Tore Supra Tokamak Research Articles

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.

Non-uniformity of the radial electric field for tokamak plasma on the low and the high toroidal magnetic field side

This paper presents a relativistic mechanism that can create a non-uniformity of a radial electric field on the low side and the high toroidal magnetic field side of tokamak plasmas. This mechanism is associated with the significant current velocity of electrons in the plasma for many experimental modes of tokamak operation. It is important that electric fields arise without the separation of charges in the plasma. The traditional approach to determining radial electric fields in tokamak plasmas gives a uniform radial electric field on magnetic surfaces inside the plasma, while there are no radial electric fields outside the plasma. The presented mechanism leads to the appearance of an electric field outside the plasma with a current, and a non-uniformity of the radial electric field on magnetic surfaces appears inside the plasma. The proposed mechanism can explain the experimentally measured non-uniformity of C+6 carbon ion toroidal rotation velocities in the Tore-Supra tokamak plasma.

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.

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.

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.

Development of Microwave Imaging Diagnostics for WEST Tokamak

WEST is an upgrade of the Tore Supra tokamak to test ITER like divertor elements over very long pulses. For the studies of MHD instabilities affected by tungsten impurities, two microwave imaging diagnostics have been developed to obtain density and temperature fluctuation maps. The core reflectometer was refurbished and now measures the core density profiles in few microseconds. Measurements of several thousand profiles in sequence with microsecond dead time (burst mode) will be available in 2019 enabling the reconstruction of density maps with unprecedented time resolution. The major issue for the implementation of an electron cyclotron emission imaging diagnostic was the heat radiation during long pulses on the large vacuum window and the first optics. Two metallic in-vessel mirrors inside the personnel access tube solved this issue. The first mirror can withstand the heat radiation without water cooling, the second mirror redirects the beam toward the personnel access port. The oblique and small view angle of this access eases also the thermomechanical constraints on the vacuum window. As the space between the port flange and the tokamak access lobby was tight, a compact optical enclosure with vertical aligned optics was designed. To ensure a good reproducibility of the mirrors and the optical enclosure positions after a maintenance period and thus avoid re-alignment, different mechanical systems are used to manipulate the mirrors and the optical enclosure and to fine tune their position and their orientation. The first channel covers the low field side and will provide 192 pixel images of temperature fluctuations with 1–2 microsecond time resolution. The installation of the in-vessel mirrors is scheduled during March 2019. Diagnostic qualification tests will be done before the restart of the experiments late May 2019.

Lévy flights on a comb and the plasma staircase.

We formulate the problem of confined Lévy flight on a comb. The comb represents a sawtoothlike potential field V(x), with the asymmetric teeth favoring net transport in a preferred direction. The shape effect is modeled as a power-law dependence V(x)∝|Δx|^{n} within the sawtooth period, followed by an abrupt drop-off to zero, after which the initial power-law dependence is reset. It is found that the Lévy flights will be confined in the sense of generalized central limit theorem if (i) the spacing between the teeth is sufficiently broad, and (ii) n>4-μ, where μ is the fractal dimension of the flights. In particular, for the Cauchy flights (μ=1), n>3. The study is motivated by recent observations of localization-delocalization of transport avalanches in banded flows in the Tore Supra tokamak and is intended to devise a theory basis to explain the observed phenomenology.

Hot spots induced by LHCD in the shadow of antenna limiters in the EAST tokamak

Hot spots induced by lower hybrid wave in experimental advanced superconducting tokamak tokamak have caused high performance experiment disruption and serious damages to the guard limiters. Experimental and theoretical analyses have been carried out to study its physical mechanism. Plasma density scan experiments indicate that the wall temperature within the hot spots enhanced by a factor of 5 and increases with the plasma density near the antenna. A lower hybrid current drive (LHCD)-only density climb experiment shows that the carbon impurity decreases to a minimum value at certain plasma density and then increases with the line averaged plasma density. A model has been developed to explain the mechanism of sputtering of graphite tiles due to hot spots as the plasma density near the LHCD antenna and the time increases. A theoretical scaling of the heat flux driven by LHCD is also presented and is consistent with the experimental scaling in the Tore Supra tokamak. The simulation results show that the total sputtering flux density has a minimum at a certain plasma density and gradually increases as the plasma density increases or decreases away from the minimum value, and the increase in parallel heat flux near the antenna would enhance the sputtering flux density. The sputtering flux density trend is qualitatively consistent with the density scan experiments. The simulated temporal evolution of sputtered flux implies that the chemical sputtering could be a candidate for the carbon impurity explosion.

