Environment In Steady-state Tokamak Research Articles

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.

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.

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.

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.

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.

3-D Electromagnetic Model of the WEST Tokamak Sector During Plasma Disruptions

WEST is an acronym for tungsten (W) Environment in Steady-state Tokamak. WEST derives from the upgrade of the Tore Supra tokamak, moving from a limiter configuration with carbon plasma-facing components (PFCs) to a divertor configuration with W PFCs. It is located in the south of France (CEA Cadarache) and it assumes a key role as a testing facility for ITER. A computational analysis campaign has been carried out to investigate electromagnetic (EM) loads during plasma disruptions in WEST, the first one with the actual WEST experimental configuration. The applied methodologies as well as the obtained results have been benchmarked and could be employed to assess the behavior of WEST and analogous tokamak machines during plasma disruptions. Specifically, the research campaign has focused on a 20° sector of the WEST tokamak and it has been carried out by means of the ANSYS APDL v19.2 code. The input disruptions have been simulated with the free-boundary plasma equilibrium code FEEQS, while the obtained results mainly relate to the evaluation of eddy currents as well as the magnetic field distribution inside the tokamak. Furthermore, two different configurations of the WEST sector have been analyzed. The analyses in Configuration A are characterized by the same axisymmetric geometrical domain and electrical resistances as in the FEEQS analysis. As a consequence, they allow making a direct comparison between FEEQS and APDL results, benchmarking the model. On the other hand, the analyses of Configuration B are based on an improved configuration, characterized by more realistic 3-D geometry, electrical properties, and loads. Therefore, these last calculations can be employed to evaluate the magnetic field distribution inside the tokamak as well as the EM loads of in-vessel components.

Spatial calibration and synthetic diagnostic of a multi-energy hard x-ray camera at WEST tokamak.

WEST (tungsten environment in steady-state tokamak) is starting operation for the first time with a water-cooled full tungsten divertor, enabling long pulse operation. Heating is provided by radiofrequency systems, including lower hybrid current drive (LHCD). In this context, a compact multi-energy hard x-ray camera has been installed for energy and space-resolved measurements of the electron temperature, the fast electron tail density produced by LHCD and runaway electrons, and the beam-target emission of tungsten at the target due to fast electron losses interacting with the divertor plates. The diagnostic is a pinhole camera based on a 2D pixel array detector (Pilatus 3 CdTe CMOS Hybrid-Pixel detector produced by DECTRIS). The novelty of this diagnostic technique is the detector's capability of adjusting the threshold energy at pixel level. This innovation provides great flexibility in the energy configuration, allowing simultaneous space and energy-resolved x-ray measurements. This contribution details two important steps in the preparation of the diagnostic operation. First, the in-vessel spatial calibration that was carried out with a radioactive source. Second, the synthetic diagnostic is obtained by the suite of codes ALOHA/C3PO/LUKE/R5-X2, which simulates LH wave propagation and absorption, as well as the fast electron bremsstrahlung production.

Real-Time Characterization of Neutron-Induced SEUs in Fusion Experiments at WEST Tokamak During D-D Plasma Operation

We conducted a real-time soft-error rate characterization of CMOS bulk 65-nm static random access memories (SRAMs) subjected to fusion neutrons during deuterium–deuterium (D-D) plasma operation at W–tungsten–Environment in Steady-state Tokamak (WEST). The test equipment, installed in the experimental hall at several locations of the tokamak, was irradiated during machine shots by a flux of particles dominated by primary 2.45-MeV neutrons. Real-time neutron metrology, neutron spectrometry, complementary characterization with monoenergetic neutrons, and Monte Carlo numerical simulations at both material and circuit levels have also been performed to analyze the experimental data. Our results suggest that higher energy neutrons, simultaneously produced by deuterium–tritium (D-T) reactions due to triton burn-up in the D-D plasma, play a significant role in the radiation response of SRAMs for which multiple cell upsets are detected and cannot be attributed to D-D neutrons.

Radiation Characterization of a Backside-Illuminated P-Type Photo-MOS Pixel With Gamma Rays and Fusion-Induced Neutrons

We report on the radiation response of a novel p-type photogate pixel. Sensors are exposed to gamma rays, and fusion neutrons during deuterium plasma operation at W–Tungsten–Environment in Steady-state Tokamak (WEST), showing dark current increase with total dose.

