Steady-state Tokamak Research Articles

XPS post-mortem analysis of plasma-facing units extracted from WEST after the C3 (2018) and C4 (2019) campaigns

Four monoblocks coming from one ITER-like plasma-facing unit from the Q3B sector of the lower divertor, named as monoblock (MB)3, MB9, MB20, and MB30, were exposed to the deuterium and helium plasma mixture during the C3 (2018) and C4 (2019) campaigns of the Tungsten Environment in Steady-state Tokamak (WEST), followed by a detailed ex-situ X-ray photoelectron spectroscopy investigation. The surface and in-depth chemistry of the tungsten monoblocks indicated the formation of a re-deposited mixture in the deposition-dominated area of the divertor, thicker than 218 and 172 nm for MB3 and MB9, respectively. The redeposition layer was dominated by a mixture of boron carbides accompanied by tungsten carbides in MB3, while in MB9, the redeposition layer was dominated by tungsten borides. The remaining two monoblocks, MB20 and MB30, were collected from the erosion region and showed similar chemical behavior with a blended mixture of oxidized and metallic tungsten followed by boron carbides within a 50 nm depth range. Boron fixation in the layers is an expected consequence of the boronizations used during the operation, but the chemical status of redeposited elements was characterized for the first time in this work.

Heating characteristic of electron Bernstein wave using high-field side X-mode injection in QUEST

Abstract A comparative investigation was conducted to assess the impact of high-field side (HFS) extraordinary (X) mode (HFS X) injection and low-field side (LFS) ordinary (O) mode (LFS O) injection of 8.2 GHz radio frequency (RF) power in the Q-shu University experimental steady-state spherical tokamak. In the case of the HFS X injection, the ratio of electron plasma frequency and RF frequency f pe / f RF was 1.3, indicating an overdense plasma, whereas in the LFS O injection, this ratio was 0.9. Distinctive features emerged in the electron temperature and density profiles between the two injection scenarios. In the HFS X injection, the profiles peaked between the fundamental electron cyclotron resonance (ECR) layer and the upper hybrid resonance layer. Conversely, in the LFS O injection, the peaks were situated near the 1st ECR layer. This implies effective excitation of the electron Bernstein wave (EBW) in the case of the HFS X injection, with the wave power being absorbed through the doppler-shifted ECR of the EBW. Density and temperature oscillations were observed only in the start-up phase of the HFS X injection, which may correlate with the presence of the left-hand cutoff of the X-mode. These findings will be contributed to understand the distinctive characteristics associated with plasma heating through EBW excitation in the HFS X injection.

Electric field effects during disruptions

Tokamak disruptions are associated with breaking magnetic surfaces, which makes magnetic field lines chaotic in large regions of the plasma. The enforcement of quasi-neutrality in a region of chaotic field lines requires an electric potential that has both short and long correlation distances across the magnetic field lines. The short correlation distances produce a Bohm-like diffusion coefficient ∼Te/eB and the long correlation distances aT produce a large scale flow ∼Te/eBaT. This cross-field diffusion and flow are important for sweeping impurities into the core of a disrupting tokamak. The analysis separates the electric field in a plasma into the sum of a divergence-free, E→B, and a curl-free, E→q, part, a Helmholtz decomposition. The divergence-free part of E→ determines the evolution of the magnetic field. The curl-free part enforces quasi-neutrality, E→q=−∇→Φq. Magnetic helicity evolution gives the required boundary condition for a unique Helmholtz decomposition and an unfortunate constraint on steady-state tokamak maintenance.

Wall conditions in WEST during operations with a new ITER grade, actively cooled divertor

