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This paper presents a spectral approach to H-optimal linear synthesis with the application for tokamak plasma control systems. Two main problems of a synthesis are discussed: the mean square optimization and the appropriate guaranteeing interpretation. The simple computational technique to solve these problems is proposed. Their implementation is illustrated by the example on the base of one variant of ITER tokamak model for the plasma vertical feedback control.

Mathematical models of the structural parametric optimization of plasma dynamics are discussed. Optimization approach to plasma dynamic is based on the consideration of trajectory ensemble. This ensemble describes transient process in tokamak subject to the initial data and external disturbances. In the framework of this approach the optimization of dynamics of the trajectory ensemble in ITER tokamak is given. The trajectories of this ensemble are perturbed at the initial point set and the set of external disturbances.

The goals of the KSTAR project are to develop a steady-state-capable advanced superconducting Tokamak and establish a scientific and technological basis for a Korean nuclear fusion power station. The KSTAR Tokamak comprises a magnet system, vacuum vessel, and cryostat, thereby facilitating vacuum conditions for plasma gas at high temperatures, along with low-temperature helium gas for cooling. The TF coil structure, a part of the magnet system, is constructed and jointed with 16 pieces at 22.5-degree intervals using a conical bolt and shear key. The main function of the conical bolt in the inner and outer inter-coil structures is to resist the in-plane and out-of-plane forces and increase the toroidal and intercoil shear stiffness. Therefore, the conical bolt must be dimensionally optimized to reduce the stresses at each connecting part. Accordingly, shape optimization of the conical bolt was carried out using SZGA, and the stresses were analyzed by ANSYS.

The relation between current density and mass density in the plasma has been derived based on the assumption that the magnetic entropy is constant in time. Over the entire range of quasi-stationary ohmic conditions, it is obtained . Here j₀ and ρ₀ are the central values of j and ρ with j₀ ≈ cB/2πR₀ (this corresponds to a value q ≈ 1 of the safety factor near the minor axis).

In this work, we present approximation of plasma position by two methods on HT-7 Tokamak. In the first method, solution of Grad–Shafranov equation are presented and one relation for plasma position is obtained. A generalized Grad–Shafranov-type equation has been used. Specific functional forms of plasma internal energy and current are used. In second method, according to magnetic fields distribution around the plasma, second relation for plasma position is obtained. Plasma position is obtained by magnetic probes. Results of the two methods are in good agreement with each other.

The structure of magnetic field lines in toroidal fusion plasmas, as in tokamaks and stellarators, represents the lowest-order description of the plasma particle behavior, up to finite Larmor and drift effects. Tokamaks with reversed magnetic shear typically present internal transport barriers that help to improve confinement through a partial or total reduction of the particle transport across magnetic surfaces. In this work, we investigate numerically particle escape in tokamaks with reversed shear in order to identify fractal structures affecting transport. These structures are quantitatively evaluated using two basic measures: the basin entropy and the Wada property.

The consideration of the three main nuclear reactions of the hydrogen isotopes (D(D, n)³He, D(D, p)T and T(D, n)⁴He) only leads to wrong results for determination of plasma parameters in magnetic confinement fusion reactors. The nuclear reaction ³He(D, p)⁴He influences the amount of produced tritium since it makes an important contribution to the charged particle energy deposition and to the temperatures. In this paper, we have considered different nuclear reactions of Deuterium–Tritium (D–T) fusion in tokamak reactor. This study has been carried out on the base of the particle and power balance equations in a zero-dimensional model then plasma parameters have been calculated. Finally, the obtained results have been compared with the theoretical results reported by other researchers.

The application of superconductivity to the large magnets required for charged particle spectroscopy in high energy physics experiments, and for plasma containment in fusion experiments, has resulted in a spectacular leap in the efficiency of these devices. First applied in the late 1960s, the technology has progressed to meet increasingly demanding goals of the experiments and has stimulated important development of the associated conductors. In this article we describe briefly the basic requirements that determine the design of the different types of magnets. This is followed by descriptions of examples of representative working and projected magnets, as well as essential auxiliary equipment. An overview is provided of ongoing development that may impact on the design of future magnets.

Plasma edge rotations and magnetohydrodynamic behaviors have been studied during radial electric field variations in IR-T1 tokamak. An external positive limiter bias has been used as an external radial electric field. The profiles of radial electric field, floating potential, poloidal and toroidal rotation velocities and MHD activities have been studied during positive limiter bias. The poloidal and toroidal velocities reduced when the bias was applied and their fluctuations on the plasma edge became smoother furthermore. A significant excitation in dominant mode (4, 1) oscillation amplitude and a sharp decrease in magnetic island rotation frequency have been seen.

