Tokamak Start-up Research Articles

An enhanced tokamak startup model

The startup of tokamaks has been examined in the past in varying degree of detail. This phase typically involves the burnthrough of impurities and the subsequent rampup of plasma current. A zero-dimensional (0D) model is most widely used where the time evolution of volume averaged quantities determines the detailed balance between the input and loss of particle and power. But, being a 0D setup, these studies do not take into consideration the co-evolution of plasma size and shape, and instead assume an unchanging minor and major radius. However, it is known that the plasma position and its minor radius can change appreciably as the plasma evolves in time to fill in the entire available volume. In this paper, an enhanced model for the tokamak startup is introduced, which for the first time takes into account the evolution of plasma geometry during this brief but highly dynamic period by including realistic one-dimensional (1D) effects within the broad 0D framework. In addition the effect of runaway electrons (REs) has also been incorporated. The paper demonstrates that the inclusion of plasma cross section evolution in conjunction with REs plays an important role in the formation and development of tokamak startup. The model is benchmarked against experimental results from ADITYA tokamak.

Separated Double-Current Layers in a High-Guide-Field Reconnection Experiment

Magnetic reconnection in the presence of a high-guide-field is utilized to heat the plasma in the merging start-up of a spherical tokamak configuration. The reconnection current layer between two spherical tokamaks was gradually compressed, and a quasi-steady current-sheet thickness was obtained. During the compression phase, the current layer split transiently into two separated layers, which provided a flattened magnetic-field-region near the X-point. This modification may affect the electron-energization mechanism in the merging start-up of a spherical tokamak in a high-guide-field.

Instrumentation and control system architecture of ECRH SST1

The Electron Cyclotron Resonance Heating (ECRH) system is an important heating system for the reliable start-up of tokamak. The 42GHz and 82.6GHz Gyrotron based ECRH systems are used in tokomaks SST-1 and Aditya to carry out ECRH related experiments. The Gyrotrons are high power microwave tubes used as a source for ECRH systems. The Gyrotrons need to be handled with optimum care right from the installation to its Full parameter control operation. The Gyrotrons are associated with the subsystems like: High voltage power supplies (Beam voltage and anode voltage), dedicated crowbar system, magnet, filament and ion pump power supplies and cooling system. The other subsystems are transmission line, launcher and dummy load. A dedicated VME based data acquisition & control (DAC) system is developed to operate and control the Gyrotron and its associated sub system. For the safe operation of Gyrotron, two level interlocks with fail-safe logic are developed. Slow signals that are operated in scale of millisecond range are programmed through software and hardware interlock in scale of microsecond range are designed and developed indigenously. Water-cooling and the associated interlock are monitored and control by data logger with independent human machine interface.

Numerical modeling of tokamak breakdown phase driven by pure Ohmic heating under ideal conditions

We have simulated tokamak breakdown phase driven by pure Ohmic heating with implicit particle in cell/Monte Carlo collision (PIC/MCC) method. We have found two modes can be differentiated. When performing breakdown at low initial gas pressure, we find that it works at lower density and current, but higher temperature, and requires lower heating power, compared to when having a high initial pressure. Further, two stages can be distinguished during the avalanche process. One is the fast avalanche stage, in which the plasma is heated by induced toroidal electric field. The other is the slow avalanche stage, which begins when the plasma density reaches 1015 m−3. It has been shown that ions are mainly heated by ambipolar field and become stochastic in the velocity distribution. However, when the induced electric field is low, there exists a transition phase between the two stages. Our model simulates the breakdown and early hydrogen burn-through under ideal conditions during tokamak start-up. It adopted fewer assumptions, and can give an idealized range of operative parameters for Ohmic start-up. Qualitatively, the results agree well with certain experimental observations.

Operation and control of high power Gyrotrons for ECRH systems in SST-1 and Aditya

