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Studying stationary regimes with high plasma confinement in a tokamak with reactor technologies (TRT) [1] involves calculating the plasma stability taking into account the influence of the current density profiles and pressure gradient in the pedestal near the boundary. At the same time, the operating limits should be determined by the parameters of the pedestal, which, in particular, are set by the stability limit of the peeling–ballooning modes that trigger the peripheral disruption of edge localized modes (ELM). Using simulation of the quasi-equilibrium evolution of the plasma by the ASTRA and DINA codes, as well as a simulator of magnetohydrodynamic (MHD) modes localized at the boundary of the plasma torus based on the KINX code, stability calculations are performed for different plasma scenarios in the TRT with varying plasma density and temperature profiles, as well as the corresponding bootstrap current density in the pedestal region. At the same time, experimental scalings for the width of the pedestal are used. The obtained pressure values are below the limits for an ITER-like plasma due to the lower triangularity and higher aspect ratio of TRT plasma. For the same reason, the reversal of magnetic field shear in the pedestal occurs at a lower current density, which causes the instability of modes with low toroidal wave numbers and reduces the effect of diamagnetic stabilization.

Chinese Fusion Engineering Testing Reactor (CFETR) is a superconducting magnet Tokamak, which has the equivalent scale with complementary function to International Thermonuclear Experimental Reactor (ITER). The vacuum vessel (VV) which has a double-layer structure, Cooling water circulating through the double-layer structure will remove the heat generated during operation. The VV will provide a high-vacuum environment for the plasma, improve radiation shielding and plasma stability and provide support for in-vessel components. The CFETR VV is composed of 16 sectors, the angle of each sector is 22.5°. The Research and Development (R&D) of the key technologies to the VV manufacture have been carried out a few years ago by Institute of Plasma Physics Chinese Academy of Science (ASIPP), including Narrow-Gap welding, cutting and non-destructive testing, Poloidal Segment (PS) and sector assembly technologies, etc. ASIPP is constructing a 1/8 sector real size VV mock-up now. The manufacture of the PS for first 1/32 sector VV mock-up has been completed in 2015, the PS will be assembled into a whole this year. This paper will describe the study of the assembly technology for the CFETR 1/32 Vacuum Vessel mock-up.

High fidelity kinetic plasma equilibrium reconstructions are an essential requirement for accurate stability and disruption prediction analyses to support continuous operation of high beta tokamak plasmas. The present kinetic equilibrium reconstructions of plasmas in the KSTAR device include plasma density and temperature profiles from Thomson scattering and ion temperature from charge exchange spectroscopy diagnostics, and allowance for fast particle pressure. In addition, up to 25 channels of motional Stark effect (MSE) diagnostic data are used to constrain the magnetic field pitch angle profile in the plasma to produce a reliable computation of the safety factor, q, profile. H-mode plasmas exhibit clear pedestal characteristics in the reconstructed pressure profile compared to internal transport barrier or L-mode plasmas. The plasma configuration and vertical position of inner strike points are validated by CCD and infrared camera images. Ideal and resistive MHD stability analyses using the DCON and resistive DCON codes utilize these kinetic equilibrium reconstructions to compare to experimental plasma stability. Equilibria with sufficiently low convergence error can provide reliable computation of ideal and resistive magnetohydrodynamic (MHD) stability analysis.

In the last 3 decades, giant steps in technology and simulations capability have been made for the benefit of all magnetic fusion concepts. This is especially true for virtually a one-of-a-kind device, a microwave-driven bumpy-torus magnetic-fusion-energy system. In the 1980s, the prototype experiment was technology limited, and it was judged based on unrealistic expectations from an incomplete understanding of the machine performance. For those experiments, the machine did not achieve the theoretical stability threshold to demonstrate good plasma confinement. This paper examines the premise that the experimental platform was underpowered, which prevented acceptance of its value for fusion. In this paper, the experimental plasma scaling data with power are reviewed. If, with the advancements of technology and simulations capability of the last 30 years, stable operation can be proven, a much improved device competitive with tokamak and stellarator performance may be possible. The advantages of this concept include steady-state operation, gentle shut down, modular construction for fabrication and maintenance, and the natural diversion of impurities.

