Tore Supra Research Articles

The effect of the electron κ-distribution on the dust particle charging in the radio-frequency thermal-sheaths

In order to investigate collisionless radiofrequency plasma sheaths containing dust particles, three models are utilized: the novel kinetic scheme Ensemble-in-Spacetime (EST) model for calculating sheath parameters, the Dust Particle Charging model, and the Single Dust Particle model. The EST model has been modified to account for κ-electron distributions. This model is applicable to radiofrequency plasma sheaths found in tokamaks equipped with an ion cyclotron radiofrequency (ICRF) wave heating system, such as JET, West(Tore Supra), EAST, ASDEX-U, and KSTAR. The calculated sheath parameters are utilized to determine the electron and ion currents of the dust particles. In the intermediate radio-frequency regime, when the ion plasma frequency is comparable to the ICRF, the flux and energy of the ions are modulated in time within the sheath. The ions are not inertialess, and the value of κ affects the electron and ion densities. As the value of κ increases, the time-averaged electron and ion densities, sheath edge position, and area of the sheath voltage-sheath charge hysteresis loop also increase. However, the ion energy distribution remains insensitive to the κ distribution. The dust particles are charged with different negative charges based on their radius, position within the sheath, κ-electron distribution, and sheath potential. These particles exhibit oscillatory motion due to their interactions with the plasma and gravitational fields and are accelerated toward the plasma core.

Development of Boron Coatings for antenna protection limiters in the WEST Tokamak

During the upgrade of Tore Supra to transform the machine into a fully metallic plasma facing environment (WEST), the Antennae Protection Limiters (APL), originally made in CFC, have been covered with a Tungsten coating with a Molybdenum interlayer (W/Mo/CFC).However, due to the aging of the W/Mo/CFC coating, the APL have been redesigned and future antenna limiters will be made of copper alloy (CuCrZr) heat sink structure with a thick boron (B) coating to reduce remaining W impurity levels during plasma operations and address the blistering issues seen in previous coatings.Among all the techniques able to realise B coating, Plasma Spray is the most appropriate technique owing to its simplicity, low cost and ability to deposit thick boron coatings on complex geometries.Various boron coatings with thicknesses up to 900 µm were produced, characterized and their thermal behavior have been assessed using a specific laser heating program and a conventional electron/ion beam facility. This paper provides an overview of the R&D activities on boron coating developments and the related validation programs showing that the thermal conductivity of deposited films is drastically reduced compared to bulk boron, preventing use without additional coating optimization.

Safety factor influence on the edge E × B velocity establishment in tokamak plasmas

This study is motivated by experiments on Tore Supra and WEST tokamaks where a deepening of the E × B velocity—governed by the radial electric field Er —near the edge is observed when the safety factor decreases in L-mode plasmas. Flux-driven global simulations of ion temperature gradient turbulence recover qualitatively the trend observed in the experiments, i.e. the E × B velocity increases when decreasing the safety factor. From these simulations, multiple clues point out the role of turbulence in the establishment of the radial electric field even though the turbulent intensity increases with the safety factor. The proposed mechanism to elucidate this phenomenon, backed up by a reduced model, is that the damping of the poloidal flow, governed by the neoclassical friction, increases more strongly with the safety factor than the turbulent drive for Er , due to the (r,θ) component of the Reynolds stress.

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.

Full-f gyrokinetic simulations of Ohmic L-mode plasmas in linear and saturated Ohmic confinement regimes

Two time slices in linear and saturated Ohmic confinement (LOC and SOC) regimes in a Tore Supra Ohmic L-mode discharge are analyzed using nonlinear global full-f gyrokinetic simulations, and qualitative features of the LOC–SOC transition are reproduced. The exhaust of carbon impurity ions is caused by ion mixing, which is driven by the toroidal field stress. The intrinsic rotation develops in the opposite direction between the LOC and SOC phases, which is characterized by different features of the mode asymmetry between trapped electron modes (TEMs) in the LOC phase and ion temperature gradient-driven modes in the SOC phase, leading to the change of the profile shear stress. Here, the mode asymmetry or the ballooning angle depends not only on the profile shear and the Er shear but also on the radial electric field Er itself. The energy fluxes of electrons and deuterium ions are dominant in the LOC and SOC phases, respectively, and the ratio of the energy confinement time between the two phases agree with the experimental value. Turbulent frequency spectra are characterized by quasi-coherent modes (QCMs) and broad-band spectra in the LOC and SOC phases, respectively. The QCMs are produced by a split of the toroidal mode number spectra of TEMs induced by the Doppler shift of poloidal E × B rotation due to Er, which is sustained in the electron diamagnetic direction by the ripple induced counter-current rotation, and enhanced by higher temperature in the LOC phase.

