Vertical Field Coils Research Articles

Performance of MT-I spherical tokamak with upgraded power supplies

This paper describes the upgradation of power supply systems for the improved performance of a small spherical tokamak (MT-I) with an aspect ratio of 1.67 in operation at the Pakistan Tokamak Plasma Research Institute (PTPRI), Islamabad. Three mutually independent power supplies are designed to generate current pulses of desired shape and magnitude in central solenoid (CS), toroidal field (TF) and vertical field (VF) coil systems. The main parts of a power supply are capacitor banks as energy storage devices, ignitrons and thyristors as fast switches for controlled power switching. A double capacitor bank topology with enhanced second capacitor bank voltage is found to be suitable for CS to improve the plasma startup conditions. Simulations are performed in MATLAB Simulink for confirmation of the designed power supply system. The Rogowski coil and current transformers are employed to measure plasma current and the coil’s current waveforms during the experiment in the presence of hydrogen as working gas. An ocean spectrometer and a photron high-speed camera (SA-8) are used for optical measurements and fast plasma imaging. An improvement in startup conditions is observed, enhancing the plasma current duration and light emission intensity recorded in the high-speed camera masked with a Hα filter.

Existence of an optimized stellarator with simple coils

An optimized compact stellarator with four simple coils is obtained from direct optimization via a coil shape. The new stellarator consists of two interlocking coils and two vertical field coils similar to those of the Columbia Non-neutral Torus (CNT) (Pedersen et al., Phys. Rev. Lett., vol. 88, 2002, pp. 205002). The optimized configuration has a global magnetic well and a low helical ripple level comparable to that of Wendelstein 7-X (W7-X) (Wolf et al., Nucl. Fusion, vol. 57, 2017, pp. 102020). The two interlocking coils have a smooth three-dimensional shape much simpler than those of advanced stellarators such as W7-X. This result opens up possibilities of future stellarator reactors with simplified coils.

New horizontal and vertical field coils with optimised location for robust decentralized plasma position control in the IGNITOR tokamak

The paper deals with the improvement of the poloidal system of the IGNITOR tokamak, and the design and simulation of the robust decentralised control systems of the plasma’s vertical and horizontal position. The results were obtained on a linear model that is constructed from the circuit equations for the coils, plasma, and vacuum vessel (VV) currents, and from the force balance equations for the plasma. Two new coils that are located close to the VV have been introduced in the IGNITOR poloidal system; specifically, a Horizontal Field Coil (HFC) that is designated for the plasma vertical position control and a Vertical Field Coil (VFC) that is used for the horizontal position control. For the first time in the design of D-shaped tokamaks poloidal systems, the location of the HFC was optimised to achieve the maximum size of the plasma controllability region in the vertical direction and the location of the VFC was determined from the optimisation of the plasma position response to the voltage step function. Although the ideal position of the VFC with fast response time is on the equatorial plane, its location usually is occupied by the other devices such as ports, diagnostic system, and heating systems, so the separated two-section VFC is used. Careful calculation shows the performance is degraded by only 20 % in this layout. For the improved IGNITOR poloidal system, the new robust decentralised plasma position control systems were designed by means of two robust approaches: H∞-optimisation and Linear Matrix Inequalities (LMI). The results of the simulation of the control systems with two types of actuator models as the multiphase thyristor rectifier and the voltage inverter in Pulse- Width Modulation (PWM) mode for the new IGNITOR poloidal system are presented.

Plasma diagnostic systems and methods used on the stellarator TJ-II

The TJ-II is a heliac-type stellarator device with major radius of 1.5 m and averaged minor radius ⩽0.22 m that has been operated at CIEMAT, Madrid since 1998. Its full magnetic field is created by a system of poloidal, central, toroidal and vertical field coils, thus it possesses a fully 3-dimensional plasma structure and a bean-shaped plasma cross-section. Although this results in a complicated vacuum-vessel layout, it has excellent port access for diagnostics (96 portholes). During its initial operational phase, it was equipped with a limited set of essential diagnostics. Since then, a broad variety and large number of both passive and active diagnostics have been installed. The former include Hα monitors, light spectrometers, an electron cyclotron emission radiometer, X-ray filter monitors, neutral particle analysers, magnetic diagnostics, as well as cameras, among others, while the latter include various laser, atomic and ion beam based diagnostics, microwave probe beams, Langmuir probes plus impurity injection techniques. In this paper, after describing the TJ-II stellarator, its heating and fuelling systems, the diagnostic systems employed are outlined and discussed briefly here. Finally, results obtained with selected systems are highlighted.