In this contribution we have presented two techniques for determination of plasma equilibrium position in IR-T1 tokamak: Relaxation and optical techniques. An analysis method of tokamak plasma equilibrium by a relaxation method with specified magnetic axis was presented. The degrees of freedom due to designated positions of the magnetic axis are possible by using poloidal field coil currents. Stable steady state tokamak plasma equilibria are calculated along with the Magnetohydrodynamic (MHD) potential energy. The plasma generates a plasma current which partially or fully cancels the magnetic field from the poloidal field coils. For low-temperature plasmas, the plasma current distribution is centrally peaked; for high-temperature plasmas, the plasma current has a hole. A centrally peaked current distribution in a low-temperature plasma is evolved into a current distribution with a hole by increasing the plasma pressure by Ohmic heating, RF heating, or by neutral beam injection heating. In the second technique, an image processing technique is used for the output signal of CCD camera. Then, plasma emission intensity profile and plasma position were obtained. Results were compared and discussed.

Controlled thermonuclear fusion via magnetic confinement, still in experimental stage, has the potential to become a viable and environment-friendly solution to the energy problem, especially for the high-power needs of modern industry. In order to optimize the operation of devices based on the tokamak principle, automatic control systems are envisaged to fulfill the requirements for the magnetic equilibrium and plasma stability, with copper coils, neutral gas injectors and microwave sources used as actuators. In present-day experiments, the implemented control loops are simple and practical, however in future devices like ITER (presently under construction) more sophisticated control design will be required, based on realistic closed-loop simulations with efficient computational tools and real-time diagnosing. For magnetohydrodynamic instability control, the system should include physics/engineering models for the plasma dynamics, the wave actuation and the diagnostic sensors, as well as controllers based on classical or modern principles. In this work, we present a model for a specific design of a controlled electromechanical millimeter-wave launcher, which executes the major part of the wave actuation, and perform numerical simulations of its open-loop dynamics and closed-loop control for scenarios relevant to tearing mode stabilization in medium-sized tokamaks.

Ignition is the point at which a nuclear fusion of hydrogen isotopes reaction becomes self-sustaining. This occurs when the energy being given off by the fusion reactions heats the fuel mass more rapidly than various loss mechanisms cool it. At this point, the external energy needed to heat the fuel to fusion temperatures is no longer needed. As the rate of fusion varies with temperature, the point of ignition for any given machine is typically expressed as a temperature. On the other hand, energy confinement time is expressed by an equilibrium state. Analytical solutions of Grad-Shafranov equation (GSE) can be used for theoretical studies of plasma equilibrium, transport and Magnetohydrodynamic stability of tokamaks. In this paper we have presented two families of exact solutions. With applying these solutions to IR-T1 tokamak, a small, air core, low beta and large aspect ratio tokamak with a circular cross section, we calculated poloidal magnetic flux. Due to the generality and high accuracy of the second exact solution for all of the magnetic configurations of interest, the result of this solution for IR-T1 tokamak is good (tokamak-relevant) equilibrium compare to the first exact solution for this tokamak. We intend to use these analytical solutions as benchmark of numerical equilibrium codes.

Analytical solutions of electromagnetism driven Grad-Shafranov equation (GSE) can be used for theoretical studies of plasma equilibrium, transport and Magnetohydrodynamic stability of tokamaks. In this research, an extended analytic solution to the Grad-Shafranov equation was presented. The solution describes standard tokamaks configurations. It allows the simulation magnetic surfaces of plasmas with elongation and triangularity, with an independent choice of pressure and plasma current. We have determined equilibrium magnetic surfaces in IR-T1 tokamak. With applying these solutions to IR-T1 tokamak, a small, air core, low beta and large aspect ratio tokamak with a circular cross section, we have calculated poloidal magnetic flux. Due to the generality and high accuracy of the first exact solution for all of the magnetic configurations of interest, the result of this solution for IR-T1 tokamak is acceptable equilibrium compare to the second one for this tokamak. These analytical solutions can be used as an input of feedback control system.

