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This is a PDF file of an article that has undergone enhancements after acceptance, such as the addition of a cover page and metadata, and formatting for readability, but it is not yet the definitive version of record. This version will undergo additional copyediting, typesetting and review before it is published in its final form, but we are providing this version to give early visibility of the article. Please note that, during the production process, errors may be discovered which could affect the content, and all legal disclaimers that apply to the journal pertain.; International audience; WEST is a full metallic tokamak with an extensive set of diagnostics for heat load measurements. In this paper, heat loads on the lower divertor of WEST are investigated using two independent methods. A first method relies on the thermal inversion of temperature measurements from arrays of thermal sensors embedded a few millimeters below the surface, while the second consists in the inversion of black body surface temperatures measured by infra-red (IR) thermography. The challenge of IR based temperature measurements in the full metal environment of WEST is addressed through a simplified model, allowing to correct for global reflections and low surface emissivities of tungsten surfaces. A large database ( L-mode discharges) is investigated. It is found that the energy absorbed by an outer divertor tile during a plasma discharge is closely estimated by the two diagnostics, over a large set of experimental conditions. A similar match is also found for the peak heat flux value on the outer target. The toroidal modulation of target heat loads by magnetic ripple is found to be consistent with the geometrical projection of a parallel heat flux component. Additionally, the heat flux channel width at the target is found to scale linearly with the magnetic flux expansion as expected. These observations give confidence in the robustness of the data from both diagnostics, and confirm the simple geometrical rules at use in the description of heat flux deposition on divertor targets. However, it is shown that the heat flux channel width estimated from infra-red thermography is about three times lower than the width estimated from embedded measurements, which is still under investigation.

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.

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.

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.

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.

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.

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.

The ongoing design of the ITER Ion Cyclotron Heating and Current Drive system (20 MW, 40-55 MHz) is rendered challenging by the wide spectrum of requirements and interface constraints to which it is subject, several of which are conflicting and/or still in a high state of flux. These requirements include operation over a broad range of plasma scenarios and magnetic fields (which prompts usage of wide-band phased antenna arrays), high radio-frequency (RF) power density at the first wall (and associated operation close to voltage and current limits), resilience to ELM-induced load variations, intense thermal and mechanical loads, long pulse operation, high system availability, efficient nuclear shielding, high density of antenna services, remote-handling ability, tight installation tolerances, and nuclear safety function as tritium confinement barrier. R&D activities are ongoing or in preparation to validate critical antenna components (plasma-facing Faraday screen, RF sliding contacts, RF vacuum windows), as well as to qualify the RF power sources and the transmission and matching components. Intensive numerical modeling and experimental studies on antenna mock-ups have been conducted to validate and optimize the RF design. The paper highlights progress and outstanding issues for the various system components. (C) 2013 ITER Organization. Published by Elsevier B.V. All rights reserved.

26th Symposium on Fusion Technology (SOFT), Porto, PORTUGAL, SEP 27-OCT 01, 2010; International audience; Infrared (IR) photothermal techniques are candidates for in situ characterisation of tokamak plasma facing components (PFC) surfaces, by means of an external thermal excitation coupled with an IR temperature measurement. Among these techniques, the laser lock-in thermography (LLIT) uses a modulated laser excitation which gives 2 major advantages: enhancement of signal to noise ratio and emissivity independence, which is a plus when the components have various and unpredictable surface quality. With this method, it is possible to develop a process, which could be used remotely, either mounted onto an in situ inspection device (articulated arm) or in a PFC test bed. This paper presents the results obtained with a continuous modulated laser heat source on particular samples (W coating on CFC substrate, C layer on graphite substrate). The identification of the experimental data with a theoretical model allows a quantitative characterisation of the layers. (C) 2011 Elsevier B.V. All rights reserved.

19th International Conference on Plasma-Surface Interactions in Controlled Fusion Devices (PSI), Univ Calif, Gen Atom, San Diego, CA, MAY 24-28, 2010; International audience; Understanding heat flux deposition processes is essential for the design of the plasma facing components allowing reliable high power steady state plasma operations. Due to the magnetic configuration of Tore Supra, the deposited heat flux is found to be particularly sensitive to tile misalignment in a region where magnetic field lines graze the surface (incident angle typically

Future fusion experiments will rely on tungsten armour tile for their plasma facing components. In order to sustain steady state operation, the components need to be cooled through an attachment to a heat sink. All current reference concepts rely on contact bonds, unfavourable for long-term application (high temperature service, cycle fatigue, thermal shocks). Three routes toward the development of thick tungsten bonds are presented here, namely functionally graded tungsten copper assembly, electron beam welding of tungsten, and coating processes. All present favourable prospects, and tend to indicate that a thick bond is possible with tungsten. Dedicated programs as well as industrial implication are however required if such concepts are to be used actually for the fabrication of large components series.