The Electron Cyclotron Resonance Heating (ECRH) system is an important heating system for the reliable start-up of tokamak. The 42GHz and 82.6GHz ECRH systems are used in tokamaks SST-1 and Aditya to carry out ECRH related experiments. The Gyrotrons are high power microwave tubes used as a source for ECRH systems. The Gyrotron is a delicate microwave tube, which deliver megawatt level power at very high voltage ∼40–50kV with the current requirement ∼10A–50A. The Gyrotrons are associated with the subsystems like: High voltage power supplies (Beam voltage and anode voltage), dedicated crowbar system, magnet, filament and ion pump power supplies, cooling, interlocks and a dedicated data acquisition & control (DAC) system. There are two levels of interlocks used for the protection of Gyrotron: fast interlocks (arcing, beam over current, dI/dt, anode voltage and anode over current etc.) operate within 10μs and slow interlocks (cooling, filament, silence of Gyrotron, ion pump and magnet currents) operate within 100ms.Two Gyrotrons (42GHz/500kW/500ms and 82.6GHz/200kW/1000s) have been commissioned on dummy load for full parameters. The 42GHz ECRH system has been integrated with SST-1 & Aditya tokamak and various experiments have been carried out related to ECRH assisted breakdown and start-up of tokamak at fundamental and second harmonic. These Gyrotrons are operated with VME based data acquisition and control (DAC) system. The DAC system is capable to acquire 64 digital and 32 analog signals. The system is used to monitor & acquire the data and also used for slow interlocks for the protection of Gyrotron. The data acquired from the system are stored online on VME system and after the shot stored in a file in binary format. The MDSPlus, a set of software tools has been used for data visualization & management for after shot analysis.

Impedance of an intense plasma-cathode electron source for tokamak startup

An impedance model is formulated and tested for the ∼1 kV, 1 kA/cm2, arc-plasma cathode electron source used for local helicity injection tokamak startup. A double layer sheath is established between the high-density arc plasma (narc≈1021 m−3) within the electron source, and the less dense external tokamak edge plasma (nedge≈1018 m−3) into which current is injected at the applied injector voltage, Vinj. Experiments on the Pegasus spherical tokamak show that the injected current, Iinj, increases with Vinj according to the standard double layer scaling Iinj∼Vinj3/2 at low current and transitions to Iinj∼Vinj1/2 at high currents. In this high current regime, sheath expansion and/or space charge neutralization impose limits on the beam density nb∼Iinj/Vinj1/2. For low tokamak edge density nedge and high Iinj, the inferred beam density nb is consistent with the requirement nb≤nedge imposed by space-charge neutralization of the beam in the tokamak edge plasma. At sufficient edge density, nb∼narc is observed, consistent with a limit to nb imposed by expansion of the double layer sheath. These results suggest that narc is a viable control actuator for the source impedance.

Probing helium nano-bubble formation in tungsten with grazing incidence small angle x-ray scattering

Helium nano-bubble formation in plasma facing materials has emerged as a major concern for the next-step fusion experiment ITER, where helium plasmas will be used during the tokamak's start-up phase. Here, we demonstrate that grazing incidence small-angle x-ray scattering is a powerful technique for the analysis of helium nano-bubble formation in tungsten. We measured helium bubbles with sizes between 1.5–2.5 nm in tungsten exposed to helium plasma at 700 °C, where a smaller number of larger bubbles were also observed. Depth distributions can be estimated by taking successive measurements across a range of x-ray incidence angles. Compared with traditional approaches in the field, such as transmission electron microscopy, this technique provides information across a much larger volume with high statistical precision, whilst also being non-destructive.

Start-up phase plasma discharge design of a tokamak via control parameterization method**Project supported by the National Natural Science Foundation of China (Grant Nos. 61104048 and 61473253) and the National High Technology Research and Development Program of China (Grant No. 2012AA041701).

The tokamak start-up is a very important phase during the process to obtain a suitable equalizing plasma, and its governing model can be described as a set of nonlinear ordinary differential equations (ODEs). In this paper, we first estimate the parameters in the original model and set up an accurate model to express how the variables change during the start-up phase, especially how the plasma current changes with respect to time and the loop voltage. Then, we apply the control parameterization method to obtain an approximate optimal parameters selection problem for the loop voltage design to achieve a desired plasma current target. Computational optimal control techniques such as the variational method and the costate method are employed to solve the problem, respectively. Finally, numerical simulations are performed and the results obtained via different methods are compared. Our numerical parameterization method and optimization procedure turn out to be effective.

Influence of plasma surface interactions on tokamak startup

The startup phase of a tokamak is a complex phenomenon involving burnthrough of the low-Z impurities and rampup of Ip, the plasma current. The design considerations of a tokamak are closely connected with the startup modeling. Plasma evolution is analysed using a zero-dimensional model. The particle and energy balance is considered of two subclasses of plasmas which are penetrable by neutral gas, together with another component, neutrals trapped in the wall. The first subclass includes plasmas being penetrated by slow neutrals of (∼few eV) temperature. The second includes plasmas being penetrated only by fast neutrals having a temperature comparable to that of the ions. The impact of impurities on energy balance is considered through their generation by ion induced desorption of adsorbed oxygen on the first wall and physical and chemical sputtering of carbon. The paper demonstrates self-consistently that the evolution of initial phase of the discharge is intimately linked to the condition of the plasma facing components (PFCs) and the resultant plasma surface interactions.