Future fusion reactors like ITER and DEMO will have all-tungsten (W) walls and long pulses. These features will make wall conditioning more difficult than in most of the existing devices. The W Environment Steady-state Tokamak (WEST) is one of the few long pulse (364 s) fusion devices with actively cooled W plasma-facing components in the world. WEST is a unique test bed to study impurity migration and plasma density control via reactor relevant wall conditioning techniques. The phase II of WEST operations began in 2022, after the installation of a new lower divertor, now entirely equipped with actively cooled, ITER grade, W monoblocks. After pump down, we baked WEST between 90 °C and 170 °C for ∼2 weeks. After 82.5 h at 90 °C and 33 h at 170 °C, vacuum conditions were stable with a vessel pressure of 6x10-5 Pa and mass spectra dominated by H2 molecules. While at 170 °C, we performed ∼40 h of D2 glow discharge cleaning (GDC) and ∼5 h of glow discharge boronization (GDB), using a 15 %-85 % B2D6-He mix and a total boron mass of ∼12 g. This was the very first GDB at such high temperature for WEST. The whole wall conditioning sequence led to a ∼10 times reduction of the H2O signal as well as to a ∼3 times reduction of the O2 signal, according to mass spectra. Once back to 70 °C, the vessel pressure was 5.5x10-6 Pa and plasma restart was seamless with ∼30 s cumulated over the very first 5 pulses and an Ohmic radiated power fraction Frad = 0.6, showing successful conditioning of the new ITER grade divertor. The effect of the first, ‘hot’ GDB faded with a characteristic cumulative injected energy of 2.45 GJ and saturation towards Frad ∼0.8. After 1.4 h and 7.5 GJ of cumulative plasma time and injected energy, we carried out a second GDB, this time at 70 °C. This ‘cold’ GDB initially led to a much lower Ohmic Frad = 0.3–0.4 but the effect lasted ∼7 times less, with a characteristic cumulative injected energy of 0.37 GJ. At the end of the campaign, we cumulated ∼3h and ∼30 GJ through repetitive, minute long pulses without any boronization. Throughout this 4-weeks-long experiment, Frad in the 4 MW heating phase evolved only marginally (from 0.5 to 0.55). This increase is mostly due to the build-up of re/co-deposited layers on both lower divertor targets.

DIII-D research to provide solutions for ITER and fusion energy

The DIII-D tokamak has elucidated crucial physics and developed projectable solutions for ITER and fusion power plants in the key areas of core performance, boundary heat and particle transport, and integrated scenario operation, with closing the core-edge integration knowledge gap being the overarching mission. New experimental validation of high-fidelity, multi-channel, non-linear gyrokinetic turbulent transport models for ITER provides strong confidence it will achieve Q ⩾ 10 operation. Experiments identify options for easing H-mode access in hydrogen, and give new insight into the isotopic dependence of transport and confinement. Analysis of 2,1 islands in unoptimized low-torque IBS demonstration discharges suggests their onset time occurs randomly in the constant β phase, most often triggered by non-linear 3-wave coupling, thus identifying an NTM seeding mechanism to avoid. Pure deuterium SPI for disruption mitigation is shown to provide favorable slow cooling, but poor core assimilation, suggesting paths for improved SPI on ITER. At the boundary, measured neutral density and ionization source fluxes are strongly poloidally asymmetric, implying a 2D treatment is needed to model pedestal fuelling. Detailed measurements of pedestal and SOL quantities and impurity charge state radiation in detached divertors has validated edge fluid modelling and new self-consistent ‘pedestal-to-divertor’ integrated modeling that can be used to optimize reactors. New feedback adaptive ELM control minimizes confinement reduction, and RMP ELM suppression with sustained high core performance was obtained for the first time with the outer strike point in a W-coated, compact and unpumped small-angle slot divertor. Advances have been made in integrated operational scenarios for ITER and power plants. Wide pedestal intrinsically ELM-free QH-modes are produced with more reactor-relevant conditions, Low torque IBS with W-equivalent radiators can exhibit predator-prey oscillations in T e and radiation which need control. High-β P scenarios with q min > 2, q 95–7.9, β N > 4, β T–3.3% and H 98y2 > 1.5 are sustained with high density ( n¯ = 7E19 m−3, f G–1) for 6 τ E, improving confidence in steady-state tokamak reactors. Diverted NT plasmas achieve high core performance with a non-ELMing edge, offering a possible highly attractive core-edge integration solution for reactors.

Development of pulsed plasma operation scenario and required conditions in JA DEMO

We have developed the pulsed plasma operation scenarios for JA DEMO, a design concept of the steady-state tokamak demonstration reactor, to clarify controls of the current profile and power required for the operation. We compare the scenarios when injecting electron cyclotron waves only and both neutral beam and electron cyclotron waves for external heating and current drive. We demonstrate current profile control that maintains the minimum value of the safety factor above one and avoids creating the local minima in the safety factor profile and power control by argon seeding that maintains the fusion power constant at the desired value and reduces the heat load on the divertor, performing long-time integrated modeling simulations. We clarify the conditions of the heating and current drive system and impurity injection system required for such control. The dependence of power control on argon anomalous transport coefficients is investigated. We have the prospect of maintaining the fusion power of 1 GW for more than two hours, i.e. obtaining the required plasma performance determined using a systems code.

