Tokamak Discharges Research Articles

Distinctive features of measuring <I>T<SUB>e</SUB></I> and <I>n<SUB>e</SUB></I> spatial distributions in the Globus-M2 spherical tokamak using method of Thomson scattering of laser radiation

The results of measuring the electron temperature and density spatial distributions in plasma of the Globus-M2 tokamak using the Thomson scattering diagnostics are presented. The diagnostics provides measurements throughout the entire tokamak discharge, starting from time of gas breakdown. The Thomson scattering data were analyzed in order to determine the positions of the last closed flux surface, the plasma magnetic axis, and the radius of inversion during the saw-tooth oscillations. The results of measurements performed during the internal reconnection of magnetic field lines are presents, as well as the dynamics of spatial distributions of electron temperature, density and pressure during the plasma transition to the H-mode. The results of measuring the electron temperature distribution in the scrape-off layer using the Thomson scattering diagnostics are also presented for distances up to 4 cm outside the last closed flux surface.

He II line intensity measurements in the JET tokamak

Abstract An understanding of the behaviour of the D or He fuel used in tokamak discharges is essential in analyses such as modelling edge and divertor transport and the erosion of the vessel walls. However, poor agreement is found between measurements made on the JET tokamak and collisional-radiative models used to predict the hydrogen-like D and He line intensities. The range of temperatures of the plasmas emitting the radiation is also limited, in contrast to that for many impurities for which a wide range is possible. This is particularly so for He II whose line intensities tend to have the same near-constant ratios in most pulses, suggesting that the emission originates in plasma regions with very similar electron temperatures. To gain understanding and to allow quantitative comparisons with theoretical models, extensive observations of the VUV Lyman series have been made, for all discharge scenarios run during three He campaigns. Those for He discharges in both JET ITER-like wall (JET-ILW) and JET C (JET-C) campaigns are presented here. He discharges have the advantage of fewer impurities resulting in less complex spectra than when D is used as the fuel. However, the characteristics of the observed discrepancies are similar in both species, allowing He to be used as a proxy for D in order to gain understanding of the discrepancy. In addition, the study of He avoids the complication of molecular species contributing to the level populations. Opacity effects are also expected to be less severe in He discharges. Nevertheless, so as to ensure that the measurements are not unduly affected by opacity, comparisons have also been made with emission from Balmer and Paschen series members. Measurements of both line intensities and their ratios are presented for all-pulse surveys and for individual pulses. In exceptional cases in which the He emission is intense a dependence on the He II line intensity is demonstrated. The discrepancy between these measurements and the theoretical models is illustrated.

Effects of vacuum magnetic field region on the compact torus trajectory in a tokamak plasma

The trajectory of the compact torus (CT) within a tokamak discharge is crucial to fueling. In this study, we developed a penetration model with a vacuum magnetic field region to accurately determine CT trajectories in tokamak discharges. This model was used to calculate the trajectory and penetration parameters of CT injections by applying both perpendicular and tangential injection schemes in both HL-2A and ITER tokamaks. For perpendicular injection along the tokamak’s major radius direction from the outboard, CTs with the same injection parameters exhibited a 0.08 reduction in relative penetration depth when injected into HL-2A and a 0.13 reduction when injected into ITER geometry when considering the vacuum magnetic field region compared with cases where this region was not considered. In addition, we proposed an optimization method for determining the CT’s initial injection velocity to accurately calculate the initial injection velocity of CTs for central fueling in tokamaks. Furthermore, this paper discusses schemes for the tangential injection of CT into tokamak discharges. The optimal injection angle and CT magnetic moment direction for injection into both HL-2A and ITER were determined through numerical simulations. Finally, the kinetic energy loss occurring when the CT penetrated the vacuum magnetic field region in ITER was reduced by by optimizing the injection angle for the CT injected into ITER. These results provide valuable insights for optimizing injection angles in fusion experiments. Our model closely represents actual experimental scenarios and can assist the design of CT parameters.

Runaway electron mitigation by 3D fields application in ASDEX Upgrade, COMPASS, and RFX-mod

Disruption-generated runaway electron (RE) beams represent a potentially severe threat for tokamak plasma-facing components. Application of properly designed 3D fields can act as a mitigation mechanism, as recently investigated in ASDEX Upgrade (AUG) and COMPASS experiments and in the tokamak discharges of RFX-mod. In all of these devices, the dynamics of the disruption are affected by the application of magnetic perturbations (MPs), and the resulting RE beam current and lifetime are significantly reduced. These experiments show, in particular, that the strength of the observed effects strongly depends on the poloidal spectrum of the applied MPs, which has been reconstructed including the plasma response. This paper reports the main findings on RE mitigation from the previously mentioned three devices, highlighting the common physics behind them and their interpretation by using the guiding center code ORBIT.