Calculation of Coil Currents for Equilibria with Reversed Plasma Current Density in J-Text Tokamak

To design the alternating current operation of the J-text tokamak, this paper studies the plasma equilibrium during the plasma toroidal current reversing. First, four alternative current reversal equilibrium configurations with circular cross section when the total toroidal current becomes zero were constructed. Then, the profile of the plasma current was used as the input of a fixed boundary equilibrium solver to calculate the poloidal field coil currents. According to the results, the relationship between the plasma current distribution and currents in ohmic heating coils and vertical field coils is discussed. Finally, the plasma equlibrium of a toroidal current reversal with total plasma current from 40 kA to –40 kA in 40 ms was designed.

Intrinsic ELMing in ASDEX Upgrade and global control system-plasma self-entrainment

It is well established that edge localized modes can be entrained to the frequency of applied global magnetic perturbations. These perturbations are delivered to the plasma using the vertical control system field coil currents. These field coils are part of an active control system that is required to maintain the plasma in a steady state. We perform time domain timeseries analysis of natural ELMing when there are no applied perturbations in the ASDEX Upgrade tokamak. We find that the plasma can transition into a state in which the control system field coil currents continually oscillate and are synchronized with oscillations in characteristic plasma parameters such as plasma edge position and total MHD energy. These synchronous oscillations have a one-to-one correlation with the naturally occurring ELMs; the ELMs all occur when the control system coil current is around a specific temporal phase. Large and small ELMs may be distinguished by the amplitude of inward movement of the edge following an ELM. Large ELMs are then found to occur preferentially around a specific temporal phase of the vertical position control coil current. Small ELMs are most likely in antiphase to this. The large and small natural ELMs occur at the opposite extrema of the oscillations in the control system vertical position control coil current. The control system coil current phase may thus provide a useful parameter to order the observed ELM dynamics. We have identified a class of natural ELMing which is a self-entrained state, in which there is a continual non-linear feedback between the global plasma dynamics and the active control system that is intrinsic to the cyclic dynamics of naturally occurring ELMs. Control system-plasma feedback thus becomes an essential component for integration into future models of natural ELM dynamics.

Development of the saddle loop sensors on the J-TEXT tokamak

To measure the amplitude and phase of the non-axisymmetric radial magnetic field generated by the locked mode, 12 saddle loop sensors are newly developed on the J-TEXT Tokamak. The saddle loop is made of flexible printed circuit (FPC) to adapt the complex installment environment and ensure the installment accuracy. In the experiment, the saddle loop measures the radial magnetic field of locked mode and the axisymmetric equilibrium magnetic fields as well as that of the corresponding eddy current. Precise compensation for the fluxes induced by the horizontal and vertical field coils is realized by utilizing the lumped eddy current circuits based on an analytical model. By using this set of saddle loop sensors, the amplitude and phase of the m/n = 2/1 locked mode are clearly measured for the case of error field locking with slow rotation and the penetration of resonant magnetic perturbations (RMPs) respectively. Here, m and n are the poloidal and toroidal mode number.

RETRACTED: Semi-emprical techniques for the specification of Shafranov shift in the large aspect ratio tokamaks

Equilibrium reconstruction consists in identifying, from experimental measurements, a distribution of the plasma current density that satisfies the pressure balance constraint. Numerous methods exist to solve the Grad–Shafranov equation, describing the equilibrium of plasma confined by an axisymmetric magnetic field. In this research, we have reviewed six techniques for determination of the plasma position in IR-T1 tokamak and the effects of superconducting coils and chamber: an optical, flux loops, magnetic probes, multipole moments techniques, and two methods based on the vertical field coil characteristics and analytical solution of Grad–Shafranov equation. In the first method an image processing technique used for the output signal of CCD camera and plasma emission intensity profile and then plasma position obtained. To compare the plasma position obtained using this method, flux loop method, magnetic probe method, multipole moments method, analytical solution of Grad–Shafranov equation, and a sensor-less method based on vertical field coil characteristics are also experimented on the IR-T1 tokamak. Results were compared and discussed.