Prospects of negative triangularity tokamak for advanced steady-state confinement of fusion plasmas in MHD stability consideration

The steady-state confinement, beta limit, and divertor heat load are among the most concerned issues for toroidal confinement of fusion plasmas. In this work, we show that the negative triangularity tokamak has promising prospects to address these issues. We first demonstrate that the negative triangularity tokamak generates the filed line rotation transform more effectively. This brings bright prospects for the advanced steady-state tokamak scenario. Given this, the MHD stability and equilibrium confinement of negative triangularity tokamak are investigated. We point out that the negative triangularity configuration with a broad pressure profile is indeed more unstable for low-n magnetohydrodynamic modes than the positive triangularity case so that the H-mode confinement can hardly be achieved in this configuration, where n is the toroidal mode number. Nevertheless, we found that the negative triangularity configuration with high bootstrap current fraction, high poloidal beta, and peaked pressure profiles can achieve higher normalized beta for low-n modes than the positive triangularity case. In a certain parameter domain, the normalized beta can reach about twice the extended Troyon limit, while the same computation indicates that the positive triangularity configuration is indeed constrained by the Troyon limit. This shows that the negative triangularity tokamaks are not only favorable for divertor design to avoid the edge localized modes but also can have promising prospects for advanced steady-state confinement of fusion plasmas in high beta.

Design of a multi-energy soft X-ray diagnostic for profile measurements during long-pulse operation in the WEST tokamak

The W Environment in Steady-state Tokamakwas designed and built to test ITER-like tungsten plasma facing components in a long pulse (∼1000 s) scenario. Recently, a multi-energy soft x-ray diagnostic (MESXR) was installed in the WEST (W Environment in Steady-state Tokamak), to understand the sources, transport and confinement of high-Z impurities. The purpose of this work is to describe the engineering challenges posed by the long pulse scenario for the integration of the MESXR diagnostic and how they were addressed and solved. The calculations of vacuum and thermal stress as well as heat transfer are presented and discussed.

Lossy compression of infrared videos in WEST (W Environment in Steady-state Tokamak).

During the 2019 C4 experimental campaign of the WEST (W Environment in Steady-state Tokamak) (France), the infrared diagnostic produced more than seven terabytes of uncompressed video data. Constraints on the computer infrastructure required for storage, backup, and especially offline access to infrared videos made the use of a compression algorithm mandatory. This paper proposes an innovative method to compress infrared videos with controlled temperature precision. This compression method is based on a controlled averaging of the video that maximizes the compression potential of standard lossless video codecs such as H264/AVC or HEVC. The combination of the loss introduction algorithm and the H264/AVC lossless video codec obtains the best compression ratio in the range of 8 to 41 with a maximum temperature error of 2 °C. This method also outperforms the JPEG-LS algorithm in terms of compression ratio and image quality for the same temperature precision.

Scaling laws of the plasma velocity in visco-resistive magnetohydrodynamic systems

We consider a visco-resistive magnetohydrodynamic modelling of a steady-state incompressible tokamak plasma with a prescribed toroidal current drive, featuring constant resistivity η and viscosity ν. It is shown that the plasma velocity root-mean-square behaves as ηf(H) as long as the inertial term remains negligible, where H stands for the Hartmann number H≡(ην)−1/2, and that f(H) exhibits power-law behaviours in the limits H≪1 and H≫1. In the latter limit, we establish that f(H) scales as H1/4, which is consistent with numerical results.

Comparison of electron temperature and density measured by helium line intensity ratio and Thomson scattering methods in ECH spherical tokamak plasma

The electron temperature and density profiles in the midplane of a spherical tokamak plasma produced by electron cyclotron heating (ECH) in Q-shu University experiment with steady-state spherical tokamak (QUEST) are measured by the helium line intensity ratio method. The measured profiles are compared with those obtained by the Thomson scattering method, and the measured temperatures and densities are found to agree within factors of ∼2 and ∼6, respectively. Taken together with the previous results of comparisons performed in the scrape-off layers of several toroidal devices, the same degree of agreement between the helium line intensity ratio method and other methods is obtained in the ranges of 7–100 eV for temperature and 4 × 1016–1 × 1019 m−3 for density.