Strategy to systematically design and deploy the ITER plasma control system: A system engineering and model-based design approach

The paper details the process of developing the ITER Plasma Control System (PCS), that is, how to design and deploy it systematically, in the most efficient and effective manner. The integrated nature of the ITER PCS, with its multitude of coupled control functions, and its long-term development, calls for a different approach than the design and short-term deployment of individual controllers. It requires, in the first place, a flexible implementation strategy and system architecture that allows system re-configuration and optimization throughout its development. Secondly, a model-based system engineering approach is carried out, for the complete PCS development, i.e. both its design and deployment. It requires clear definitions for both the PCS role and its functionality, as well as definitions of the design and deployment process itself. The design and deployment process is shown to allow tracing the relationships of the many individual design and deployment aspects, such as system requirements, assumed operation use-cases and response models, and eventually verification and functional validation of the system design. The functional validation will make use of a dedicated PCS simulation platform that includes the description of the control function design as well as plant, actuator and sensor models that enable the simulation of these functions. By establishing a clear understanding of the interconnected steps involved in designing, implementing, commissioning, and operating the system, a more systematic approach is achieved. This ensures the completion of a comprehensive design that can be deployed efficiently, hence preventing the loss of precious operational time needed to debug and retune control functions and more importantly avoiding tokamak discharge disruptions.

Hydrogen removal by electron cyclotron wall conditioning with neon gas and its impact of tokamak plasma start-up on the QUEST spherical tokamak

Electron cyclotron wall conditioning with neon gas (Ne-ECWC) has been performed on the normal conducting spherial tokamak QUEST with metal walls under a trapped particle configuration with O-mode EC waves including X-mode polarization with a frequency of 8.2 GHz and an injection power of 16 kW. The Ne-ECWC removes hydrogen from the wall with small neon retention. The Ne-ECWC decreases hydrogen recycling at the following tokamak discharges, contributing to an improvement of the following tokamak plasma start-up: the plasma current increases and the start-up timing of the plasma current shifts forward. However, defects such as voids and bubbles are formed on tungsten surface exposed to the Ne-ECWC plasma.

Numerical analysis for the free-boundary current reversal equilibrium in the AC plasma current operation in a tokamak

In recent decades, tokamak discharges with zero total toroidal current have been reported in tokamak experiments, and this is one of the key problems in alternating current (AC) operations. An efficient free-boundary equilibrium code is developed to investigate such advanced tokamak discharges with current reversal equilibrium configuration. The calculation results show that the reversal current equilibrium can maintain finite pressure and also has considerable effects on the position of the X-point and the magnetic separatrix shape, and hence also on the position of the strike point on the divertor plates, which is extremely useful for magnetic design, MHD stability analysis, and experimental data analysis etc. for the AC plasma current operation on tokamaks.

Line identification of extreme ultraviolet spectra from aluminum ions in EAST Tokamak plasmas

Extreme ultraviolet (EUV) spectra emitted from aluminum in the 5-340 Å wavelength range were observed in Experimental Advanced Superconducting Tokamak (EAST) discharges. Several spectral lines from aluminum ions with different degrees of ionization were successfully observed with sufficient spectral intensities and resolutions using three fast-time-response EUV spectrometers. The line identification uses three independent state-of-art computational codes for the atomic structure calculations, which provide the wavelengths and the collisional and radiative transition rate coefficients. These programs are hullac (Hebrew University—Lawrence Livermore Atomic Code), autostructure, and fac (Flexible Atomic Code). Using three different codes allows us to resolve some ambiguities in identifying certain spectral lines and assess the validity of the theoretical predictions.