Nuclear fusion is a nuclear reaction in which two or more atomic nuclei (such as a deuterium–tritium) come very close and then collide at a very high speed and join to form a new high energy nucleus (Helium). Determination of accurate plasma horizontal position during plasma discharge is essential to transport it to a control system based on feedback. The solutions of Grad-Shafranov equation (GSE) analytically can be used for theoretical studies of plasma equilibrium, transport and magneto hydrodynamic stability. Here we have presented specific choices for source functions, kinetic pressure and poloidal plasma current, to be quadratic in poloidal magnetic flux and derive an analytical solution for Grad-Shafranov equation. With applying this solution to NSTX tokamak, we calculated poloidal magnetic flux, toroidal current density and normalized pressure profiles for this tokamak. Toroidal and poloidal flows can considerably change the equilibrium parameters of tokamak. These effects on the equilibrium of tokamak plasmas are numerically investigated using the code FLOW. As a comparative approach to equilibrium problem, the code is used to model equilibrium of NSTX tokamak for case pure toroidal flow. Comparison of the results of these two methods for NSTX tokamak shows good agreement between two and that our analytical solution can be served as good benchmark against the equilibrium code FLOW.

The main goal of this paper is the investigation of different techniques for improving the plasma confinement in IR-T1 Tokamak. For this purpose, magnetic fluctuations determined by Mirnov coils were analyzed using singular value decomposition (SVD) and wavelet techniques. Also, electric fluctuations specified by Langmuir-Ball pen probe were studied and analyzed for calculations of plasma potential and electron temperature. In this research, investigation was also made into influence of Resonant Helical Magnetic Field (RHF) (with both modes L = 2 and L = 3) on plasma stability and plasma confinement. Results show that cold biased limiter with a positive voltage and emissive biased limiter with a negative voltage were the most effective condition on increase of plasma stability and reduction of magnetic fluctuations. Also, the results demonstrated that RHF with mode L = 3 has an effective role in plasma confinement improvement.

It is very important to calculate the equilibrium magnetic field configuration exactly and to estimate the parameters of the device needed for the equilibrium and stability in the design and operate the low-aspect-ratio tokamak, one of the axis-symmetrical torus devices which is widely used in the research of nuclear fusion using plasma. In the previous researches, there were the theoretical and numerical interpretation methods for the high-aspect-ratio tokamak plasma. But for the low-aspect-ratio tokamak plasma, numerical methods such as the finite element method and iteration method are usually used. In this research, we interpreted and examined the features of magnetohydrodynamic equilibrium of the low-aspect-ratio tokamak plasma in combination with the Green function method and finite element method which has high numerical value accuracy.

Spontaneous toroidal rotation of the plasma is observed in the spherical tokamak QUEST with the help of electron cyclotron resonance (ECR) heating and without the use of any externally injected momentum. Several vertical magnetic field ( $B_{z}$ ) configurations with varying mirror ratio ( $M$ ) (a measure of field curvature) are applied and evolution of rotation is studied with the help of Doppler spectroscopy of bulk and impurity ions. Significant toroidal rotation ( $V_{\phi } \sim 6$ km/s) is initiated in the open magnetic field configuration during the initial plasma breakdown phase, which is later sustained ( $V_{\phi } \sim 20$ km/s) in a closed magnetic field configuration in steady state. Rotational velocity is primarily along the cocurrent direction and is found to be proportional to the $B_{z}$ strength and the resulting plasma current. High $M$ and $B_{z}$ are demonstrated to be the two specific external controls by which rotation can be initiated in the plasma. The rotation in open field lines is found to be initiated at the ECR layer in the slablike plasma, which is evolved to produce a sustained rotation in the natural divertor inboard poloidal field null equilibrium in QUEST.

