the study of the plasma wall interactions, temperature measurements are foreseen with Mid Wave Infra-Red (MWIR: 3-5 µm) thermography systems to cover most of ITER chamber including the divertor (tungsten material) and first wall (beryllium material) [1]. Absolute temperature measurements from 70 °C, the cooling temperature of the PFUs, up to 3500 °C (W melting temperature at 3422 °C), are considered. Those measurements require the knowledge of the emissivity of the tungsten PFU surfaces as function of the temperature and wavelength used for the IR monochrome thermography system. This paper summarizes the emissivity measurements performed on tungsten surfaces in High Heat Flux (HHF) test beds and the WEST tokamak [2] and uses the experimental emissivity values to compute the expected uncertainties in ITER (T/T) using monochrome and bicolor thermography techniques. Results show that monochrome technique is not able to fulfill the T/T < 10% requirement, while bicolor technique is able to reduce the temperature uncertainty below 10%. Laboratory bicolor thermography measurements using a filter wheel (including 6 MWIR interference filters) have been successfully performed up to 830°C with temperature uncertainty T/T < 3%. In the WEST tokamak, the two dimensions emissivity map is varying by a factor of five within 27 cm along a PFU, from 0.13 to 0.65, showing that the monochrome IR thermography will have difficulties to stay within the ITER requirements T/T < 10% from 70 °C up to 3500 °C.

The Faraday Screen (FS) is the plasma facing component of ITER ion cyclotron heating antennas shielding. The requirement for the high heat exhaust, and the limitation of the temperatures to minimize strain and thus offer sufficient resistance to fatigue, imply the need for high conductivity materials and a high cooling flow rate. The FS bars are constructed by a hipping process involving beryllium tiles, a pure copper layer, a copper chrome zirconium alloy for the cooling channel and a stainless steel backing strip. Two FS bars small scale mock-ups were manufactured and tested under high heat flux. They endured 15,000 heating cycles without degradation under nominal heat flux, and revealed growing flaws when the heat flux was progressively augmented beyond. In this case, the ultrasonic test confirms a strong delamination of the Be tiles.

International audience; A new Infrared diagnostic has been developed by CEA-IRFM and installed in the WEST tokamak to measure surface temperature of the actively cooled W-monoblocks components as foreseen for the ITER Divertor, with a very high spatial resolution of 100µm. The goals are to investigate the effects of the shaping of these components on the heat load deposition pattern, the evolution of pre-damaged components specifically introduced in WEST, the behavior of the leading edges regarding the assembling tolerances between adjacent monoblocks, and finally to contribute to the specification assessment of the ITER divertor units. In WEST, each Plasma Facing Unit is composed of 35 W-monoblocks of individual surface of 28x12mm. To analyze heat load pattern and phenomena on such tiny surfaces, the leading edges and in the narrow gaps between monoblocks (400-500µm), a 100µm spatial resolution is required. Then, a Very High spatial Resolution (VHR) infrared diagnostic has been specially developed at CEA-IRFM. The VHR operates at 1.7µm wavelength to take advantage of the dynamic of the signal for the temperature range (400 to 3600°C). The VHR infrared diagnostic is now operational above the divertor sector made of actively cooled W-monoblocks and graphite inertial components with W coating. This paper gives a description of the diagnostic.

The WEST project (Tungsten (W) Environment in Steady State Tokamak) is targeted at minimizing risks in support of the ITER divertor strategy. Part of the machine protection system will be based on Short Wave InfraRed (SWIR) thermography diagnostic which consists in monitoring and controlling in real time the power load on the plasma facing components through the surface temperature measurements. The inherent availability objective of such a machine protection diagnostic is essential for WEST operation. A functional analysis of the IR system from highest level main functions down to basic operational functions has been developed. The availability of the initial design has been assessed by making a RAMI (Reliability, Availability Maintainability and Inspectability) analysis. Despite mitigation actions to reduce the frequency of potential failures and their time to repair, the availability required by the project could not be reached. With the aim of achieving the availability target, a recommendation was made to consider an alternative design. This paper presents a RAMI analysis of the IR thermography diagnostic whose results have led to modifying the design of antennas protection system to a more available system as required by the WEST project.