First results of EAST edge modeling with SOLPS‐ITER 3.2.0 on Extended Grid

AbstractThe processes in far scrape‐off layer (SOL) and the plasma interaction with the first‐wall (FW) elements may notably affect the tokamak discharge, since they define fuel and impurity recycling, material erosion and redeposition, wall surface heating, etc. For a long time, the far SOL description in most plasma edge transport codes was insufficient or absent at all, and so particle and heat fluxes onto the FW (except divertor plates) were out of consideration. Recently some codes, for example, SOLEDGE and SOLPS‐ITER are upgraded allowing for the extension of the computational grid up to real walls and for corresponding account of the vacuum vessel shape and all in‐vessel elements. The new release of SOLPS‐ITER (the version 3.2.0) required a development of a new code data structure and new approach to numerical approximation of fluid equations compatible with unstructured non‐orthogonal computational grid. Intensive testing of the new code in different conditions is still required. In the present contribution, such a testing is performed for the EAST disconnected double null (DDN) L‐mode discharge. For the first time, the SOLPS‐ITER 3.2.0 modeling results with drifts and currents turned on are presented, and a comparison to former SOLPS‐ITER version (3.0.8) is performed. The far SOL transport and its effects on the discharge performance are studied by comparing the computational results obtained on several meshes which differ by their width in equatorial midplane. A single null (SN) one (with mesh width limited by distance to the secondary separatrix) was examined versus two DDN meshes (one with actual and one with artificially extended targets to make mesh wider) and a true unstructured (the widest) mesh. The notable difference in results obtained on different meshes appears in those places where plasma density does not vanishes at the computational domain boundaries. For the cases on true unstructured (the widest) mesh, the particle and heat fluxes onto central column, limiters, far SOL part of targets, dome umbrella and other EAST far SOL in‐vessel structures are calculated for the first time by SOLPS‐ITER, allowing assessment of the plasma interaction with those surfaces.

Real-Time Plasma Magnetic Control System with Equilibrium Reconstruction Algorithm in the Feedback for the Globus-M2 Tokamak

To control the plasma shape during a tokamak discharge, it is necessary to calculate the plasmashape in real-time. The rate requirements for the shape calculations are especially high for tokamaks with asmall radius, such as Globus-M2 (St. Petersburg, Russia). A real-time magnetic plasma control system forthe Globus-M2 tokamak with flux and current distribution identification (FCDI) algorithm for the plasmaequilibrium reconstruction in feedback is presented. The control system contains discrete one-dimensionaland matrix proportional-integral-derivative controllers synthesized by the matrix inequality method usingthe plasma LPV model calculated on experimental data, and carries out the coordinated control of the plasmaposition and shape as well as the compensation for the scattered field of the central solenoid. The FCDI algorithmis improved for the operation in the real-time mode, and makes it possible to reconstruct the plasmashape in 20 μs. The digital control system with a feedback algorithm was simulated on a real-time test bench,consisting of two Speedgoat Performance Real-Time Target Machines (RTTM), and demonstrated the averageTask Execution Time (TET) value in 67 μs.

Study of an improved single neuron PID control algorithm in the Tokamak plasma density control system

Tokamak is an important device for controlled nuclear fusion research. The plasma electronic density control system (PEDCS) is an important system for controlling the Tokamak discharge process, which should be of high stability, rapidity, and accuracy. Gas seeding systems are widely used in many Tokamak devices to achieve plasma electronic density control. According to the mechanism model analysis for the plasma electronic density object, an adapted single neuron proportion integration differentiation (PID) control algorithm with the radial basis function (RBF) neural network tuning is studied. The principle and the implementation of the intelligent control algorithm are described in detail in this paper. The intelligent controller enables the system to optimize the PID parameters online according to the density state in the discharge process. The experimental results show that the adapted algorithm achieves a good control effect and also improves the control performance. The proposed method provides a useful reference for Tokamak devices and other similar control systems.

Tungsten dust injection into the STOR-M tokamak

Controlled tungsten (W) dust injection experiments have been performed in the STOR-M tokamak. Tokamak discharges were initiated after a known delay time with respect to the injection of W micro-spheres. The plasma parameters were detrimentally affected, proportional to an increased amount of dust within the plasma volume. High speed camera imaging of the injected dust revealed that the dust particles experience a strong force in the counter-plasma current direction. This force decreased for high dust number density due to the associated decline of plasma parameters.