The new magnetic diagnostics in the WEST tokamak.

The WEST tokamak consists of a major upgrade of the superconducting medium size tokamak Tore Supra aiming at testing ITER divertor components. Such modification has required rebuilding a full set of magnetic diagnostics. The project was started in 2013 and completed in 2016. The diagnostic consists of a set of 469 sensors (421 pick-up coils, 36 flux loops, and 12 Rogowski coils) installed in the WEST vacuum vessel. New analog integrators have been developed in order to obtain the magnetic field and flux from the raw signal of the sensors. During the startup phase of WEST, plasma currents of the order of a few kilo amperes were measured despite much larger current of the order of hundreds of kilo amperes flowing in nearby conducting structures. The diagnostic is now fully operational and exhibits a noise level of about 0.5 mT on the magnetic field, and 2.0 mWb on flux loops allowing identifying the plasma boundary with an accuracy of a few millimeters on a 2 ms time cycle.

Virtual Reality: Lessons learned from WEST design and perspectives for nuclear environment

From 2012–2016, the Tore Supra tokamak was upgraded in an x-point divertor device within the frame of the WEST project. A huge design activity was made to modify the whole configuration of Tore Supra including the in-vessel components. Some of the most noticeable changes are the addition of inner divertor coils inside the vacuum vessel and the replacement of all plasma facing components.Fitting all the new components in a previously existing environment was probably one of the major challenges of the project. To improve the assembly plan and the design of toolings, Virtual Reality (VR) was introduced in the design process.For complex assembly operations, simulations were developed for VR headsets from CAD models. From the requirements expressed by the design team, the VR engineers developed dedicated tools to address the specific issues. Engineers, designers and operators were then able to experience the assembly conditions in a 3D 360° real size immersive environment and interact with the models thanks to a large set of dedicated tools which are developed in the VR lab. This approach showed impressive results for issues such as accessibility, complex kinematics of large components and design of handling tools. It allowed reducing the number of mock-ups and helped the WEST assembly team to reduce the number of unforeseen assembly issues. It was for example used to simulate the assembly of the divertor: insertion through the ports, toroidal displacement on rails, and assembly with gaps lower than 1 mm.Virtual Reality will also provide innovative solutions to help designers in the anticipation of challenges linked to operation of a nuclear facility: operator’s training for nuclear maintenance, virtual mock-ups for qualification of maintenance, maintenance scenario optimization with real time simulations… A first module plotting the dose rate is currently being tested. On the one hand, it would for example allow designers to improve the design by adapting maintenance operations to any complex 3D map of dose rate, and on the other hand, it would permit workers to train for operations in radioactive environments (showing no go areas, hot spots…).

Poloidal asymmetries of flows in the Tore Supra tokamak

Simultaneous measurements of binormal velocity of density fluctuations using two separate Doppler backscattering systems at the low field side and at the top of the plasma show significant poloidal asymmetry. The measurements are performed in the core region between the radii 0.7 &amp;lt; ρ &amp;lt; 0.95, over a limited number of L-mode discharges covering a wide range of plasma conditions in the Tore Supra tokamak. A possible generation mechanism by the ballooned structure of the underlying turbulence, in the form of convective cells, is proposed for explaining the observation of these poloidally asymmetric mean flows.