Automation of Aditya tokamak plasma position control DC power supply

Plasma position control is essential for obtaining repeatable high temperature, high-density discharges of longer duration in tokamaks. Recently, a set of external coils is installed in the vertical field mode configuration to control the radial plasma position in ADITYA tokamak. The existing capacitor bank cannot provide the required current pulse beyond 200ms for position control. This motivated to have a DC power supply of 500A to provide current pulse beyond 200ms for the position control. The automatization of the DC power supply mandated interfaces with the plasma control system, Aditya Pulse Power supply, and Data acquisition system for coordinated discharge operation. A high current thyristor circuit and a timer circuit have been developed for controlling the power supply automatically for charging vertical field coils of Aditya tokamak. Key protection interlocks implemented in the development ensure machine and occupational safety. Fiber-optic trans-receiver isolates the power supply with other subsystems, while analog channel is optically isolated. Commissioning and testing established proper synchronization of the power supply with tokamak operation. The paper discusses the automation of the DC power supply with main circuit components, timing control, and testing results.

Radial Compression to Transiently Achieve Higher Beta Poloidal in a Tokamak

An increase in the magnetic strength of the vertical field will, in general, compress and heat plasmas, which thus allows high temperature, high density, and high beta to be reached. In this paper, according to the scaling laws for the magnetic compression, the scaling range of both the major-radius compression and the minor-radius compression is figured out. Then, the comparison between the major-radius compression and the minor-radius compression is carried out. It demonstrates that the minor-radius compression can obtain higher ${\beta }_{p}$ of plasma. Consequently, we try to significantly enhance discharge parameters by the minor-radius compression by inserting the sets of the vertical-field coils in the experimental advanced superconducting tokamak (EAST) vacuum chamber. Afterward, the minor-radius compression based on EAST shot #34128 is deduced with ${C}_{a}=1.5$ and ${C}_{R}=0.925$ , which increases ${\beta }_{p}$ by 76.19%. It notes that the minor-radius compression is a promising way to effectively improve ${\beta }_{p}$ in a tokamak. Finally, the preliminary design of the minor-radius compression, which includes the control system, the in-vessel vertical-field coils, and its fast-control power supply, is described.

RETRACTED ARTICLE: Magnetic Studies of Tokamak Plasma Equilibrium Based on Magnetic System Materials and Characteristics

In this contribution, we have studied the magnetic system of ITER, and plasma equilibrium using the semi-empirical technique. Tokamak magnet systems consist of four main sub-systems: Toroidal field coils, Central solenoid coils, Poloidal field coils, and Correction coils. The plasma horizontal position is calculated from the vertical field coil characteristics. The calculation is made focusing on the vertical field coil current and voltage changes due to a horizontal displacement of plasma column. The results are compared and discussed.

HIDRA: Hybrid Illinois Device for Research and Applications

The Hybrid Illinois Device for Research and Applications (HIDRA) is a medium sized classical stellarator using a l = 2, m = 5 configuration with a major radius R = 0.72 m and minor radius a = 0.19 m. HIDRA will initially be operated with 26 kW of magnetron heating (2.45 GHz) and will operate with a magnetic fields B0 = 0.087 T to 0.5 T. Electron temperatures up to Te = 20 eV and densities up to ne = 1×1018 m-3 are expected with Bernstein wave heating (OXB). HIDRA has a flexible magnetic configuration due to the addition of vertical field coils. HIDRA will be used mainly in the development of new dedicated plasma material interaction experiments in a fusion type environment. Development of multi-scale and multi phase materials adaptive to extreme environment will be a focus of HIDRA and UIUC’s expertise with in-situ diagnostics of materials will open up new opportunities for innovative material testing. HIDRA will also serve as an education and training the next generation of plasma and fusion scientists and engineers. Basic plasma physics with an emphasis on plasma material interactions will be a focus of HIDRA using established diagnostic techniques as well as the development of new diagnostics for understanding the basic plasma physics and plasma material interactions.