The physics of tokamak start-up

Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases, inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. International Thermonuclear Experimental Reactor, the National Spherical Torus Experiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in the solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection, and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current.

Development of a Quasi-Steady Equilibrium Field System for Plasma Merging ST Startup Experiments on the UTST Device

A novel equilibrium field (EF) system has been developed for use in the merging start-up of spherical tokamak (ST) plasma in the University of Tokyo Spherical Tokamak (UTST) device. Since the UTST device utilizes external poloidal field (PF) coils to form two STs on the top and bottom of the device by huge induction voltages, it is necessary to decouple the slow EF coil system from the fast-swing PF coils. A new EF system using electric double-layer capacitors was constructed and evaluated through the merging start-up experiment in the UTST. Using a very long duration of the EF current waveform and a thick magnetic shield successfully reduced the induction voltages from the PF coils and prolonged the equilibrium time constant.

Control Issues Related to Start-Up of Tokamaks

Developing robust and reproducible start-up scenarios is essential for all tokamaks and especially for burning plasma devices. A tokamak start-up sequence is complex and calls on control of different types of plasmas, from nearly collisionless low-temperature electrons in a large neutral background to a more conventional diverted high-temperature fully ionized plasma during the ramp-up phase.

Plasma Startup Design of Fully Superconducting Tokamaks EAST and KSTAR With Implications for ITER

Recent commissioning of two major fully superconducting (SC)-shaped tokamaks, Experimental Advanced Superconducting Tokamak (EAST) and Korean Superconducting Tokamak Advanced Research (KSTAR), represents a significant advance in magnetic fusion research. The key to commissioning success in these complex and unique tokamaks was as follows: 1) use of a robust, flexible plasma control system (PCS) based on the validated DIII-D design; 2) use of the TokSys design and modeling environment, which is tightly coupled with the DIII-D PCS architecture for first-plasma scenario development and plasma diagnosis; and 3) collaborations with experienced internationally recognized teams of tokamak operations and control experts. We provide an overview of the generic modeling environment and plasma control tools developed and validated within the DIII-D experimental program and applied through an international collaborative program to successfully address the unique constraints associated with the startup of these next-generation tokamaks. The unique characteristics of each tokamak and the machine constraints that must be included in device modeling and simulation, such as SC coil current slew rate limits and the presence of nonlinear magnetic materials, are discussed, along with commissioning and initial operational results. Lessons learned from the startup experience in these devices are summarized, with special emphasis on ramifications for International Thermonuclear Experimental Reactor (ITER).

Tokamak Start-Up Modeling and Design for EAST First Plasma Campaign

The Experimental Advanced Superconducting Tokamak (EAST) was the first shaped tokamak of mega-ampere scale to achieve plasma utilizing a fully superconducting poloidal field coil system, and it is addressing ITER relevant superconducting constraints associated with the breakdown, plasma formation, and initial plasma current ramp. Electric field production for plasma start-up is severely limited in fully superconducting machines as a consequence of constraints associated with coil and lead voltages and eddy current heating in the superconducting coils. Such constraints motivate the use of electromagnetic modeling codes to design start-up scenarios for these devices. The successful first plasma campaign of the EAST superconducting tokamak was greatly facilitated by extensive and careful planning, development of appropriate modeling, simulation and diagnostic tools, a highly flexible plasma control system, and a highly experienced international collaboration team. We describe the design and modeling tools used to develop the first plasma scenario along with results of their application in the start-up campaign. Control design tools and plasma control algorithms utilized during the first campaign are discussed. Key physics, engineering, and operations results of the first plasma campaign are presented, including observations relevant to future devices such as ITER.

