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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 100 μm. 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 28 × 12 mm. To analyze heat load pattern and phenomena on such tiny surfaces, the leading edges and in the narrow gaps between monoblocks (400–500 μm), a 100 μm spatial resolution is required. Then, a Very High spatial Resolution (VHR) infrared diagnostic has been specially developed at CEA-IRFM. The VHR operates at 1.7 μm wavelength to take advantage of the dynamic of the signal for the temperature range (400–3600 °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. [ABSTRACT FROM AUTHOR]

The emissivity of metallic materials is low and varies with temperature and wavelength inducing errors on surface temperature measurements. High heat flux experiments on beryllium were carried out to investigate the thermal behavior of bulk Be tiles. Thermal modeling aiming at determining the surface and bulk temperatures have been performed using ANSYS®. A Be tile was exposed to heat flux with power density ranging between 1 and 7 MW/m 2 . Surface temperatures were measured using an infrared camera in the 3–5 μm range and two-color pyrometers, one at short wavelengths (1.5–1.7 μm) and one at mid IR range wavelengths (2–4 μm) range. Both the IR camera and two-color pyrometers do not provide accurate temperature measurements on melted Be due to changes in the emissivities and emissivity ratio induced by surface modifications. [ABSTRACT FROM AUTHOR]

An imaging spectrometer using a sapphire prism as dispersing element has been conceived at Tore Supra for the spectral range of 1–4 μm. It measures simultaneously at various wavelengths the temperature on distributed high heat-flux elements under plasma impact with 36 optical fibers, four of which are ZrF4 fibers. It employs an InSb focal plane array detector (256*320 pixels) behind a silicon filter and a ZnS window yielding a dynamic range of 200–1500 °C with 20 ms temporal resolution. The fiber transmission and the spatial variation of gain and background of the camera are calibrated using a light source with integrating sphere. With a blackbody source one determines the nonlinearity of the average gain and controls its stability during operation. The spectral dispersion of about 24 nm/pixel is determined with interference filters and controlled with a spectral lamp. The measurements at various wavelengths allow to determine the temperature distribution in the field of view. [ABSTRACT FROM AUTHOR]

Since tungsten (W) has been selected as material for the ITER divertor target the knowledge of the emissivity in the infrared (IR) wavelength range becomes mandatory. A dedicated setup has been developed to measure the emissivity in the wavelength range 1.7–4.75 μm and the temperature domain 200–850 °C. The paper presents the emissivity measurements on W samples coming from actively cooled bulk W components (ITER-like) and W-coated graphite component of the WEST divertor. W samples with damaged surfaces generated by transient heat load as those observed in fusion machines, including micro-cracks and crack network, are also investigated. The dependence on wavelength, temperature and surface state are shown and discussed. [ABSTRACT FROM AUTHOR]

One of the challenges that International Thermonuclear Experimental Reactor (ITER) will face during plasma operation is to determine the absolute divertor surface temperature of the actively water-cooled (70 °C, 3 MPa) tungsten plasma-facing units (PFUs) to ensure their integrity. The expected steady-state heat flux up to 10 MW/m2 is close to the operational limit and so one of the goals of the thermographic system is to make reliable measurements (wavelength band: 3.5– $4.5~\mu \text{m}$) with relatively low error bars to avoid tungsten melting and material damages. The tungsten emissivity is low and dependent on wavelength, temperature, and surface state (roughness, cracks, oxidation, and erosion/deposition processes) which can evolve during the time along with plasma operation. An error on the absolute emissivity can lead to large absolute temperature errors, and consequently to either a reduction of the operational window to fulfill safety limits, or an increased risk regarding the integrity of the components. For the past years, a number of emissivity measurements have been performed with ITER-like PFU mock-ups made of different grades of tungsten, with different techniques, from different material makers, and from different manufacturers. The overall emissivity discrepancy is large, indicating that tungsten emissivity is a major issue for temperature measurement and wall protection. The accuracy of the temperature measurement is investigated for both monocolor and bicolor IR thermography techniques. We present laboratory tests performed with an IR camera equipped with a rotating filter wheel for bicolor temperature measurement. It shows that the bicolor technique can reach high precision $\Delta \text {T}/\text {T} < 10{\%}$ at high temperature assuming the emissivity ratio is constant regarding the two selected wavelengths. [ABSTRACT FROM AUTHOR]