The present contribution gives an overview of our work on non-conventional magnetic coil sensors for diagnostics and plasma stability control of nuclear fusion experiments in tokamaks. Instead of wire wound around a core, these devices consist of printed conductor wire coils on ceramic substrates, and are based on LTCC (low-temperature co-fired ceramic) and thick-film technology, which allow creation of monolithic multilayer coils with excellent stability. For 3D sensing, an innovative modular design combining LTCC coils and an alumina base has been developed. Finally, the important aspects of integration, manufacturing, mounting and interconnection are discussed.

The operation scenarios for plasma breakdown and current ramp-up phases in JT-60SA tokamak have been developed and verified in simulation using the TOSCA code. The induced current in the conducting elements such as vacuum vessel and stabilizing plate increases to the comparable level of plasma current of ∼600 kA during the breakdown phase and thus enhances the strength of stray field. The optimized scenarios for half and full pre-magnetization cases satisfied the conditions required for the plasma initiation. At the initial plasma, the vertical magnetic field required to sustain the plasma position was controlled by the outer equilibrium field (EF) coil currents which compensate for a vertical field due to a large eddy current. The condition for the formation of divertor configurations given by the combination of the magnetic flux for plasma and the plasma current has been satisfied which enables us to develop the operational scenarios with a smooth transition from a limiter to a divertor configuration.

In tokamaks, plasma emits as a volumetric Soft-X-ray (SXR) source. Emitted X-rays can give very useful information about plasma stability, shape and impurity content. Measuring the Soft X-ray (SXR) radiation ([0.1–20 keV]) of magnetic fusion plasmas is a standard way of accessing valuable information on particle transport and MagnetoHydroDynamic. Generally, like at Tore Supra in France, the analysis is performed with a 2D tomographic system composed of several cameras equipped with detectors like Silicon Barrier Diodes spread in periphery of the tokamak. Unfortunately, the strong constraints imposed by the environment of a tokamak reactor (high neutron fluxes, gamma and hard X-ray emission, high magnetic field and high radiofrequency powers) do not authorize to install in a close vicinity of the machine such detectors. We have thus investigated the possibility of using polycapillary lenses to transport the SXR information to several meters from the plasma, not necessarily in a straight line. The idea is to protect the SXR detector from the entire environment by a proper shielding. Different polycapillary lenses could be used for that purpose and have been tested in collaboration with CELIA (CEA–CNRS) of Bordeaux. Transmission of the order of 20% where observed for the low energetic part of the spectrum (down to 3 keV) while still 10% were observed for the remaining part (from 3 to 10 keV). In parallel a model of polycapillary transmission has been developed and validated against experiment. Results are presented confirming the great potential of polycapillary lenses for SXR transmission in tokamak plasma. Studies of the influence of geometrical parameters like diameter and curvature of the channels, on the photons transmission is also presented.

In the MHD tokamak plasma theory, the plasma pressure is usually assumed to be isotropic. However, plasma heating by neutral beam injection and RF heating can lead to a strong anisotropy of plasma parameters and rotation of the plasma. The development of MHD equilibrium theory taking into account the plasma inertia and anisotropic pressure began a long time ago, but until now it has not been consistently applied in computational codes for engineering calculations of the plasma equilibrium and evolution in tokamak. This paper contains a detailed derivation of the axisymmetric plasma equilibrium equation in the most general form (with arbitrary rotation and anisotropic pressure) and description of the specialized version of the SPIDER code. The original method of calculation of the equilibrium with an anisotropic pressure and a prescribed rotational transform profile is proposed. Examples of calculations and discussion of the results are also presented.

Plasma confinement plays an important role in fusion study. Applying an external voltage using limiter biasing system is proved to be an efficient approach for plasma confinement. In this study, the position of the limiter biasing system was changed to investigate the effect of applying external voltages at different places to the plasma. The external voltages of ±200 V were applied at the different positions of edge, 5 mm and 10 mm inside the plasma. Then, the main plasma parameters were measured. The results show that the poloidal turbulent transport and radial electric field increased about 25–35% and 35–45%, respectively (specially when the limiter biasing system was placed 5 mm inside the plasma). Also, the Reynolds stress is experienced its maximum reduction about 5–10% when the limiter biasing system was at 5 mm inside the plasma and the voltage of +200 V was applied to the plasma. When the limiter biasing system move 10 mm inside the plasma, the main plasma parameters experienced more instabilities and fluctuations than other positions.