Electron cyclotron heating assisted start-up experiments in J-TEXT

Electron cyclotron heating (ECH) breakdown and burn-through assist has been adopted to make the ITER start-up, which uses a low toroidal electric field of about 0.3 V/m, more robust. Related studies have been carried out on many fusion machines such as DIII-D, JT-60U, Tore Supra and FTU. However, the required ECH power to ensure an effective breakdown assist at ITER is not yet clear. Experiment was also performed in J-TEXT to determine the minimum ECH power requirements to assist breakdown and develop a better physics description of the process. The breakdown loop voltage for a successful shot was reduced from 34 to 3.7 V (corresponding to 0.56 V/m) by 300 kW X2-mode ECH power. The critical ECH power to assist the successful formation of a tokamak discharge is about 200kW, and below this value, ionization caused by ECH power is observed, however, the tokamak discharge start-up does not succeed. The effect of different loop voltages and ECH pulse-width on start-up was also studied. As injecting ECH power at a proper time, low ECH power can achieve a similar pre-ionization result as the case with high power. Extremely low breakdown voltage leads to a higher toroidal field later when putting a capacity to continue discharging in J-TEXT. The earlier shutdown of ECH power caused a failed discharge.

Operational domain for the new 3MW/1000s ECRH System on WEST

The ECRH system formerly used in Tore Supra is being upgraded to start on WEST in 2023, at a power level of 1MW and frequency of 105 GHz. Its ultimate 3MW/1000s capability is expected to enlarge the WEST operational domain by increasing margins with respect to H-mode access, and by providing additional flexibility in terms of achievable scenarios using impurity and/or MHD control. This flexibility is made possible using an antenna based on three steerable mirrors for controlled power injection. In order to determine an appropriate range of EC wave injection angles for WEST scenarios, the fast and reliable ray-tracing code REMA has been interfaced with the WEST IMAS database. This allows the EC power damping rate to be quickly assessed, as well as deposition profiles to be predicted in realistic plasma conditions. Based on a typical WEST discharge at central magnetic field B0~3.6 T, central line-averaged electron density nl~4 × 1019 m−3 and central electron temperature Te0~3keV, ray-tracing calculations have been performed. Comprehensive poloidal and toroidal angle scans, as well as variations of Bt, nl and Te0 with respect to the reference parameters have allowed an adequate range of injection angles to be determined for efficient use of ECRH and/or ECCD in typical WEST scenarios, and compared with the mechanical limits set by the antenna mechanical characteristics. In order to further characterize the effect of this new power source in WEST scenarios, EC wave deposition and current profiles from ray-tracing calculations have been included in integrated simulation codes. It has been shown that this additional power source could allow central electron heating to be achieved, potentially alleviating the issue of radiative collapse caused by impurities observed in some situations.

A new 3MW ECRH system at 105 GHz for WEST

The aim of the WEST experiments is to master long plasma pulses (1000s) and expose ITER-like tungsten wall to deposited heat fluxes up to 10 MW/m2. To increase the margin to reach the H-Mode and to control W-impurities in the plasma, the installation of an upgraded ECRH heating system, with a gyrotron performance of 1MW/1000s per unit, is planned in 2023. With the modifications of Tore Supra to WEST, simulations at a magnetic field B0∼3.7T and a central density ne0∼6 × 1019 m−3 show that the optimal frequency for central absorption is 105 GHz. For this purpose, a 105 GHz/1MW gyrotron (TH1511) has been designed at KIT in 2021, based on the technological design of the 140 GHz/1.5 MW (TH1507U) gyrotron for W7-X. Currently, three units are under fabrication at THALES. In the first phase of the project, some of the previous Tore Supra Electron Cyclotron (EC) system components will be re-installed and re-used whenever possible. This paper describes the studies performed to adapt the new ECRH system to 105 GHz and the status of the modifications necessary to re-start the system with a challenging schedule.

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.