The Formation of a Tokamak-like Plasma in Initial Experiments Using an Outboard Plasma Gun Current Source

Solenoid-free tokamak startup via point-source DC helicity injection is demonstrated on the Pegasus Toroidal Experiment using a high current density, low impurity plasma gun mounted near the outboard midplane. A threshold in the vacuum vertical magnetic field strength that allows the injected current filament to relax into a tokamak-like topology is observed. A simple 2-D model of the vacuum magnetic field suggests this threshold is the maximum field strength that allows a toroidally connected field null to form. Discharges with I p ≈ 17 kA are produced using less than 2 kA of injected current and no inductive drive. The tokamak-like discharges exhibit current decay times about five times longer than the injected current decay, expansion of the plasma into the vacuum region and a significant increase in the line-integrated density.

Progress in the development of heating systems towards long pulse operation for KSTAR

Construction of the Korea superconducting tokamak advanced research (KSTAR) tokamak is in its final phase. For the long-pulse KSTAR discharges, the ion cyclotron range of frequencies (ICRF) and neutral beam injection (NBI) heating systems are expected to play important roles through a selective heating of ions and electrons, control of the plasma pressure and current profiles, a core fuelling and beam diagnostics for the KSTAR. In addition, the ICRF system is expected to be used for possible discharge cleaning and assisting in the tokamak startup. In this paper, the recent progress in the development of the ICRF and the NBI heating systems is described. The four-strap ICRF antenna has been successfully tested for a voltage up to 41 kV for a pulse length of 300 s (to 46 kV for 20 s) in a test chamber. A prototype KSTAR NBI system has been developed. At present, the system has successfully produced a 1 MW beam power for 200 s and a 3.5 MW output beam power for 4 s.

Second harmonic electron cyclotron pre-ionization in the DIII-D tokamak

Second harmonic 60 GHz electron cyclotron (EC) pre-ionization using low field side (LFS) X-mode launchers has been effective in producing target plasmas for startup of the DIII-D tokamak with electron densities comparable to fundamental LFS pre-ionization plasmas using the same EC system. A visible Bremsstrahlung array showed that breakdown occurred at the 2nd harmonic resonance location and after a few milliseconds the EC driven plasma filled the entire vessel, independent of the resonant location, which was varied from near the inner wall to the centre of the torus. The power threshold for ionization of ∼0.4 MW was observed in DIII-D for these 2nd harmonic pre-ionization experiments. An orbit following code calculated that cold (0.03 eV) electrons could be heated to energies above 20 eV where the ionization of the neutral deuterium gas can occur. Scaling from DIII-D to ITER indicates that electron cyclotron heating (ECH) 2nd harmonic pre-ionization and initial plasma formation is possible if ITER operation at reduced toroidal field is desired, but additional experiments are required to extrapolate the entire ECH start-up scenario from DIII-D to ITER.

Tokamak Start-up under Assistance of RF Waves

To improve the start-up behavior of tokamak discharges and realize the low loop voltage start-up is required by performance of large scale, full superconductor tokamaks. In recent years, some kinds of RF wave have been used to assist the start-up and some exciting results have been gained. This paper introduce the investigation on both in physical principle and experimental research of the start-up process, in which high frequency RF waves were used to assist it.

Study of plasma current start-up of the middle-sized tokamak with force-balanced coils

For a high field tokamak device, we have developed force-balanced coils (FBCs) which have nonuniform-pitch multi-pole helical-configuration to reduce centering electromagnetic forces without using cancellation coils. As a superconductivity magnet, the FBC configuration is favorable to the achievement of high critical current densities because of a J c‖ B effect. In the tokamak operation with the FBCs, the rise of the toroidal field synchronizes with the plasma current ramp up because the FBCs function both as toroidal field coils (TFCs) and as primary coils for ohmic heating. The torsional force exerted on the FBCs is comparable and the net centering force is reduced to 20% compared with those on conventional TFCs of the same dimensions.

Transients in D-T and D-3He Tokamak Fusion Reactors

A time-dependent, volume-averaged particle and power balance code is used to investigate reactivity transients during tokamak startup and after sudden changes in the plasma confinement, fueling rates, and impurity concentrations in deuterium-tritium (D-T) and D-3He fusion reactors. For a given H-mode factor fH relative to the ITER89-P scaling law, a very narrow range of values, limited by quenching of the fusion burn due to ash accumulation and by exceeding operational limits, is found to sustain steady fusion burn. The dependence of the large power overshoot taking place shortly after ignition due to ash accumulation on the assumed ρ and fH is examined. To alleviate the excessive external heating power requirements for D-3He-reactor startup, schemes utilizing D-T fusion reactions are considered. Because of power transients of several hundreds of megawatts in reactors operating at a gigawatt level of fusion power, triggered by very small changes in the plasma confinement (