Measurement of eddy-current distribution in the vacuum vessel of the Sino-UNIted Spherical Tokamak.

Eddy currents have an important effect on tokamak plasma equilibrium and control of magneto hydrodynamic activity. The vacuum vessel of the Sino-UNIted Spherical Tokamak is separated into two hemispherical sections by a toroidal insulating barrier. Consequently, the characteristics of eddy currents are more complex than those found in a standard tokamak. Thus, it is necessary to measure and analyze the eddy-current distribution. In this study, we propose an experimental method for measuring the eddy-current distribution in a vacuum vessel. By placing a flexible printed circuit board with magnetic probes onto the external surface of the vacuum vessel to measure the magnetic field parallel to the surface and then subtracting the magnetic field generated by the vertical-field coils, the magnetic field due to the eddy current can be obtained, and its distribution can be determined. We successfully applied this method to the Sino-UNIted Spherical Tokamak, and thus, we obtained the eddy-current distribution despite the presence of the magnetic field generated by the external coils.

A Simplified Method Based on Magnetic Measurement for Determination of Iron-Core Tokamak Plasma Current Center

For the safe and efficient discharge operation, it is essential to real-time determine the plasma current center and guide the parameters modification in the vertical field coil current feedback control system. In this paper, we analyze the limitation of the method of measuring the plasma displacement by using single-turn loops, which was used in the vertical field coil current feedback control system on iron-core tokamak HT-7. In addition, we introduce a new method to estimate the plasma current center using the external magnetic measurements on iron-core tokamak. This method has been validated by several HT-7 discharges. The simulation and discussion of measured values by this method is presented.

Plasma current start-up experiments without a central solenoid in the iron core STOR-M tokamak

Reproducible plasma current start-up without a central solenoid (CS) has been demonstrated using the outer ohmic heating (OH) coils in the iron core STOR-M tokamak (Mitarai et al 2014 Fusion Eng. Des. 89 2467–71). Although the outer OH coil current saturates the iron core eventually, it has been demonstrated that the plasma current can be maintained during the iron core saturation phase. In this work, further studies have been conducted to investigate the effects of the turn number of the outer OH coils (N = 4 or N = 6) in the CS-less discharges and to evaluate the plasma stability with respect to the n-decay index of the vertical magnetic field. For the loose coupling of the iron core with N = 4 turns, the plasma current can be sustained after the additional third capacitor bank is applied near the iron core saturation phase, showing the slow transition from the unsaturated to the partially saturated phase. For the case of stronger coupling of N = 6 turns, the plasma current is increased at the same fast bank voltage, but the main discharge is shortened from 35 to 20 ms. As the magnetizing current is smaller due to stronger coupling between the OH coils and the plasma current, the transition from the unsaturated to the saturated phase is slightly difficult at present. The present experimental results suggest a feasible operation scenario in a future spherical tokamak (ST) at least using loose iron core coupling for smoother transition from the unsaturated to the saturated iron core phase. Thus, a reliable plasma current start-up by the outer OH coils and the current ramp-up to a steady state by additional heating power and vertical field coils could be considered as an operation scenario for future ST reactors with an iron core transformer.

Novel divertor design to mitigate neutron irradiation in the helical reactor FFHR-d1

The heat flux at the divertor in a fusion reactor is considered to have a peak of >10MW/m2. In a design study of the helical reactor FFHR-d1, the feasibility of employing a copper alloy for divertor cooling pipes was investigated; however, radiation in the divertor area would quickly damage the copper alloy. The neutron load on the divertor can be reduced by a blanket arrangement; nevertheless, in the present divertor structure, irradiation damage of materials on the inboard side of the torus remains relatively high. If the divertor could be moved to an area receiving much less radiation, then the lifetimes of divertor materials should increase. In this paper, a novel divertor structure is introduced in which the coil-support structure is modified to create a region receiving relatively low amounts of radiation without changing the geometry of the helical or vertical field coils. Using this proposed design would increase the lifetime of the copper alloy in divertor components to more than an estimated six years. In addition, the divertor could be accessed from either the upper or lower sides of the device, simplifying maintenance.