Design and analysis of actively-cooled, edge-transport diagnostic for long-pulsed operation in WEST

Next step fusion devices that will operate in steady-state will require complex plasma-facing components (PFCs) that can survive the harsh environment over long timescales not common in current devices. This will require robust plasma facing surfaces that are integrated with active cooling systems. In a collaboration between CEA and ORNL, a plasma-interacting diagnostic is being designed for the W Environment in Steady-state Tokamak (WEST) in Cadarache, France which requires plasma-facing protection like those needed for steady-state PFCs. This integrated diagnostic studies edge transport and impurity migration within WEST, and is planned to include imbedded temperature Langmuir probe sensors, as well as removeable sample slots for ex-situ surface analysis of plasma-material interactions. The entire assembly is expected to move into the plasma edge for periods up to 1000 s having an energy removeable capability of ∼6 kW and seeing a peak heat flux of more than 7 MW/m2. The assembly compliments WEST high fluence campaigns that plan for multiple 1000 s pulses. Within these specifications, the assembly will require a refractory metal plasma facing surface and integral cooling in order to function within the limited space allotted for such diagnostics. Because of the limited space and linear actuator needs, additive manufacturing of the high heat flux working end of the assembly is being considered which could allow for precision cooling-channels and lighter weight designs. Conceptual design along with simulation and analysis results will be presented for this complex diagnostic with novel PFCs.

Flux pumping studies and Ti peaking by q-profile shaping using ECCD in ASDEX Upgrade

On ASDEX Upgrade a prioritized ‘advanced Tokamak’ program has been run during the last two experimental campaigns, focusing on the effect of non-standard current profiles on the behavior of high beta plasmas. Two lines of non-standard (i.e. non-sawtoothing) q-profiles are followed: (1) plasma self-organization of centrally flat q-profiles (qmin ≈ 1) via ‘flux pumping’, here based on a naturally occurring continuous (1,1) mode and (2) externally shaped q-profiles with qmin > 1, allowing for more experimental freedom of the q-profile to be established. The advantages and requirements of both concepts are discussed. The invited presentation at the workshop focuses on specific effects of Electron Cyclotron Current Drive (ECCD). With respect to flux pumping it could be shown that increasing the plasma beta raises the flux pumping capability. More central coECCD can be redistributed if beta is increased. This is in line with qualitative theory. For the case with qmin > 1, it is shown that small variations of the ECCD profile can modify the peaking of the central ion temperature. For larger radii no significant changes of the kinetic profiles have been observed as the q-profile is changed. Hfactors have not exceeded 1.2, well below the assumptions of several models for steady state tokamak operation. To achieve these changes of q at larger radii we have used ctr-ECCD in the plasma center, increasing the ohmic current globally. The net effect of central ctr-ECCD and increased ohmic current is a strong off-axis current drive, allowing H-modes with Te ≈ Ti ≈ 8 keV and n¯e ≈ 6 · 1019m−3 ≈ 0.5 nGW at q95 = 4 to be run stationary with qmin ≈ 1.4. The design of these discharges was strongly supported by inter-shot model based optimization.

U-Net for temperature estimation from simulated infrared images in tokamaks

Surface temperature measurement is critical for the safety and proper operation of nuclear fusion reactors such as tokamaks, which operate at high temperature [100–3600 °C] in the presence of complex and unknown surface reflectance properties, causing a reflected flux that disturbs measurement interpretations. This paper describes a numerical approach based on machine learning techniques to estimate the surface temperature of in-vessel components from infrared images, despite the presence of reflections and the unknown emissivity of materials. Our contributions are two-fold: for the first time in the infrared domain, we generate a huge dataset of simulated images thanks to a novel GPU implementation of a Monte Carlo infrared ray tracer. This implementation offers the great advantage of enabling the generation of a very large amount of synthetic images for different plasma scenarios. Hence the second contribution of this paper: we propose an infrared inverse lighting model, based on a Convolutional Neural Network (CNN) trained with the images generated by our ray tracer. The proposed approach is applied on a numerical prototype of W Environment Steady state Tokamak (WEST), including complex simulated thermal scenes. Using synthetic test data, and without providing the surface optical properties, the mean temperature error predicted by the CNN on ITER-like wide angle view of WEST is evaluated to 9% with standard plasma scenarios and 22% in case of exotic thermal scenes.