One of the most promising candidates to produce clean nuclear fusion energy is the tokamak, a device magnetically confining an extremely hot plasma (i.e. an ionized gas) where the fusion reactions take place. To produce nuclear fusion energy using tokamak devices, it is crucial that the poloidal magnetic flux (characterized by the so-called q profile) and the plasma internal energy are tightly controlled to avoid magnetohydrodynamic instabilities and to reach the high pressures and temperatures that are needed for high fusion-power density. Simultaneous control of the q profile and the internal energy is challenging for a number of reasons: the system is nonlinear, there are significant parameter uncertainties and large disturbances, the available number of actuators is small, and the actuation authority is limited from a control perspective. Therefore, a variable-structure controller is proposed in this work to tackle this plasma control problem since this type of controllers can typically diminish the impact of serious disturbances and nonlinearities while still leading to good performance. Simulations and recent experiments on the DIII-D tokamak in a challenging high-confinement (H-mode) plasma regime show that this control approach does indeed lead to good and repeatable control of the q profile and the internal energy.

Tokamak plasma rotation is widely recognized for its importance to heat confinement and plasma stability. In this work we consider control of the plasma rotation profile with the aim of building a control strategy suitable for testing various rotation profiles for stability characteristics and reaching desired operating conditions. To obtain a control-oriented model of the toroidal rotation profile evolution, a simplified version of the momentum balance equation is combined with scenario-specific models for the momentum sources. Various momentum sources including on-axis and off-axis neutral beam injection and the non-axisymmetric field coils, which provide rotation damping, allow not only control of the bulk plasma rotation, but also control of the profile shape. A feedback controller is designed in a model predictive control framework to regulate the rotation profile while satisfying constraints associated with the desired plasma stored energy and β (kinetic to magnetic pressure ratio) limits.

In the present RF-driven plasma with a lot of high-energy electrons, there may be anisotropic plasma pressure, which makes the usual equilibrium analysis difficult, but the Cauchy condition surface method can reconstruct the plasma shape precisely regardless of the anisotropy. In addition, the plasma current effect in the open magnetic surfaces outside of the closed magnetic surfaces is considered in the RF-driven divertor plasma. In the reconstruction process, singular value (SV) decomposition is used and optimal criterion function for generalized cross validation is estimated concerning truncation or reduction of the small-SV components.

One of the most promising devices for realizing power production through nuclear fusion is the tokamak. To maximize performance, it is preferable that tokamak reactors achieve advanced operating scenarios characterized by good plasma confinement, improved magnetohydrodynamic stability, and a largely noninductively driven plasma current. Such scenarios could enable steady-state reactor operation with high fusion gain, the ratio of produced fusion power to the external power provided through the plasma boundary. For certain advanced scenarios, control of the spatial profile of the plasma current will be essential. The complexity of the current profile dynamics, arising due to nonlinearities and couplings with many other plasma parameters, motivates the use of model-based control algorithms that can account for the system dynamics. A first-principles-driven, control-oriented model of the current profile evolution in low-confinement mode (L-mode) discharges in the DIII-D tokamak is employed to address the problem of regulating the current profile evolution around desired trajectories. In the primarily inductive L-mode discharges considered in this paper, the boundary condition, which is dependent on the total plasma current, has the largest influence on the current profile dynamics, motivating the design of a boundary feedback control law to improve the system performance. The backstepping control design technique provides a systematic method to obtain a boundary feedback law through the transformation of a spatially discretized version of the original system into an asymptotically stable target system with desirable properties. Through a nonlinear transformation of the available physical actuators, the resulting control scheme produces references for the total plasma current, total power, and line averaged density, which are tracked by existing dedicated control loops. Adaptiveness is added to the control scheme to improve upon the backstepping controller's disturbance rejection and tracking capability. Prior to experimental testing, a Simserver simulation was carried out to study the controller's performance and ensure proper implementation in the DIII-D Plasma Control System. An experimental test was performed on DIII-D to test the ability of the controller to reject input disturbances and perturbations in initial conditions and to demonstrate the feasibility of the proposed control approach.