Design and implementation of data acquisition system for magnets of SST-1

The magnet system of the Steady-State Superconducting Tokamak-1 at the Institute for Plasma Research, Gandhinagar, India, consists of sixteen toroidal field and nine poloidal field. Superconducting coils together with a pair of resistive PF coils, an air core ohmic transformer and a pair of vertical field coils. These magnets are instrumented with various cryogenic compatible sensors and voltage taps for its monitoring, operation, protection, and control during different machine operational scenarios like cryogenic cool down, current charging cycles including ramp up, flat top, plasma breakdown, dumping/ramp down and warm up. The data acquisition system for these magnet instrumentation have stringent requirement regarding operational flexibility, reliability for continuous long term operation and data visualization during operations. A VME hardware based data acquisition system with ethernet based remote system architecture is implemented for data acquisition and control of the complete magnet operation. Software application is developed in three parts namely an embedded VME target, a network server and a remote client applications. A target board application implemented with real time operating system takes care of hardware configuration and raw data transmission to server application. A java server application manages several activities mainly multiple client communication over ethernet, database interface and data storage. A java based platform independent desktop client application is developed for online and offline data visualization, remote hard ware configuration and many other user interface tasks. The application has two modes of operation to cater to different needs of cool-down and charging operations. This paper describes application architecture, installation and commissioning and operational experience from the recent campaigns of SST-1.

Study of Plasma Startup on HL-2A Tokamak

For pure Ohmic drive both plasma breakdown and startup on HL-2A are on the low field side where the field null configuration is located, and the intensity of Dα emission confirms the plasma formation area. The current of vertical field coils is an important factor to facilitate plasma breakdown and a certain value is required for prompt plasma formation. The effect of the prefill pressure on plasma breakdown has been studied by scanning the prefill pressure. High and low prefill pressure will cause failure of the discharge and only in an appropriate range of the pressure can obtain a reproducible prompt discharge. A new plasma startup scenario has been introduced for bad wall conditions using early formation of a divertor configuration, the result demonstrates that it can be successful startup even if the plasma facing components have absorbed a lot of impurities.

Application of multipole moments and magnetic techniques for determination of tokamak plasma shift

AbstractIn this contribution, we presented approaches for the determination of tokamak plasma column shift based on multipole moments and magnetic techniques. First, we presented analytical details for using this technique. Then, the principle of different models based on this technique for the design and fabrication of six coils will be presented: four modified Rogowski coils (two cosine coils and two sine coils) and two saddle coils (saddle sine coil and saddle cosine coil). Also, to compare the results, the flux loops technique is used. Because of continuous measurements of magnetic field distribution around the tokamak plasma using multipole coils, this technique gives us more reliable information about the plasma current displacement. Moreover, we deduced the plasma current and plasma boundary centers shift using the two semi-empirical techniques in the IR-T1 tokamak. First, the plasma current center is calculated from the vertical field coil characteristics. The calculation is made focusing on the vertical field coil current and voltage changes due to a horizontal displacement of the plasma column. Also, the plasma boundary center shift was measured from the external magnetic coils. The results from these two techniques are compared and discussed.

Effects of Hot Limiter Biasing on Tokamak Runaway Discharges

In this research hot limiter biasing effects on the Runaway discharges were investigated. First wall of the tokamak reactors can affects serious damage due to the high energy runaway electrons during a major disruption and therefore its life time can be reduced. Therefore, it is important to find methods to decrease runaway electron generation and their energy. Tokamak limiter biasing is one of the methods for controlling the radial electric field and can induce a transition to an improved confinement state. In this article generation of runaway electrons and the energy they can obtain will be investigated theoretically. Moreover, in order to apply radial biasing an emissive limiter biasing is utilized. The biased limiter can apply +380 V in the status of cold and hot to the plasma and result in the increase of negative bias current in hot status. In fact, in this experiment we try to decrease the generation of runaway electrons and their energy by using emissive limiter biasing inserted on the IR-T1 tokamak. The mean energy of these electrons was obtained by spectroscopy of hard X-ray. Also, the plasma current center shift was measured from the vertical field coil characteristics in presence of limiter biasing. The calculation is made focusing on the vertical field coil current and voltage changes due to a horizontal displacement of plasma column.