Long plasma duration operation analyses with an international multi-machine (tokamaks and stellarators) database

Combined high-fusion performance and long-pulse operation is one of the key integration challenges for fusion energy development in magnetic devices. Addressing these challenges requires an integrated vision of physics and engineering aspects with the purpose of simultaneously increasing time duration and fusion performance. Significant progress has been made in tokamaks and stellarators, including very recent achievement in duration and/or performance. This progress is reviewed by analyzing the experimental data (109 plasma pulses with a total of 3200 data points, i.e. on average 29 data per pulse) provided by ten tokamaks (in alphabetical order: Axially Symmetric Divertor Experiment Upgrade, DIII-D, Experimental Advanced Superconducting Tokamak, Joint European Torus, JT-60 Upgrade, Korea Superconducting Tokamak Advanced Research, tokamak à configuration variable, Tokamak Fusion Test Reactor, Tore Supra, W Environment in Steady-State Tokamak) and two stellarators (Large Helical Device and Wendelstein 7-X) expanding the pioneering work of Kikuchi (Kikuchi M. and Azumi M. 2015 Frontiers in Fusion Research II: Introduction to Modern Tokamak Physics (Springer)). Data have been gathered up to January 2022 and coordination has been provided by the recently created International Energy Agency-International Atomic Energy Agency international Coordination on International Challenges on Long duration OPeration group. By exploiting the multi-machine international database, recent progress in terms of injected energies (e.g. 1730 MJ in L-mode, 425 MJ in H-mode), durations (1056 s in L-mode, 101 s in H-mode), injected powers, and sustained performance will be reviewed. Progress has been made to sustain long-pulse operation in tokamaks and stellarators with superconducting coils, actively cooled components, and/or with metallic walls. The graph of the fusion triple products as a function of duration shows a dramatic reduction of, at least two orders of magnitude when increasing the plasma duration from less than 1 s to 100 s. Indeed, long-pulse operation is usually reached in dominant electron-heating modes at reduced density (current drive optimization) but with low ion temperatures ranging from 1 to 3 keV for discharges above 100 s. Difficulties in extending the duration may arise from coupling high heating powers over long durations and the evolving plasma-wall interaction towards an unstable operational domain. Possible causes limiting the duration and critical issues to be addressed prior to ITER operation and DEMO design are reported and analyzed.

Anomaly classification by inserting prior knowledge into a max-tree based method for divertor hot spot characterization on WEST tokamak

The divertor of WEST (W Environment in Steady-state Tokamak) is the main component for plasma control and exhaust. It receives high heat fluxes, which can cause damage to plasma facing units above the allowable heat flux. Improving the operation safety on the actively cooled tungsten divertor is being researched in place at WEST, toward providing divertor monitoring solution for ITER. Divertor operation safety relies on detecting, monitoring, and classifying all hot spots on the divertor surface using infrared (IR) cameras. In this paper, a method based on max-tree representation and attributes of IR images is used to classify normal from abnormal strikelines on the divertor. The proposed method requires only high-level prior knowledge of abnormal temperatures and divertor structure but does not require any labeled data, unlike existing methods, such as support vector machines (SVMs) or convolutional neural networks (CNNs). The max-tree classifier method is tested on real IR images from the WEST tokamak and shows that 88% of hot spots are accurately classified with a small enough calculation duration that can be performed between two pulses.

The design and test of the electrical insulation break for the EAST Divertor primary heat transfer system

Experimental Advanced Superconducting Tokamak (EAST) is one of the important steady-state divertor Tokamak physical experiment platforms in the world. The main functions of the high-temperature and high-pressure Divertor Primary Heat Transfer System (DIV PHTS) are to provide high-pressure cooling water for the divertor to remove the heat load deposited on the divertor under the plasma operation mode and to provide high-temperature nitrogen gas for the divertor system to achieve vacuum condition during the baking mode. The stability and reliable operation of the DIV PHTS play an important role in the EAST tokamak plasma physics experiment. The high-temperature and high-voltage electrical insulation break are one of the important components of the system. The main function of the Electrical Insulation Break (EIB) is to break the closed circuit of the heat transfer system to avoid eddy current during plasma discharge affecting the magnetic field distribution characteristics of the tokamak. On the other way, it also needs to meet the requirement, there is no leak of the cooling water and nitrogen gas during cooling and baking operation mode. This paper proposes the design of an electrical insulating break (EIB) to meet the design requirements of EAST tokamak at first. Then the electric field analysis is carried out by using a finite element model to evaluate its electrical insulation characteristics. And the sealing structural characteristics are verified by applying the finite element method under two design modes. Finally, the sealing test result under the high-temperature baking and high-pressure cooling modes are presented to validate the design.