In this contribution, we have presented two techniques for the determination of plasma equilibrium position in IR-T1 tokamak: relaxation and optical methods. An analysis method of tokamak plasma equilibrium by a relaxation method with a specified magnetic axis is presented. The degrees of freedom due to designated positions of the magnetic axis are possible by using poloidal field coil currents. Stable steady-state tokamak plasma equilibria are calculated along with the magnetohydrodynamic potential energy. The plasma generates a plasma current which partially or fully cancels the magnetic field from the poloidal field coils. For low-temperature plasmas, the plasma current distribution is centrally peaked; for high-temperature plasmas, the plasma current has a hole. A centrally peaked current distribution in a low-temperature plasma is evolved into a current distribution with a hole by increasing the plasma pressure by Ohmic heating, radio frequency heating, or by neutral beam injection heating. In the se...

The aim of this paper is to present an optimization procedure for the design of the Poloidal Field (PF) coil system of a tokamak reactor. Starting from the target plasma shape, a simplified toroidal field (TF) coil shape is derived to satisfy the basic requirements of bending-free coil and maximum (usually about 1%) TF ripple at the plasma boundary. Then, the PF coils, which are mechanically attached to the TF, optimal number, and locations, are computed. This computation is done by optimally matching at the desired plasma boundary the external magnetic flux that keeps the plasma magneto hydro dynamic (MHD) equilibrium with the magnetic flux generated by the PF coils. The external magnetic flux is matched at the plasma boundary by optimally selecting the principal components of the modified mutual inductance matrix M* that minimize the plasma boundary deformation from its reference shape. Since the M* accounts for the plasma response through the linearization of the Grad-Shafranov equation, the solution found is correct to first order and it substantially simplifies the lengthy, nonlinear computations that are presently used to achieve the same goal.

Determinations of the poloidal beta, internal inductance, plasma energy, plasma pressure, plasma temperature, plasma resistance, plasma effective atomic number, magneto-hydrodynamics (MHD) activity, Runaway electrons energy and energy confinement time are essential for tokamak experiments and optimized operation. Also some of the plasma information can be deduced from these parameters, such as plasma toroidal current profile, and MHD instabilities. In this contribution we investigated about measurements of some plasma parameters as well as MHD activity and Runaway electrons energy. For this purpose we used the magnetic diagnostics and a hard X-ray spectroscopy in IR-T1 tokamak. A hard X-ray emission is produced by collision of the Runaway electrons with the plasma particles or limiters. The mean energy was calculated from the slope of the energy spectrum of hard X-ray photons. In this paper in order to measure energy of the Runaway electrons, we obtained hard X-ray energy in every 5 ms intervals, from the beginning to the end of plasma. Results indicated mean energy of Runaway electrons is maximum during the 0–5 ms interval.

Research on controlled nuclear fusion has been largely concentrated on plasma confinement using toroidal magnetic fields. Toroidal systems are complex. A simpler magnetic confinement system may provide a valuable platform for understanding fusion plasmas. The linear mirror machine has delivered good performance with the potential of giving a direct conversion of nuclear energy into electric power. The GAMMA-10 (G-10) linear mirror confinement system at Tsukuba University demonstrated the principle of the direct conversion of plasma energy into electric power on a small scale from the exhaust plasma in the exterior divertor chamber. The tokamak fusion system has to prove that the 10 to 15 MA of plasma current can be sustained continuously with acceptable efficiency. Plasma confinement is due to the magnetic field from the plasma current in tokamaks. There is room for creative new solutions in the magnetic confinement of fusion plasmas, and consideration is given for the alternative approach of using a linear machine with high magnetic mirror fields and the direct conversion of the high temperature escaping plasma to electric power.