Some experiments with the tunnel probe in a low temperature magnetized plasma

Experiments were performed using a Tunnel Probe (TP) inside the weakly-ionised plasma of the Linear Magnetized Plasma Device (LMPD). The TP is designed as a concave probe, which should annihilate the problem of sheath expansion in the ion branch of the I-V characteristic. As the ion saturation current is consequently well defined, the ion parallel current and plasma density can be more accurately calculated. Furthermore the ratio between the ion saturation currents on the two collectors (tunnel ring and the back-plate) can be used to derive the electron temperature. The TP has been repeatedly used with success on the former Castor and Tore-Supra tokamaks and will be used on the upgraded version of Tore-supra, namely the WEST tokamak, as well [1, 2]. It was however never used successfully in a low-temperature plasma.We studied the feasibility of the TP use in a low-temperature plasma for direct measurements of plasma temperature and density. The various probe characteristic dimensions, such as the distance between the two collectors, the aperture size and the probe radius were varied to see influence of the individual probe feature. We also varied the level of magnetization of the charged particle species, the background gas pressure (which influences the electron energy distribution function), the plasma density (important for the ratio between the λD and the ion Larmor radius). The sensitivity of the probe alignment to the magnetic field lines was also studied. We found, that the ion saturation current does not necessarily saturate and that the probe works according to expectations only in a limited amount of regimes.

Measurement of plasma current in Tokamaks using an optical fibre reflectometry technique

An optical time-domain reflectometer sensitive to the polarization of light is proposed for the measurement of plasma current in the Tore Supra fusion reactor. The measurement principle relies on the Faraday effect i.e. on the generation of a circular birefringence along an optical fiber subject to an axial magnetic field. The circular birefringence induces a polarization rotation that can be mapped along the fiber thanks to an opticaltime domain reflectometer followed by an linear polarizer. A proper fitting of the measurement trace then allows determining the applied plasma current. The sensor has been experimentally validated on the Tore Supra tokamak fusion reactor for a plasma current range going from 0.6 to 1.5 MA. A maximum error of 13.50% has been observed for the lowest current.

Tungsten coating by ATC plasma spraying on CFC for WEST tokamak

In the field of fusion experiments using a tokamak, the plasma facing components (PFC) are the closest object to the hot plasma. Due to the plasma-wall interaction, the material composing the PFC may enter the plasma and disturb the experiments. In the past, the main material for PFC was carbon (CFC, graphite), while the future reactors like ITER will be fully metallic, in particular tungsten. The Tore Supra tokamak has been transformed in an x-point divertor fusion device within the frame of the WEST (W (tungsten) Environment in Steady-state Tokamak) project in order to have plasma conditions close to those expected in ITER. The PFC other than the divertor has been coated with W to transform Tore Supra into a fully metallic environment. Different coating techniques have been selected for different kind of PFC. This paper gives an overview on the coating process used for the antennae protection limiter, the associated validation programme and concludes on the adequacy of the W coating with the WEST experimental programme requirements and gives perspectives on the development to be pursued.

Software demonstration of quasi on line parametric identification related to thermal state in tokamak plasmas

In this paper, estimation of space and time varying thermal diffusivity and heating source in a nonlinear parabolic partial differential equation (PDE) describing heat transport in tokamak plasma is considered. An iterative regularization method based on a Conjugate Gradient Method (CGM) is implemented to deal with model and measurement errors. The estimation strategy is investigated in the infinite dimensional framework. Real data from the Tore Supra (TS) tokamak are provided to evaluate the performance of the proposed strategy.

Width of turbulent SOL in circular plasmas: A theoretical model validated on experiments in Tore Supra tokamak

The relation between turbulent transport and scrape off layer width is investigated in circular plasmas toroidally limited on the inner wall. A broad set of experimental observations collected in the Tore Supra scrape off layer is detailed and compared to turbulent interchange models. Blob E × B drift velocities measured in experiments agree reasonably well with an analytical model derived for isolated blobs. Based on a time averaged particle flux balance, it is also shown that the SOL width depends on both the blob drift velocity and a blob intermittency parameter, which is so far not predicted by isolated blob models. A set of 2D isothermal turbulence simulations are used to derive a power law regression of the density width function of global control parameters. Quantitative agreement is found between this regression and experimental density widths measured in Tore Supra, over a large set of plasma conditions. The sensitivity to control parameters (major radius, safety factor and normalized Larmor radius) is roughly explained by the sensitivity of the blob velocity model. The predictions are also extended to power decay length in limited plasma configurations. For ITER start-up phases, the predicted power decay length fall in the range of extrapolations based on multi-machine regressions.