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 ITER 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 °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. [ABSTRACT FROM AUTHOR]

• Unique 0.1 mm/pixel IR thermography diagnostic for plasma-wall interaction studies. • Improved accuracy and reliability on field of view positioning. • Operational on WEST tokamak since 2020. • Diagnostic is accurate to the monoblock. The Very High spatial Resolution (VHR) infrared thermography diagnostic developed by CEA-IRFM is installed in the WEST tokamak to measure surface temperature of the actively cooled W-monoblocks (MB) of the Plasma Facing Units (PFUs), which are similar to the International Thermonuclear Experimental Reactor (ITER) ones. It is a major diagnostic to investigate the behavior of PFU during experimentation thanks to its very high spatial resolution (0.1 mm / pixel), movable field of view (FoV) and the wide-range temperature measurement. In particular, it can study the evolution of monoblocks mounted in the tokamak, the influence of the shaping of these components submitted to high heat load and the behavior of the leading edges regarding the assembling tolerances between adjacent monoblocks. Finally, such diagnostic directly contributes to the specification assessment of the ITER divertor units. This diagnostic was operational during the 2019 WEST experimentation and its enhanced design for the 2020 WEST experimentation campaign. In order to monitor 30 × 12 mm monoblocks temperature, the VHR IR diagnostic is able to move its narrow FoV (64 × 51 mm) to specific targets on the observable divertor area (360 × 420 mm). Position and focus steering are achieved through motor-controlled mirrors and lenses operated by a dedicated Labview application on a remote unit. Mirror positioning occurs between pulses under normal WEST operation conditions, and can withstand thermal radiation and disruptions during pulses. Repeatability, hysteresis, mechanical components reliability are critical parameters to provide accurate positioning (<1 mm) of the FoV. The tuning and calibration of these actuators in accordance with WEST environment is crucial to guarantee the best position of the FoV and optical performances. This paper gives the FoV positioning implementation details and performance of the VHR in WEST. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

A tangential infrared thermography system is installed in the WEST tokamak to observe the thermal scene of the first wall and the divertor components through a wide-angle tangential view. The goal is to provide and monitor the heat load deposition on the plasma facing components. About one-sixth of the chamber is observed, including sectors of the baffle, the lower and upper divertors, one inner bumper, vertical displacement event (VDE) protections, one ICRH antenna, and other parts of the wall. It is used for real time machine protection by monitoring temperature thresholds in delimited region of interest, and for analysis of normal or specific heat load events during operation such as VDE, ELM, disruption or runaways. This wide-angle view uses one aspherical and one on-axis plane mirrors, a sapphire window and three lenses for the objective of the camera. The optical line is optimized for two wavelengths 1.7 and 4 µm. The field of view is 60° on a 512×640 pixels Focal Plane Array. The endoscope is fully actively cooled. The thermal scene is complex to interpret given the fully metallic and radiative environment and the uncertainties on the emissivity which is angular-dependent and changes with the surface properties. This involves significant inaccuracy on the recovery of the real temperature. In this context, an interpretation by modeling approach is better suited, based on ray-tracing simulations taking into account the optical properties of materials. For instance, this allowed discriminating reflections patterns from real thermal events in the wide-angle view. A description of the wide-angle infrared diagnostic and its performances is presented as well as experimental measurements obtained in the WEST Tokamak. [ABSTRACT FROM AUTHOR]