New method for reshaping of the WEST Lower Hybrid launcher to reach long pulse operation

The aim of WEST (tungsten-W Environment in Steady-state Tokamak) experiments is to master long pulses (1000s) and submit the ITER-grade tungsten divertor to power fluxes of 10 MW/m2. The lower hybrid current drive (LHCD) system, composed of a fully active multijunction (FAM-LH1) launcher and a passive active multijunction (PAM-LH2) launcher, with a total capability of 7 MW/1000s, is essential to meet these goals. The LH1 launcher was reshaped before the start of WEST to match the new plasma shape and thus improve its coupling capability. During the 2019–2021 experimental campaign, long pulse operation (∼ 55 s) was successfully and routinely performed with PLH = 3 MW and good LH coupling properties. Thanks to infrared monitoring during long pulses, the new toroidal shape of the LH1 launcher was found to avoid overheating on the front face edges. It also highlighted the existence of hot spots on the LH2 launcher's front face edges, limiting its power capability in high power and long pulse scenarios. A reshaping inside the vacuum vessel of the LH2 launcher was therefore performed during the shutdown in 2021. This paper describes the reshaping methods used for the two launchers and the tools developed for the reshaping of LH2 inside the tokamak.

First 3D modelling of tungsten erosion and migration in WEST discharges adopting a toroidally non-symmetric wall geometry

Numerical analysis is a useful tool to investigate tungsten (W) sources and transport across plasma in W Environment Steady state Tokamak (WEST) plasma discharges, as it highlights physical mechanisms not always directly observable in experiments. Modelling activities were performed to study W erosion from WEST plasma-facing components (PFCs), as well as W migration through the plasma. For the first time, it was adopted a toroidally asymmetric wall geometry consisting of toroidally localized objects representing WEST antennas. To simulate WEST boundary plasma, 3D non-axisymmetric SOLEDGE transport simulations were performed with simplifying assumptions (pure deuterium plasma, a fluid model for neutrals). Results were then used as background for ERO2.0 runs to model W migration. On the sides of the toroidally localized objects, two thin stripes modelled WEST W antenna protections. Simulations suggest that particles eroded from the antennas protections may dominate the core W contamination in the analysed wall configuration. The findings suggest that these 3D non-axisymmetric models may be needed on a broader range of plasma conditions and wall configurations to accurately model the W migration in WEST.

Software platform for imaging diagnostic exploitation applied to edge plasma physics and real-time PFC monitoring

In current tokamaks, imaging diagnostics (visible/infrared) have become increasingly used for real-time control of the temperature of the Plasma Facing Components (PFC) and for off-line physical analysis. Developing a full acquisition and control system based on a network of multiple cameras is a complex task requiring advanced software tools. Furthermore, such systems produce a large quantity of data encouraging end users to share tools and codes for data access and analysis.In this paper, we introduce a software platform, named Thermavip, dedicated to these tasks. The main contribution of this software is to gather under the same framework different functionalities for (1) firm real-time video acquisition and PFC monitoring, (2) offline data access and display, (3) signal and video processing applied to plasma physics, (4) video annotation and thermal events database management. This software platform is widely used on the WEST (W Environment in Steady-state Tokamak) tokamak (France) for offline study of infrared (IR) and almost all WEST diagnostics data, and online display of IR videos and temperature time traces for critical components. Thermavip platform is also used on the HADES (High heAt LoaD tESt) facility (France) and the EAST (Experimental Advanced Superconducting Tokamak) tokamak (China) for offline signal analysis, and on the Wendelstein 7-X (W7-X) stellarator (Germany) for the IR diagnostic acquisition, PFC real-time (RT) monitoring, online display and offline video analysis. In addition, Thermavip provides a full featured video annotation tool used on both WEST and W7-X to build and manage thermal events datasets. These annotations are used as input dataset to train deep-learning models, like Region Based Convolutional Neural Networks, for automatic detection and classification of thermal events based on IR movies.The Thermavip framework is composed of a unique C++ Software Development Kit (SDK) providing high-level classes for offline multi-sensor data analysis, firm real-time processing and online visualization of these sensor data. Its strength comes from its unique component block architecture, allowing to build multicore and distributed processing pipelines for both offline and firm real-time applications. Thermavip architecture relies on a versatile plugin mechanism to extend its functionalities using C++ code or Python scripts, facilitating the integration on new machines.