The WEST (Tungsten (W) Environment in Steady state Tokamak) project is a partial rebuild of the Tore Supra tokamak to make it an X-point metallic environment machine aimed at testing ITER technologies in relevant plasma environment. For the safe operation of the WEST tokamak, infra-red (IR) thermography is a crucial diagnostic as it is a sound and reliable way to detect hotspots or abnormal heating patterns on the plasma facing components (PFCs). Thus WEST will be fitted with middle/short-IR (1.5–2 μm or 3–5 μm) cameras in the upper port plugs to get a full view of the critical PFCs (in particular the new lower divertor) and radio-frequency (RF) heating antennas and one camera at the equatorial level to monitor the new upper divertor and the first wall. This paper describes the design of the up-to-date optical system along with the hydraulic analysis and the thermal and mechanical finite element analysis conducted to ensure adequate heat extraction capabilities. Boundary conditions and simulation results will be presented and discussed as well as technological solutions retained.

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.

To operate long plasma discharge in tokamak equipped with actively cooled plasma facing components (PFC), infrared (IR) thermography is a key diagnostic. Indeed IR data are used for both PFC safety monitoring, to avoid material degradation and water leak, and various physics studies on plasma-wall interaction. The IR monitoring is becoming even more crucial with today metallic PFCs. This is the case for the WEST project (Tungsten (W) Environment for Steady State Tokamak), which aims at installing a W divertor in Tore Supra (TS), in order to operate the 1st tokamak with a full W actively cooled divertor in long plasma discharges. The IR thermography system for the WEST project described in this paper will consist of a set of 3 different diagnostics: (1) Six cameras located in upper ports viewing the full W divertor, which reuse a part of the existing diagnostic of TS. (2) Five novel views located behind the inner protection panels for the antennas monitoring, based on an innovative imaging fibers bundle technology. (3) A tangential wide angle view located in a median port, for the upper divertor and first wall monitoring.

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.

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.

Actively cooled plasma facing components (PFCs) have to exhaust high heat fluxes from plasma radiation and plasma–wall interaction. Critical heat flux (CHF) event may occur in the cooling channel due to unexpected heat loading or operational conditions, and has to be detected as soon as possible. Therefore it is essential to develop means of monitoring based on precursory signals providing an early detection of this destructive phenomenon, in order to be able to stop operation before irremediable damages appear. Capabilities of CHF early detection based on acoustic techniques on PFC mock-ups cooled by pressurised water were already demonstrated. This paper addresses the problem of the detection in case of flow rate reduction and of flow dilution resulting from multiple plasma facing units (PFU) which are hydraulically connected in parallel, which is the case of ITER divertor. An experimental study is launched on a dedicated mock-up submitted to heat loads up to the CHF. It shows that the measurement of the acoustic waves, generated by the cooling phenomena, allows the CHF detection in conditions similar to that of the ITER divertor, with a reasonable number of sensors. The paper describes the mock-ups and the tests sequences, and comments the results.

Analyses of erosion and deposition over the toroidal pump limiter of Tore Supra were performed combining 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-homogeneous, in agreement with a deuterium impoverishment of the deep layers.

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.

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).

Lock-in thermography was applied to determine the thermal contact conductance of a W-layer (140 μm) on a CFC-substrate. A lock-in thermography system together with a pulse repetition rate Nd:YAG laser (1064 nm, 1–500 Hz pulse repetition rate) for layer heating was applied for phase shift measurements on the W-layer. A numerical model for direct phase shift calculations was developed and applied to rapid determination of the Fourier amplitudes and phases of the temperature. Thermal conductance coefficients were obtained by comparing the experimental and simulation phase shifts.