A national project of the HT-7U superconducting tokamak is now under design and construction in ASIPP. Developing candidate materials for plasma facing components (PFC) of the HT-7U device is a crucial importance due to its operation circumstance under high heat flux and steady state. A series of doped CFC, multi-element doped graphite, thick SiC gradient coatings and tungsten coatings on carbon based materials (CBM) were developed by cooperation in China, evaluation experiments have been systematically performed in China and Japan by cooperation of Core University Program. The behavior of doped graphite and thick SiC gradient coatings on doped graphite under steady state and pulsed high heat flux, and HT-7 limiter plasma irradiation was investigated. Tungsten coating on CBM is now also under development being considered as one of the promising candidate plasma facing materials (PFM) for divertor plates of HT-7U device. The primary results about candidate PFM of HT-7U device are presented in this paper.

In this paper a data driven methodology to automatically derive an interpretable Fuzzy Logic Classifier (FLC) has been applied to the problem of confinement regime identification in the Joint European Torus. The approach has been developed explicitly to handle the complexities of the inference process in Magnetic Confinement Nuclear Fusion (MCNF). The resulting FLC on the one hand attains very good performance in terms of generalization and classification, on the other hand provides a series of rules which can be easily interpreted and contributing to a very good first, intuitive understanding of the physics involved.

The Fusion Advanced Studies Torus (FAST) conceptual has been proposed as possible European ITER Satellite facility with the aim of preparing ITER operation scenarios and helping DEMO design and R&D. Insights into ITER regimes of operation in Deuterium plasmas can be obtained from investigations of nonlinear dynamics that are relevant for the understanding of alpha particle behaviours in burning plasmas by using fast ions accelerated by heating and current drive systems. In this paper the plasma scenarios that can be studied on FAST are presented, with emphasis on the aspect of its flexibility in terms of both performance and physics that can be investigated. Plasma position and shape control studies are also discussed.

The basic elements of an integrated model-based control strategy for extrapolating present-day advanced tokamak scenarios to steady state operation are described. Taking advantage of the large ratio between the time scales involved in the magnetic and thermal diffusion processes, the model identification procedure makes use of a multiple time scale approximation. The methodology is generic and can be applied to any device, with different sets of heating and current drive actuators, controlled variables and/or parameter profiles. It has been applied to experimental data from JET and JT-60U, and satisfactory models have been obtained. First closed loop experiments were performed on JET with three H&CD actuators to control the safety factor profile. Simultaneous real-time control of the q-profile and toroidal velocity profile on JT-60U has been simulated.

In order to design the control system for plasma current, shape and position the structural parametric optimization of transient processes is suggested. Optimization approach to plasma dynamic is based on the consideration of transient processes of the full-sized control object that is closed by a regulator of a decreased dimension. It is suggested to use an integral performance criterion as a functional that allows optimizing the transient process, perturbed at the initial point set and the set of external disturbances. In the framework of this approach the optimization of plasma dynamics of the ITER tokamak is given.

Mathematical model and optimization at the initial stage of discharge in the ITER tokamak. Results and computer realization.

The ECRH system at ASDEX Upgrade is currently extended from 1.6 MW to 5 MW. The extension so far consists of 2-frequency units, which use single diamond-disk vacuum-windows to transmit power at the natural resonances of these disks (105 & 140 GHz). For the last unit of this extension two additional intermediate non-resonant frequencies are foreseen, requiring new window concepts. For the torus a polarisation-independent double-disk window has been developed. For the gyrotron a grooved diamond disk is actually favoured, for which the grooved surfaces act as anti-reflective coating. Since ASDEX Upgrade operates with completely W-covered plasma facing components, central ECRH is often applied to suppresses W-accumulation in the plasma center. In order to extend the operational range for central ECRH, X3- and O2-heating schemes were developed. Both are characterized by incomplete single-path absorption. For X3 heating, the X2 resonance at the pedestal on the high field side is used as a 'beam-dump', for the O2 scheme a specific reflector tile on the inner heat shield enforces a second path through the plasma center. The geometry for NTM control had to be modified to allow simultaneous central heating. In real-time the ECRH position can be determined either by ray-tracing based on real-time equilibria and density profiles or from ECE for modulated ECRH power. Fast real-time ECE also allows to determine the NTM position. Further major physics applications of the system are summarized.

The electrostatic electron Bernstein wave (EBW) can provide localised on- and off-axis heating and current drive in typically overdense (high-β) spherical tori (ST) where the usual electromagnetic EC modes are cut-off. Hence, the EBW is a candidate for plasma control and stabilisation in such devices. We present here a modelling of EBW heating and current drive in realistic ST conditions, particularly in typical NSTX equilibria and in model equilibria for NHTX [1] and MAST Upgrade [2,3]. The EBW injection parameters are varied in order to find optimized scenarios and possible ways to control the deposition location and the driven current. It is shown that EBWs can be deposited and efficiently drive current at any radial location.

The application of superconductivity to the large magnets required for charged particle spectroscopy in high energy physics experiments, and for plasma containment in fusion experiments, has resulted in a spectacular leap in the efficiency of these devices. First applied in the late 1960s, the technology has progressed to meet increasingly demanding goals of the experiments and has stimulated important development of the associated conductors. In this article we describe briefly the basic requirements that determine the design of the different types of magnets. This is followed by descriptions of examples of representative working and projected magnets, as well as essential auxiliary equipment. An overview is provided of ongoing development that may impact on the design of future magnets.