In nuclear fusion devices, such as Tore Supra, the plasma facing components (PFC) are in carbon. Such components are exposed to very high heat flux and the surface temperature measurement is mandatory for the safety of the device and also for efficient plasma scenario development. Besides this measurement is essential to evaluate these heat fluxes for a better knowledge of the physics of plasma-wall interaction, it is also required to monitor the fatigue of PFCs. Infrared system (IR) is used to manage to measure surface temperature in real time. For carbon PFCs, the emissivity is high and known (ε ~ 0.8), therefore the contribution of the reflected flux from environment and collected by the IR cameras can be neglected. However, the future tokamaks such as WEST and ITER will be equipped with PFCs in metal (W and Be/W, respectively) with low and variable emissivities (ε ~ 0.1-0.4). Consequently, the reflected flux will contribute significantly in the collected flux by IR camera. The modulated active pyrometry, using a bicolor camera, proposed in this paper allows a 2D surface temperature measurement independently of the reflected fluxes and the emissivity. Experimental results with Tungsten sample are reported and compared with simultaneous measurement performed with classical pyrometry (monochromatic and bichromatic) with and without reflective flux demonstrating the efficiency of this method for surface temperature measurement independently of the reflected flux and the emissivity. [ABSTRACT FROM AUTHOR]

The integration of laser techniques in an embedded system is studied in this paper to propose an embedded system that is able to remove deposited layers, to collect ablated matter, and to perform an "online" characterization of ablated matter in a tokamak. Due to environmental constraints (particularly, pressure), this innovative system, which can be used between two plasma sessions (without pressure and temperature breaks) presents some limitations. We will see that, due to its limitations, the system (currently at a concept stage) will be favorably used to clean small areas or as a diagnostic system. Results presented in this paper are based on studies and simulations. An experimental validation is underway. [ABSTRACT FROM AUTHOR]

Abstract: During Tokamak operation, Plasma Facing Components (PFCs) are subjected to severe interaction with plasma. As a consequence and independently of the PFCs composition, materials eroded and then re-deposited in the form of layers on the surfaces, can flake and produce dusts. These fragile structures are able to trap part of the hydrogenated species (tritium for example) in vessel inventory. For safety reasons, it is mandatory to measure and to control vessel dust and tritium inventory. Up to now, laser techniques are a part of the most promising methods able to solve these ITER open issues. Of special interest are laser systems loaded on a miniature tool that can be attached to a Multi Purpose Deployer (MPD) and used for laser treatments (détritiation and other), for PFCs chemical analysis as well as for micro particles recovery of dust produced during laser ablation. Such a system (Laser Ablation System Kit: LASK) is currently under development at IRFM and the following presentation will describe the current achievements of this project and the perspectives. In this paper, we will present an innovative compact system, which, loaded on a Multi Purpose Deployer, could allow operation in a harsh environment (pressure range from atmospheric to Ultra High Vacuum and temperature up to 120°C). According to the process conditions, different treatments can be performed: at low laser fluence, PFCs thermal treatment will be expected, while at high laser fluence material will be ablated allowing Dust (and T) recovery as well as chemical analysis of material. This “in-line” chemical analysis based on Laser Induced Breakdown Spectroscopy (LIBS) enables the ablation process to be controlled and preserves the substrate integrity. The paper will be focussed on the methodology followed during the LASK development and the method used to determine a laser process window able to remove co-deposited film without damaging the bulk material and taking into account external parameter variation (Multi Purpose Deployer vibrations for example). The first design of the system is proposed that complies with the process requirements and the external constraints. Special emphasis will be given on limitations, and alternatives to these limitations will be proposed. [Copyright &y& Elsevier]