Extraction of quasi-coherent modes based on reflectometry data

The identification of turbulence sources would drive to a deeper understanding of confinement dynamics in tokamak plasmas. Turbulence results from a mixture of instabilities corresponding to sources at different timescales and spatial scales. Using poloidal correlation reflectometry and multi-pin Langmuir probe, it was shown in the T-10 and the Tokamak Experiment for Technology Oriented Research (TEXTOR) tokamaks that the reflectometry frequency spectrum is the superposition of several components: broadband component, quasi-coherent (QC) modes and low-frequency components. The relevance of QC modes is associated with their link with the trapped electron mode instability. This link was exhibited in the transition from the linear ohmic confinement (LOC) to the saturated ohmic confinement (SOC) regime. A method is presented in this paper to extract the QC mode component from the reflectometry data, enabling its separation from the broadband component and the study of its time evolution. It is a first step toward the discrimination of turbulence sources. The central idea explores a way to combine the approach of signal processing and machine learning. The continuous wavelet transform on the basis of complex Morlet wavelet has proved to be efficient in providing a decomposition of a signal at different scales over time for fluctuation tackling; clustering techniques, such as the mini-batch K-means, are able to tackle clusters at different scales. The method was applied to Tore Supra and TEXTOR reflectometry data. In Tore Supra, the amplitude of the extracted QC mode component decreases during the LOC–SOC transition. In TEXTOR, the amplitude of the coherent spectra of the extracted QC mode component is similar to the experimental coherent spectra obtained through correlation reflectometry. The developed method permits the extraction of components, preserving their physical and statistical properties.

Energy transfer of trapped electron turbulence in tokamak fusion plasmas

The first principle gyrokinetic simulations of trapped electron turbulence in tokamak fusion plasmas demonstrate the energy transfers from the most linearly unstable modes at high k_theta rho _isim 1 to intermediate k_theta via parametric decay process in a short period of linear-nonlinear transition phase. Dominant nonlinear wave-wave interactions occur near the mode rational surface msimeq nq. In fully nonlinear turbulence, inverse energy cascade occurs between a cutoff wave number k_c and k_theta rho _isim 1 with a power law scaling |phi (k_theta )|^2propto k^{-3}, while modes with k<k_c are suppressed. The numerical findings show fair agreement with both the weak turbulence theory and realistic experiments on Tore Supra tokamak.

Collisionless losses of fast ions in the divertor tokamak test due to toroidal field ripple

In this paper we analyze fast ion motion in the divertor tokamak test (DTT) device (Albanese et al 2017 Nucl. Fusion 57 016010). It is planned that DTT will be heated through a mix of 45 MW heating power, including 15 MW negative-ion-based neutral beam heating (NNBI) which is currently being developed by Consorzio RFX in Padova, Italy (Agostinetti et al 2019 Fusion Eng. Design 146 441–446). An issue for DTT is that a toroidal field (TF) ripple with a maximum value of about ∼0.42% (with respect to the on-axis magnetic field B 0) is expected on the low-field side, and this ripple interacts with fast ions through the rather well-known phenomena of ripple-precession resonances, in addition to prompt losses of ions which do not complete a full orbit in the poloidal plane. We will show that, with the planned geometry of NNBI, prompt losses are negligible, and ripple-precession losses amount to a maximum of 0.15%. The calculations are performed with the guiding center code Orbit using two different equilibria, and a beam with an energy of 400 keV and the injection angle α inj = 40° (measured w.r.t. the first wall), which corresponds to a pitch of injected particles λ = v ∥/v ≈ sin α inj = 0.65. The main resonances are of the form ω b − nNω d = 0, ω b and ω d being the bounce and precession frequency, respectively, N = 18 the ripple periodicity and 3 ⩽ n ⩽ 6 are the toroidal wavenumbers of the resonances. Although collisionless interaction with the TF ripple does not pose a serious threat to the NNBI project, an open question remains as to whether the presence of these resonances will interact with fast particles accelerated by Alfvén eigenmodes, and if stochastization of the resonances is possible in DTT, as was observed in the past in TORE SUPRA.

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.

New, improved analysis of correlation ECE data to accurately determine turbulent electron temperature spectra and magnitudes (invited).

Turbulent electron temperature fluctuation measurement using a correlation electron cyclotron emission (CECE) radiometer has become an important diagnostic for studying energy transport in fusion plasmas, and its use is widespread in tokamaks (DIII-D, ASDEX Upgrade, Alcator C-Mod, Tore Supra, EAST, TCV, HL-2A, etc.). The CECE diagnostic typically performs correlation analysis between two closely spaced (within the turbulent correlation length) ECE channels that are dominated by uncorrelated thermal noise emission. This allows electron temperature fluctuations embedded in the thermal noise to be revealed and fluctuation level and spectra determined. We have demonstrated a new, improved CECE coherency-based analysis for calculating the temperature fluctuation frequency spectrum and level, which has been verified both numerically through the simulation of synthetic ECE radiometer data and through analysis of experimental data from the CECE system on DIII-D. The new formulation places coherency-based analysis on a firm foundational footing and corrects some currently published methodologies. This new method accurately accounts for bias error in the coherence function and correctly calculates noise levels for a fixed data record length. It provides excellent accuracy in determining temperature fluctuation level (e.g., <10% error) even for a small realization number in the ensemble average. The method also has a smaller uncertainty (i.e., error bar) in the power spectrum when compared to the more standard cross-power method when evaluated at low coherency. Direct calculation of system noise level using correlation between randomized intermediate frequency signals is recommended.