Extensive research has been conducted to find operating scenarios that optimize the plasma performance in nuclear fusion tokamak devices with the goal of enabling the success of the ITER project. The development, or planning, of these advanced scenarios is traditionally investigated experimentally by modifying the tokamak's actuator trajectories, such as the auxiliary heating/current-drive (H&CD) scheme, and analyzing the resulting plasma evolution. In this work, a numerical optimization algorithm is developed to complement the experimental effort of advanced scenario planning in the DIII-D tokamak. Two properties related to the plasma stability and performance are the safety factor profile ( q -profile) and the normalized plasma beta ( β N ). The optimization algorithm goal is to design actuator trajectories that steer the plasma to a target q -profile and plasma β N , such that the achieved state is stationary in time, subject to the plasma dynamics (described by a physics-based, nonlinear, control-oriented partial differential equation model) and practical plasma state and actuator constraints, such as the maximum available amount of H&CD power. This defines a nonlinear, constrained optimization problem that we solve by employing sequential quadratic programming. The optimized trajectories are then tested through simulation with the physics-based model and experimentally in DIII-D.

In this work the effects of Cold Biased Limiter (CBL) and Emissive Biased Limiter (EBL) have been individually investigated, in both positive and negative polarities, on the width and frequency of magnetic islands. The effects were examined using singular value decomposition and wavelet techniques on mirnov coil signals. The results show that the application of EBL with both positive and negative polarities has been more effective on plasma stability compared with CBL. Comparing different polarities of CBL and EBL revealed that the positive polarity was more effective on the width of magnetic islands than negative polarity. The greatest impact occurs during EBL with positive polarity, in which reduction is observed in both width of magnetic islands and emission of Hα radiation. Besides, intensity and frequency of magnetic islands are reduced from 50 to 25 kHz in around 1 ms after bias application. Meanwhile, the minimal effect on width and frequency of magnetic field occurs in CBL with negative polarity.

Edge plasma turbulence was investigated over a wide range of plasma and field parameters in the IR-T1 tokamak for the first time. Fluctuation levels and spectra were measured using two arrays of Langmuir probe in the region r/a = 0.75–1.2. Under almost all conditions the edge plasma was turbulently unstable, with a broadband fluctuation spectrum in the drift wave range of frequencies f = 10–1000 kHz. A stable state was observed only in the very cold, low-current discharge formed at unusually high neutral filling pressure. Otherwise, the relative fluctuation level as monitored by the ion saturation current was very high, in the range $$ {{\tilde{\rm J}^{ + } } \mathord{\left/ {\vphantom {{\tilde{\rm J}^{ + } } {\bar{\rm J}^{ + } }}} \right. \kern-0pt} {\bar{\rm J}^{ + } }} \cong 0.2{-}0.8 $$ , while the fluctuation power spectra were roughly invariant in shape. The relative fluctuation level was always highest near the wall and decreased monotonically toward the plasma centre.

Many tokamaks have observed that sawteeth of sufficient duration may trigger neoclassical tearing modes (NTMs) that lead to plasma performance degradation. In this paper, TCV's ability to accurately control the period of individual sawteeth is exploited, using localized electron cyclotron resonance heating and current drive (ECRH and ECCD), to trigger NTMs under controlled conditions, providing an excellent environment for the study of the seeding of NTMs by sawtooth crashes. The TCV experiments show evidence of a fast formation of seed islands with poloidal/toroidal mode numbers m/n = 3/2 and 2/1 within a few tens of microseconds following the sawtooth crash. Crashes of sawteeth with a longer period duration are observed to generate larger seed islands but also increase the plasma stability to conventional tearing modes. While these two effects compete, the NTM stability is found to decrease with increasing sawtooth period. The seed island size can be reduced and, thereby, the NTM stability improved, by increasing the value of the safety factor q(95). Alternatively, NTM stability can be increased by application of preemptive ECRH at the resonant surface of the NTM. Preemptive ECRH is found to enlarge the plasma operational domain by improving the conventional tearing stability and by reducing the coupling between the driving (m/n = 1/1 or 2/2) and the driven modes (m/n = 2/1 or 3/2), resulting in smaller sawtooth generated seed islands. The efficiency of preemptive ECRH increases when sufficient ECRH power is applied in a short time interval prior to the sawtooth crash.