High heat flux testing of newly developed tungsten components for WEST

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

WEST tokamak hard x-ray tomography inversion

Long-pulse steady-state discharges with lower hybrid current drive were successfully obtained in the limiter configuration machine, Tore Supra. The hard x-ray (HXR) diagnostic system in Tore Supra associated with tomographic inversion provided a good approximation of the lower hybrid power deposition profile. The accurate plasma current control through this tomographic reconstruction leads to an improvement of plasma confinement. Tore Supra was upgraded to WEST, Tungsten (W) Environment in Steady-state Tokamak. The plasma geometry was changed to the D-shaped geometry. Therefore, a novel inversion code for WEST has been developed. The method used in the code is based on the Bessel function of the first kind of order zero. The code validation was performed using phantom emissivity profiles and WEST experimental data. It has been found that the line radiation of the tungsten divertor in WEST and noise significantly affect the inversion. The local HXR emissivity profile with acceptable accuracy can be achieved by applying a singular value decomposition filter and suppressing the lines of sight that intersect the divertors.

Calibration of a versatile multi-energy soft x-ray diagnostic for WEST long pulse plasmas.

A compact multi-energy soft x-ray diagnostic is being installed on the W Environment in Steady-state Tokamak (WEST), which was designed and built to test ITER-like tungsten plasma facing components in a long pulse (∼1000 s) scenario. The diagnostic consists of a pinhole camera fielded with the PILATUS3 photon-counting Si-based detector (≲100 kpixel). The detector has sensitivity in the range 1.6-30 keV and enables energy discrimination, providing a higher energy resolution than conventional systems with metal foils and diodes with adequate space and time resolution (≲1 cm and 2 ms). The lower-absorption cut-off energy is set independently on each one of the ∼100 kpixels, providing a unique opportunity to measure simultaneously the plasma emissivity in multiple energy ranges and deduce a variety of plasma parameters (e.g., Te, nZ, and ΔZeff). The energy dependence of each pixel is calibrated here over the range 3-22keV. The detector is exposed to a variety of monochromatic sources-fluorescence emission from metallic targets-and for each pixel, the lower energy threshold is scanned to calibrate the energy dependence. The data are fit to a responsivity curve ("S-curve") that determines the mapping between the possible detector settings and the energy response for each pixel. Here, the calibration is performed for three energy ranges: low (2.3-6 keV), medium (4.5-13.5 keV), and high (5.4-21 keV). We determine the achievable energy resolutions for the low, medium, and high energy ranges as 330eV, 640eV, and 950 eV, respectively. The main limitation for the energy resolution is found to be the finite width of the S-curve.

Multi-energy calibration of a PILATUS3 CdTe detector for hard x-ray measurements of magnetically confined fusion plasmas.

A multi-energy hard x-ray pin-hole camera based on the PILATUS3 X 100K-M CdTe detector has been developed at the Princeton Plasma Physics Laboratory for installation on the Tungsten Environment in Steady State Tokamak. This camera will be employed to study thermal plasma features such as electron temperature as well as non-thermal effects such as fast electron tails produced by a lower hybrid radiofrequency current drive and the birth of runaway electrons. The innovative aspect of the system lies in the possibility of setting the threshold energy independently for each of the ∼100k pixels of the detector. This feature allows for the measurement of the x-ray emission in multiple energy ranges with adequate space and time resolution (∼1cm, 2ms) and coarse energy resolution. In this work, the energy dependence of each pixel was calibrated within the range 15 keV-100 keV using a tungsten x-ray tube and emission from a variety of fluorescence targets (from yttrium to uranium). The data corresponding to pairs of Kα emission lines are fit to the characteristic responsivity ("S-curve"), which describes the detector sensitivity across the 64 possible energy threshold values for each pixel; this novel capability is explored by fine-tuning the voltage of a six-bit digital-analog converter after the charge-sensitive amplifier for each of the ∼100k pixels. This work presents the results of the calibration including a statistical analysis. It was found that the achievable energy resolution is mainly limited by the width of the S-curve to 3 keV-10 keV for threshold energies up to 50 keV, and to ≥20 keV for energies above 60 keV.