WEST is a full W tokamak with an extensive set of diagnostics for heat load measurements especially in the lower divertor. It is composed by infrared thermography, thermal measurement with thermocouples and fibre Bragg grating embedded few mm below the surface and flush mounted Langmuir probes. A large database including different magnetic equilibrium and input power is investigated to compare the heat load pattern (location, amplitude of the peak and heat flux decay length) on the inner and outer strike point regions: from the first ohmic diverted plasma (obtained during the second experimental campaign C2 in 2018) up to the high power (8 MW total injected) and high energy (up to 90 MJ injected energy in lower single null configuration) experiments performed in the last experimental campaign (C4 in 2019). Concerning the peak location, a good agreement (<1 cm) is obtained between thermal inversions and flush-mounted LP measurements. The peak heat flux from the whole set of diagnostics is in good agreement and mainly in the ±20% range, while the heat flux decay length reported on the target shows significant discrepancy between diagnostics and location in the machine (±40% range). Despite such discrepancy, heat flux decay length at target is found to scale mainly with the magnetic flux expansion through the variation of the X-point height, as expected. The improved plasma performances achieved during C4 enabled to reach significant heat load in the divertor, up to 6 MW m−2 with 4 MW of additional heating power showing the capability to reach the ITER relevant heat load (10 MW m−2 steady state) with about 7 MW of additional power in L-mode discharge. The heat load distribution is clearly asymmetric with a 3/4 and 1/4 distribution on the outer and inner strike point region respectively for the parallel heat flux. [ABSTRACT FROM AUTHOR]

ITER (www.iter.org) will take advantage of tungsten (W) actively cooled plasma facing units (PFUs) in the lower divertor. These PFUs will receive a steady state plasma heat flux of 10 MW m−2 and up to 20 MW m−2 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 the ITER chamber, including the divertor (tungsten material) and first wall (beryllium material) [1]. Absolute temperature measurements from 70 °C and the cooling temperature of the PFUs, up to 3500 °C (W melting temperature at 3422 °C), are considered. Those measurements require knowledge of the emissivity of the tungsten PFU surfaces as a 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 test beds and the WEST tokamak [2] and uses the experimental emissivity values to compute the expected uncertainties in ITER (ΔT/T) using monochrome and bicolor thermography techniques. The results show that the monochrome technique is not able to fulfill the ΔT/T < 10% requirement, while the bicolor technique is able to reduce the temperature uncertainty below 10%. Laboratory bicolor thermography measurements using a filter wheel (including six MWIR interference filters) have been successfully performed up to 830 °C with temperature uncertainty ΔT/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 staying within the ITER requirements ΔT/T < 10% from 70 °C up to 3500 °C. [ABSTRACT FROM AUTHOR]

Twelve ITER-like plasma-facing units made of tungsten were exposed in the WEST tokamak divertor, with three plasma-facing units (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, which is 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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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 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 λ q t , 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. [ABSTRACT FROM AUTHOR]

• We presented the WEST full tungsten actively cooled divertor, mentioning its relevance with the ITER one. • Main features including key elements related to the large-scale production, the reception tests, and the installation into the WEST tokamak are reported. • Main manufacturing issues as well as the reception strategy, in particular the handling of non-conformities, are highlighted. • Results related to the assessment of heat removal capability by IR-thermography and high heat flux test are discussed. The WEST (W - for tungsten - Environment in Steady-state Tokamak) platform aims to support the ITER tungsten divertor technology, manufacturing and long pulse operation. For this purpose, the Tore Supra tokamak has been deeply modified (mainly the installation of a lower and upper divertor coil structure), enabling a restart in 2016 and a first phase of operation (Phase I) until the end of 2020. Following this first phase, a limited number of actively cooled PFUs (Plasma Facing Units), based on ITER divertor technology, were industrially produced and implemented into the WEST vacuum vessel. The second phase of operation (Phase II) will be with a full actively cooled divertor. The first WEST experimental campaign with this new configuration (Phase II) started in 2022. This paper reports on the key elements related to the large-scale industrial production, the reception tests, and the installation of the ITER-like divertor components (including embedded instrumentation) into the WEST tokamak. Main manufacturing issues as well as the reception strategy, in particular the handling of non-conformities, are highlighted, mentioning its relevance with ITER divertor specifications. Results related to the assessment of heat removal capability by IR thermography examination and high heat flux test, as well as the He-leak tightness control are also discussed. [ABSTRACT FROM AUTHOR]

Abstract 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. [ABSTRACT FROM AUTHOR]