Mechanical analysis of WEST divertor support plate

The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the TER tungsten divertor strategy. The WEST project aims to test W monoblock Plasma Facing Units (PFU) under long plasma discharge (up to 1000s), with thermal loads of the same magnitude as those expected for ITER. Therefore the divertor is a key component of the WEST project, and so is its support structure, which has to handle strong mechanical loads. The WEST upper and lower divertor are made of 12 30° sectors, each one composed of 38PFU that can be made of tungsten, CuCrZr or graphite. A generic 316L stainless steel 30° conic support plate is used to hold the 38PFU together, regardless of their material. The PFUs are fixed on the support plate thanks to 152 Xm19 stainless steel fixing elements (4 per PFU), and in each of this fixing element an Aluminium-Nickel-Bronze alloy (Al-Ni-Br) pin is engaged in a slotted hole, in order to allow thermal expansion in the length direction of the PFU. The support plate is fixed on the divertor coil casing thanks to 10M10 screws. Mechanicals loads which act on the PFUs are transmitted to the support plate through the fixing elements. These loads are due to Vertical Displacement Event (VDE), disruptions and thermal expansion of the PFU. First the different load cases, PFU configurations and scenario are presented. Then an ANSYS plastic mechanical simulation is performed in order to validate the number of cycles of the support plate for each scenario: 30,000 cycles in steady-state and 3000 cycles in VDE. Finally reactions forces from the previous ANSYS simulation are used in order to calculate the stress in the M10 screws.

Plasma Current Measurement in Thermonuclear Fusion Reactors Using a Photon-Counting POTDR

In this letter, we present a fiber-optic current sensor for the measurement of plasma current in thermonuclear fusion reactors. The sensor includes a low-birefringence optical fiber placed around a section of the vacuum vessel that hosts the fusion reaction. The interrogating unit consists of a polarization-sensitive photon-counting optical time domain reflectometer that enables to determine the light polarization rotation induced by the plasma current. The measurement principle is demonstrated by means of the Jones formalism. The sensor has been experimentally validated on the Tore Supra tokamak fusion reactor for a plasma current range going from 0.6 to 1.5 MA. A maximum error of 13.50% has been observed for the lowest current.

Overview of the different processes of tungsten coating implemented into WEST tokamak

The Tore Supra tokamak is being transformed in an x-point divertor fusion device within the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the ITER tungsten divertor strategy. The WEST project aims at testing W monoblock Plasma Facing Units (PFU) under long plasma discharge, with thermal loads of the same magnitude as those expected for ITER. The others Plasma Facing Components (PFC) will also be modified and coated with W to transform Tore Supra into a fully metallic environment. Different coating techniques have been selected, taking into account the specifications of the various PFC: heat loads, complex geometries (length up to 1m) and different substrates (CuCrZr for actively cooled PFC, graphite and CFC for other components). This paper gives an overview on the different processes used and the associated validation program and concludes on the adequacy of the W coating with the WEST experimental program requirements.

The role of the plasma current in turbulence decrease during lower hybrid current drive

The interaction of radio frequency (RF) waves with edge turbulence has resurfaced after the results obtained on many tokamaks showing that edge turbulence decreases when the ion cyclotron frequency heating (ICRH) is switched on. Using the lower hybrid (LH) waves to drive current into tokamak plasmas, this issue presented contradicting results with some tokamaks (FTU &amp; HT-7) showing a net decrease, similar to the ICRH results, and others (Tore Supra) did not. In this article, these apparent discrepancies among tokamaks and RF wave frequencies are removed. It is found that turbulence large-scale structures in the scrape-off layer decrease at high enough plasma currents (Ip) on the Tore Supra tokamak. We distinguish three regimes: At low Ip's, no modification is detected with statistical properties of turbulence similar to ohmic plasmas even with PLH reaching 4.8 MW. At moderate plasma currents, turbulence properties are modified only at a high LH power. At high plasma currents, turbulent large scales are reduced to values smaller than 1 cm, and this is accompanied by a net decrease in the level of turbulence of about 30% even with a moderate LH power.