The ISTTOK tokamak ( Ip = 4 kA, BT = 0.5 T, R = 0.46 m, a = 0.085 m) is one of the few tokamaks with regular alternate plasma current (AC) discharges scientific programme. In order to improve the discharge stability and to increase the number of AC discharge cycles a novel control system was developed. The controller acquires data from 50 analog-to-digital converter (ADC) channels of real-time diagnostics and measurements: tomography, Mirnov coils, interferometer, electric probes, sine and cosine probes, bolometer, current delivered by the power supplies, loop voltage and plasma current. The system has a control cycle of 100 μs during which it reads all the diagnostics connected to the advanced telecommunications computing architecture (ATCA) digitizers and sends the control reference to ISTTOK actuators. The controller algorithms are executed on an Intel ® Q8200 chip with 4 cores running at 2.33 GHz and connected to the I/O interfaces through an ATCA based environment. The real-time control system was programmed in C++ on top of the Multi-threaded Application Real-Time executor (MARTe). To extend the duration of the AC discharges and the plasma stability a new magnetising field power supply was commissioned and the horizontal and vertical field power supplies were also upgraded. The new system also features a user-friendly interface based on HyperText Markup Language (HTML) and Javascript to configure the controller parameters. This paper presents the ISTTOK control system and the consequent update of real-time diagnostics and actuators.

A method is proposed for stability analysis of the locked and rotating resistive wall modes (RWMs) in tokamaks. The method is based on the relations describing the balance of energy permeating through the vessel wall. This is a natural extension of the traditional energy approach to the plasma stability tasks which allows incorporation of the energy outflow (absent in the classical energy principle) and its dissipation in the wall. The proposed method covers the locked and rotating modes with a complex growth rate. Its efficiency is proved by derivation of a general dispersion relation for such modes with further reduction to particular consequences for slow and fast RWMs. It is shown that in the latter case, when the skin depth becomes smaller than the wall thickness, the mode rotation essentially amplifies its damping, weakening and even suppressing the instability. This effect was earlier found in the frame of the slab model [V. D. Pustovitov, Phys. Plasmas 19, 062503 (2012)]. Here, it is confirmed with equations valid for toroidal geometry, which are obtained as a supplement to the standard energy principle. The presented results predict strong rotational stabilization of the fast RWMs, which occurs at the mode rotation frequency above a critical level. The estimates are given to allow comparison of these predictions with experimental results.

The main purpose of this paper is the determination of plasma equilibrium parameters, as well as current density and safety factor profiles, in IR-T1 tokamak. For this purpose, a diamagnetic loop with its compensation coil and an array of magnetic probes were designed, constructed, and installed on the outer surface of the IR-T1 chamber, and then, the poloidal beta and poloidal and radial magnetic fields were measured. Moreover, a few approximate values of the internal inductance for different possible profiles of the plasma current density were also calculated. Then, from these two techniques results, the plasma horizontal displacement was determined. In addition, from the results, the current density and q-profiles were obtained.

The tokamak is a magnetic-confinement device where a plasma is confined with the final purpose of generating power from fusion reactions. Unfortunately, working points with favorable fusion conditions in tokamaks are normally found in a region in which the plasma may be thermally unstable. Therefore, regulation of the plasma temperature and density to produce a certain amount of fusion power while avoiding thermal instabilities, known as burn control, is one of the key issues that need to be solved for the success of burning plasma tokamaks such as ITER. Most previous controllers make use of approximate linearization techniques. In the present work, a model-based control approach using nonlinear techniques is proposed. This nonlinear control approach avoids approximate linearization of the model, is applicable to a greater range of operating conditions, and is stable against a larger set of perturbations. In addition to conventional actuation, like modulation of the auxiliary power and modulation of the fueling rate, the in-vessel coil system is considered as a new actuator. The in-vessel coils have the capability to generate non-axisymmetric magnetic fields that modify the plasma confinement, which influences the plasma energy dynamics. A model is proposed to account for the influence that the in-vessel coil actuation has on the plasma confinement. Finally, the effectiveness of the controller is demonstrated via a simulation study for an ITER-like scenario.