Plasma facing components temperature measurement is required to ensure safe high power for long pulse tokamak operation and for physics studies. During the WEST phase 1, a set of 20 thermocouples (TC) and 4 fibers Bragg grating (FBG) have been exposed to five experimental campaigns with 3889 plasmas for a total plasma duration of about 7h15 and 21.2 GJ of cumulated injected energy. This paper presents the performances in terms of response time and high temperature sustaining assess with high heat flux tests performed in the HADES facility. The response time of the TC remains at the same level as before their installation in the divertor. However, the FBG have shown clear increase of their response time depending on the location due to a degradation of the thermal contact of the graphite adhesive and the FBG or component. Finally, one regenerated FBG has been exposed to high temperature, showing that absolute temperature measurement is found reliable up to 1100 °C with the start of some spectrum degradations. Full erasure of the grating is obtained after 10 s at 1200 °C well above the stable temperature of 900 °C of the regenerated FBG. [ABSTRACT FROM AUTHOR]

The main objective of WEST is to study the behavior of the ITER-like divertor monobloc components, test their resistance and ageing under ITER relevant heat loads. For the upcoming phase 2 of WEST, the lower divertor is now fully equipped with the ITER-like actively cooled bulk tungsten Plasma Facing Units (PFU), designed to withstand 10MW/m² in steady state. This divertor is also equipped with an enlarged set of ThermoCouples (TCs x 16) and Fiber Bragg Gratings (FBGs) probes (x5) with 14 measurement spots each to evaluate the heat loading during the experiments. This paper presents the new TC and FBG diagnostics as well as the associated inverse method used to estimate the surface heat flux from the embedded temperature measurements in the divertor. The performance of each diagnostic will be evaluated with synthetic measurements generated with realistic heat flux intensity and distribution. The results of a parametrical analysis shows that each diagnostic is able to estimate the spatial distribution of the expected heat fluxes in WEST. [ABSTRACT FROM AUTHOR]

Abstract: On the limiter of Tore Supra, the heat loads map computed from deconvolution of IR surface temperatures shows good agreement with calorimetry measurements. This experimental heat pattern allows deducing the heat fluxes in the scrape-off layer using a 3D magnetic calculation and assuming only parallel heat transport along field lines. This calculation leads to an underestimation of the power circulating in the edge plasma according to the power balance, similarly to RFA measurements. The comparison between experimental heat loads on the limiter and modelling also shows a spreading of heat fluxes near the LCFS that cannot be explained only by parallel transport in the SOL. [Copyright &y& Elsevier]

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 wavelength-multiplexed 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. [ABSTRACT FROM AUTHOR]

Abstract: In the course of the development of a concept for a spectrally resolving thermography diagnostic for the ITER divertor using optical fibres experimental development work has been carried out in three different areas. Firstly ZrF4 fibres and hollow fibres (silica capillaries with internal AG/AgJ coating) were tested in a Co60 irradiation facility under γ irradiation up to doses of 5kGy and 27kGy, respectively. The ZrF4 fibres suffered more radiation induced degradation (>1db/m) then the hollow fibres (0–0.4db/m). Secondly multi-colour pyroreflectometry is being developed towards tokamak applicability. The emissivity and temperature of tungsten samples were measured in the range of 700–1500°C. The angular working range for off normal observation of the method was 20–30°. The working distance of the method has been be increased from cm to the m range. Finally, encouraging preliminary results have been obtained concerning the application of pulsed and modulated active thermography. [Copyright &y& Elsevier]

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 (∼0.1 mm/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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Abstract: A laser ablation technique has been adapted for removal of the deposited layers on carbon plasma facing components from Tore Supra. This paper describes in detail the experiments performed to adjust the scan parameters and optimize the laser ablation technique. The goal was to reduce the process duration in order to be able to treat large surfaces. To remove layers up to 50μm thick, with an Ytterbium fiber laser of 1mJ, the process duration limit is 20h/m2. The conditions to reach 1h/m2 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. [Copyright &y& Elsevier]