Ex-situ quantification of impurity deposition depth on HL-2A divertor graphite tile by laser-induced breakdown spectroscopy

During tokamak discharges, the Plasma Wall Interaction (PWI) would lead to impurity deposition and fuel retention on the surface of the Plasma Facing Materials (PFMs). Laser-induced breakdown spectroscopy (LIBS) has emerged as a promising technique to facilitate in-situ, real-time, online characterization of the impurity deposition and fuel retention on the wall of tokamaks for PWI studies. The purpose of this investigation is to quantify the impurity deposition layer on the graphite tile removed from the dome region of the HL-2A tokamak lower divertor by using the ex-situ LIBS approach. The investigations indicate that the primarily deposited impurity elements are Fe, Cr, Ni, Mn, Ca, Al, Cu, Si, C, which may originate from the stainless steel first wall, divertor, limiter or the siliconized wall conditioning due to the PWI process in HL-2A tokamak. This has been further supported by observing a significantly positive correlation between the Cr, Ni, Mn, Al and Fe elements by spectral cumulative intensity linear regression analysis. The 3D morphology of the laser-ablated craters is reconstructed using a confocal microscope, revealing the laser ablation rates for the deposition layer and the substrate graphite tile to be 348 nm/pulse and 220 nm/pulse, respectively, under the laser fluence of 10 J/cm2. Then the thickness of the deposition layer is determined by using the correlation coefficient method. Along the poloidal direction of HL-2A tokamak, the deposited impurity shows an inhomogeneous pattern, with the thickness of the deposition layer varying from 1.74 μm to 5.92 μm. Finally, the absolute depth distribution of each element is also quantitatively measured.

Cross-machine comparison of runaway electron generation during tokamak start-up for extrapolation to ITER

A cross-machine comparison of global parameters that determine the runaway electron (RE) generation and loss process during tokamak start-up was carried out with the aim to extrapolate these to ITER. The study found that all considered discharges, also those that do not show signs of RE, are non-thermal at the start, i.e. have a streaming parameter larger than 0.1. During the current ramp-up the electric field, E, remains above the critical value, E c, that allows RE in the plasma. The distinction to be made is not if RE can form but, if sufficient RE can form fast enough such that they are detected or start to dominate the dynamics of the tokamak discharge. The dynamics of the value of E, density and temperature during tokamak are key to the formation of RE. It was found that larger devices operate with E closer to E c, due to their higher temperatures, hence the RE generation is relatively slower. The slower time scales for the formation of RE, estimated to be of the order of 100s of ms in ITER simplifies the development of avoidance schemes. The RE confinement time is also an important determinant of the entire process and is found to increase with the device size. The study also revealed that drift orbit losses, a mechanism often attributed as the main RE loss mechanism during the early tokamak discharge, are actually more difficult to achieve. RE losses might be more likely attributed to RE diffusion due to magnetic turbulence.

Drift of ablated material after pellet injection in a tokamak

Pellet injection is used for fuelling and controlling discharges in tokamaks, and it is foreseen in ITER. During pellet injection, a movement of the ablated material towards the low-field side (or outward major radius direction) occurs because of the inhomogeneity of the magnetic field. Due to the complexity of the theoretical models, computer codes developed to simulate the cross-field drift are computationally expensive. Here, we present a one-dimensional semi-analytical model for the radial displacement of ablated material after pellet injection, taking into account both the Alfvén and ohmic currents which shortcircuit the charge separation creating the drift. The model is suitable for rapid calculation of the radial drift displacement, and can be useful for e.g. modelling of disruption mitigation via pellet injection.

Statistical analysis of similarity of plasma parameters profiles at quasi-stationary stage of discharge in JET tokamak

A statistical analysis is carried out of the similarity of profiles of plasma parameters: electron temperature T e, density n e, and pressure P e, at the stage of quasi-stationary plasma current (the so called flat-top stage) of about 9000 discharges in the JET tokamak, among the discharges from #84458 to #99419 that covers almost the entire database of successful discharges in the JET ITER-like wall machine. For all these parameters, the existence of universal profiles is shown as functions of the normalized minor radius of the plasma column ρ. The discharge universal profiles (DUPs) are obtained by dividing the space-time-dependent profile by its value at the center of the plasma column or by the space-averaged value of this parameter in the region ρ ⩽ ρ max = 0.5–1, and subsequent averaging over time at the flat-top stage of each discharge. The machine universal profiles (MUPs) are obtained by averaging over time of flat-top stage of all discharges. It is shown that for 86% of the discharges, the time-averaged relative root-mean-square deviation of P e profile from the DUP is less than 20% at ρ ⩽ 0.8, and for the MUP, it is less than 17%. A similar picture is observed for the profiles of T e and n e/√T e. An auxiliary heating, P aux ⩾ 1 MW or 10 MW, was present at, respectively, 44% and 21% of the total duration of the flat-top stage of discharges. It is shown that large, ∼100% in magnitude, jumps of T e at ρ = 0, caused by the switch-on of auxiliary heating, can be described with a ∼20% accuracy by the jumps of the space-averaged temperature if the DUP is used to describe the stationary shape of the temperature spatial profile. The results obtained illustrate the degree of plasma self-organization in tokamaks.