Long discharges in a steady state with D2 and N2 on the actively cooled tungsten upper divertor in WEST

Nitrogen (N2) will be used in ITER to enhance the radiative fraction to ∼90%, thereby cooling the edge plasma and preventing damage to the plasma-facing components. However, the reactivity of N2 with hydrogen isotopes can lead to the formation of tritiated ammonia (NT3). This should be considered in terms of the in-vessel tritium inventory, the regeneration of the cryo pumps, and the processes in the ITER de-tritiation plant. In the ‘W’ Environment in Steady-state Tokamak (WEST), a series of long L-mode discharges (∼50 s), with a constant N2 seeding from the outer strike point region has been performed on the upper actively cooled divertor. In the absence of active pumping, the N2 balance shows steady-state retention during plasma discharge, and is partially (∼35%) released in between discharges. Although a significant amount of N2( 18.65 Pa m3) has been injected, the wall still exhibited N2 pumping capabilities. Under these conditions, as long as this N2 reservoir is not saturated, there is not enough N available for the detectable threshold of ND3 formation to be reached. In these WEST experiments, no ammonia is detected during the pulse or after the pulse in the outgassing phase. These results are consistent with and complementary to the N2 seeded experiments performed in the Joint European Torus (JET) with its ITER-like wall and in the Axially Symmetric Divertor Experiment (ASDEX) upgrade.

From tore Supra to WEST: Evolution of the CODAC infrastructure

In 2013 Tore Supra limiter tokamak at CEA Cadarache went through major changes to become the WEST tokamak (Tungsten [W] Environment in Steady state Tokamak). The tokamak was upgraded into an X point divertor device. Taking advantage of its long discharge capability, WEST's goal is to minimize technological and operational risks of the actively cooled tungsten divertor for ITER.At the same time, Control Data Access and Communication (CODAC) system has been redesigned in order to satisfy the performance and evolutions needed by researchers, and to facilitate maintenance and evolution for the CODAC team.This paper describes the upgrade of the CODAC system from Tore Supra to WEST.

The Commissioning of the WEST Tokamak: Experience and Lessons Learned

The integrated commissioning of the tungsten (W) environment in steady-state tokamak (WEST) was started with the superconducting magnet cool-down and completed one year after by the injection of 2.5 MW of additional power. It consisted of several steps: local system commissioning, integrated commissioning without plasma, and first plasma operation. This article reviews these WEST commissioning phases and highlights the experiences and the lessons learned. The magnet cooling down operation was initiated in late September 2016. The vacuum vessel was pumped out three weeks before the first plasma. The impregnation and curing of the two in-vessel divertor coils was performed at 180 °C during two days contributing to the baking of the vacuum vessel. In the meanwhile, rehearsal sessions for testing the CODAC systems were organized every month. The toroidal, poloidal, and divertor field coils were energized and their performance was tested successfully allowing moving to the first plasma operation phase. After a few attempts, plasma breakdown was achieved confirming that all systems were properly running and synchronized. Due to several micro air leaks on fiber optic feedthroughs and issues with the design of the in-vessel stabilizing plates, plasma current ramp-up could not be achieved in a first step. Once these issues have been fixed, confined plasmas could be obtained. Diverted plasmas were then achieved and the plasma current was increased up to 800 kA for more than 10 s. Up to 2.5 MW of additional power was eventually successfully coupled to the plasma.

Thermomechanical simulation of flat tiles mock-ups under high heat flux

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

Incorporating magnetic equilibrium information in Gaussian process tomography for soft X-ray spectroscopy at WEST.

Gaussian process tomography (GPT) [J. Svensson, JET Internal Report EFDA-JET-PR(11)24, 2011 and D. Li, J. Svensson, H. Thomsen, F. Medina, A. Werner, and R. Wolf, Rev. Sci. Instrum. 84, 083506 (2013)] is a recently developed tomography method applied earlier to soft X-ray (SXR) spectroscopy on WEST-Tungsten (W) Environment in Steady-state Tokamak. The short execution time of the algorithm makes GPT an important candidate for providing real-time information on impurity transport and for fast MHD control. In earlier work, GPT has shown its flexibility by providing good reconstruction results without background information about the magnetic equilibrium. On the other hand, information about the magnetic flux surface geometry can in general be useful for additional regularization of the solution. In this paper, we develop a way to take into account the equilibrium information, by constructing a covariance matrix of the prior Gaussian process depending on the flux surface geometry. The GPT method is validated using synthetic SXR emissivity profiles relevant to WEST plasmas and compares favorably with the classical algorithm based on minimization of the Fisher information.