WestBox: An object-oriented software component for WEST CODAC

Abstract COntrol, Data Acquisition and Communication (CODAC) real-time software codes are key elements for the operation of a fusion device both for the machine protection and for the optimization of the experiments. In 2013, following the WEST (W -for tungsten- Environment Steady-state Tokamak) upgrade of Tore-Supra, the whole legacy acquisition system has been re-factored. The WEST CODAC framework which inherited the older implementations used over 20 years on TORE Supra has been reorganized into primary software components corresponding to basic technical features, such as Database Access, Inter-process Communications, Real-time Device Management, Timing System Network, Finite-State Machine, and Shared Memory Network. In addition, a new modern middleware component named Westbox has been developed, to simplify developments for native diagnostic upgrades and cross-platform data acquisition hardware and software from guest collaborators.

Recent progress on turbulence and multi-scale interactions in tokamak plasmas

Studies on nonlinear interactions and multi-scale physics are crucial in the physical understanding of the turbulence and transport in the complex nonlinear system of fusion plasmas. This paper reviews recent studies on the turbulence and the related transport. The prominent results are mainly from tokamaks, such as HL-2A, Tore Supra and J-TEXT. First, turbulence behaviors and its role in transport will be introduced. Critical gradients in heat and particle transports have been observed and reproduced by simulations. Turbulence modes and its role in the transport are identified using advanced diagnostics. The measured turbulence transition is consistent with the observation of particle convection reversal and the simulation prediction. In addition, the excitation of an electromagnetic turbulence by either edge self-accumulated or externally seeded impurities has been observed. The impurity-driven turbulence plays an essential role in regulating the impurity transport. Then, the effect of plasma flows on turbulence and multi-scale interactions (micro-turbulence, meso-scale fluctuations, and large scale MHD activities) will be presented. Moreover, a new meso-scale electrostatic fluctuation has been observed. The results indicate that geodesic acoustic modes and magnetic fluctuations can transfer energy through nonlinear synchronization. At last, the enhancement of nonlinear interactions by external actuators will be reported. Theoretical studies in comparisons with experimental findings will be discussed.

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.

Reconstruction of hollow areas in density profiles from frequency swept reflectometry

The standard density profile reconstruction techniques are based on the WKB approximation of the probing wave’s phase, making them unable to properly reconstruct blind areas in the cut-off frequency profile. The reconstruction suffers a significant immediate error that is not rapidly damped. It is demonstrated that even though no reflections occur inside the hollow region causing the blind area, the higher probing frequencies that propagate through it carry information that can be used to estimate its properties. The usually ignored full-wave effects were investigated with the use of full-wave simulations in 1D, with special attention paid to the frequency band where they are dominant. A database of perturbation signals was simulated on five-dimensions of parameters and an application of the database inversion was demonstrated for a magnetic island in a Tore Supra discharge. The new adapted reconstruction scheme improved the description of the density profile inside the hollow region and also along 10 cm after it.

A limit-cycle model for internal transport barrier oscillations

A 0-D model for the time evolution of the electron temperature T and current density j is derived (following a systematic procedure) from the cylindrical 1-D electron heat transport and current density diffusion equations. The stationary regimes (stable fixed points) of the deduced dynamical system are analysed and it is shown that the model reproduces well the cases of total diffusion (no sources), of a pure Ohmic (OH) discharge and of a constant external heating scenario. Moreover, it is seen that, as the fraction of externally driven non-inductive current applied off-axis is increased, the system moves from an OH regime into an internal transport barrier (ITB) regime, where j is reversed and the negative magnetic shear reduces the heat diffusivity, thus increasing T at the core. When the external power deposition is made proportional to both T and j, limit-cycle oscillations, which resemble those of the O-regime at Tore Supra, are found. The 1-D transport equations are also solved numerically and an ITB oscillatory regime with features similar to the experiments is found, namely, a period of oscillation that is of the order of the resistive time scale, along with a decrease of the oscillation's amplitude with increasing frequency.

Software quality management approach for WEST CODAC

The WEST project has consisted in modifying the Tore Supra tokamak magnetic configuration to achieve X-point by adding divertor coils in the vacuum vessel and to test ITER-like plasma facing components. Major changes and significant developments have been performed on the measurement systems (diagnostics); the control, data access and communication; the plasma control system, and the actuators. To keep track of the software developments and step by step upgrade performed on any subsystem of the tokamak, a software quality management procedure has been implemented. This procedure consists also of a validation workflow ensuring that only validated version of the software are used during a plasma discharge. Such procedure is successfully used since 2 years.