Use of intrinsic hysteresis for the active control of internal transport barriers in magnetically confined fusion plasmas

In magnetically confined fusion devices, control of internal transport barriers (ITBs) is important both to enhance and suppress the turbulent transport to improve confinement control. Barrier control should allow for the improvement of confinement to aid in achieving the needed fusion criteria while also permitting the degradation of confinement to control profiles and clean the device by moving out the impurities accumulated near the core. In this work, we present a novel control scenario that takes advantage of the hysteresis intrinsic to transport barriers to easily cycle through enhanced and degraded confinement regimes. The control scenario is illustrated using a five-field simplified transport model for an ITB using typical parameters of a neutral beam injection-heated DIII-D tokamak discharge. Pellets and ion cyclotron resonance frequency power are used as control knobs for this active control scenario. These knobs adequately modify at will the local gradients and, therefore, the growth rates and shearing rates, allowing for easy and efficient control of the barrier by taking advantage of the barrier hysteresis. The result is a control cycle that could be operated with a relatively small amount of power in high performance regimes which, nowadays, typically require large power to control. It may also have advantages to avoid, or at least ameliorate, the appearance of magnetohydrodynamic instabilities in the barrier region.

Experimental and numerical modeling of plasma start-up assisted by electron drift injection on J-TEXT

Start-up is one of the critical phases for tokamak discharge. The electron drift injection (EDI) system has been developed on J-TEXT for start-up studies. A breakdown experiment with EDI-assisted start-up has been conducted, which verified the effect of pre-ionization by EDI to achieve start-up at a lower ohmic field voltage. A zero-dimensional (0D) model has been developed to explain the effect of EDI quantitatively. The comparison between the experiment and simulation verified the credibility of this model. Based on this model, a comparison between pure ohmic heating start-up and EDI-assisted start-up was presented, showing that EDI improved ionization, causing a lower delay to the peak of hydrogen ionization and radiation losses and a smoother rise in the electron and ion energy. This result quantitatively verified the pre-ionization effect of EDI on start-up . The effects of injecting different currents and electron energy were investigated. A better pre-ionization effect was realized by increasing the injected current, which can be a reference for the upgrading of the EDI system.

Automatic identification of edge localized modes in the DIII-D tokamak

Fusion power production in tokamaks uses discharge configurations that risk producing strong type I edge localized modes. The largest of these modes will likely increase impurities in the plasma and potentially damage plasma facing components, such as the protective heat and particle divertor. Machine learning-based prediction and control may provide for the automatic detection and mitigation of these damaging modes before they grow too large to suppress. To that end, large labeled datasets are required for the supervised training of machine learning models. We present an algorithm that achieves 97.7% precision when automatically labeling edge localized modes in the large DIII-D tokamak discharge database. The algorithm has no user controlled parameters and is largely robust to tokamak and plasma configuration changes. This automatically labeled database of events can subsequently feed future training of machine learning models aimed at autonomous edge localized mode control and suppression.

Electrostatic turbulence in EAST plasmas with internal transport barrier

Based on first-principles nonlinear gyrokinetic simulations, the electrostatic turbulence properties in the internal transport barrier (ITB) region of an Experimental Advanced Superconducting Tokamak discharge (#93890) are investigated. Specifically, ITBs with steep density and temperature gradients are located in the weakly negative magnetic shear region at the plasma center. In the linear stage, the growth rate and frequency of the ion temperature gradient (ITG) mode increase significantly due to resonant excitation by trapped electrons. That is, the resonance between trapped electrons and the ITG becomes strong due to the precession drift reversal of trapped electrons by the negative magnetic shear and Shafranov shift. Meanwhile, the trapped electron mode is stable in the ITB region due to only a very small fraction of electrons precessing in the direction of the electron diamagnetic drift. Nonlinear simulations show that, after considering the non-adiabatic effect of trapped electrons, the heat conductivity of ions and the turbulence intensity increase by at least a factor of 7 compared with the results only considering the adiabatic effect of electrons. The zonal charge density of trapped electrons can partially cancel that of ions, which weakens the intensity of the zonal flow, and consequently reduces the zonal flow regulation and enhances the turbulent transport.