To develop Continuous Wave (CW) high power klystrons for fusion experiments, calorimetric matched loads absorbing the RF power are necessary. To test and adjust the parameters of the new klystrons TH2103C [1] able to produce 700 kW/CW at a frequency of 3.7 GHz for upgrading the RF power available in Tore Supra LHCD transmitter, SPINNER GmbH has successfully developed, in collaboration with the CEA, and manufactured a water load capable to absorb the RF power with a Voltage Standing Wave Ratio (VSWR)

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 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 μm 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.

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.

Acoustic monitoring is a method under development to indicate the occurrence of critical heat flux (CHF) events on plasma facing components exposed to high heat fluxes (HHFs) from plasma wall interaction, in order to be able to stop plasma operation before irremediable damages appear. It is a non-intrusive promising method thanks to the property of acoustic waves to propagate in channels and to the CHF acoustic precursory indicators which have been observed in several previous HHF experimental studies. This method is on a preliminary assessment stage and HHF experiment on relevant mock-up is an opportunity to collect and analyse data for improving the method efficiency. This paper deals with the post-processing of acoustic signals recorded on a European ITER divertor qualification prototype, on which an unexpected loss of flow accident occurred and caused a cooling channel rupture. The acoustic data have been analysed in the same way as for a CHF scenario, in order to seek precursory indicators of this kind of event.

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.

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.

Up-to-date development of actively cooled high heat flux (HHF) plasma facing components (PFC) prototypes only allows reduced margins with regards to the ITER thermal requirements. Additionally, perfect quality cannot be ensured along series manufacturing: the presence of flaws which impair the heat transfer capability of the component, in particular at the interface armour/heat sink appears to be statistically unavoidable. In order to ensure a successful series production, a qualification methodology of actively cooled high heat flux plasma facing components is proposed. Secondly, advanced non-destructive techniques developed for HHF PFC commissioning are detailed with definition of acceptance criteria. Finally, innovative diagnostics for in situ monitoring during plasma operations or tokamak shutdowns are investigated in order to prevent immediate damage (safety monitoring); or evaluate component degradation (health monitoring). This work takes into account the relevance to Tore Supra, and is applied to W7X and ITER Divertor HHF PFC.

A non destructive testing technique – so called modulated photothermal thermography or lock-in thermography – has been set-up for plasma facing components examination. Reliable measurements of phase contrast were obtained on 8 mm carbon fiber composite (CFC) armoured W7-X divertor component with calibrated flaws. A 3D finite element analysis allowed the correlation of the measured phase contrast and showed that a 4 mm strip flaw can be detected at the CFC/copper interface.

Plasma facing components (PFC) for modern steady state fusion devices have to withstand heat fluxes up to 20 MW m−2. In case of internal defects in the components, a critical heat flux event can occur at an incident heat flux lower than expected in a defect-free component. It is therefore essential to develop in-vessel means for on line safety monitoring to prevent PFC destruction, and health monitoring during shutdown to control PFC damage. Two methods are developed in this paper, covering these topics.

Combined magnetron sputtering and ion implantation (CMSII) is a deposition technique involving simultaneous magnetron sputtering and high energy ion bombardment of the coating during its growth. A high voltage pulse discharge (U=40 kV, τ=20 μs, f=25 Hz) is superposed over the magnetron deposition and in this way, positive ions are accelerated to the components to be coated, bombarding initially the substrate and then the coating itself. In the framework of the ITER-like wall project this method was applied to produce nanostructured W coatings on the carbon fibre composite (CFC) substrate. These coatings have been characterized in terms of adhesion, thickness, structure and resistance to high thermal loads (up to 23.5 MW m−2). Based on the results of these tests, which are presented in this paper, CMSII technology was selected for coating about 1100 tiles with a 10 μm tungsten layer for the JET first wall and divertor.