Advanced Tokamak Research Articles

Advanced scenarios for ITER operation

In thermonuclear fusion research using magnetic confinement, the tokamak is the leading candidate for achieving the conditions required for a reactor. An international experiment, ITER, is proposed as the next essential and critical step on the path to demonstrating the scientific and technological feasibility of fusion energy. ITER is to produce and study plasmas dominated by self-heating. This would give unique opportunities to explore, in reactor relevant conditions, the physics of α-particle heating, plasma turbulence and turbulent transport, stability limits to the plasma pressure and exhaust of power and particles. Important new results obtained in experiments, theory and modelling, enable an improved understanding of the physical processes occurring in tokamak plasmas and give enhanced confidence that ITER will achieve its goals. In particular, progress has been made in research to raise the performance of tokamaks, aimed to extend the discharge pulse length towards steady-state operation (advanced scenarios). Standard tokamak discharges have a current density increasing monotonically towards the centre of the plasma. Advanced scenarios, on the other hand, use a modified current density profile. Different advanced scenarios range from (i) plasmas that sustain a central region with a flat current density profile (zero magnetic shear), capable of operating stationary at high plasma pressure, to (ii) discharges with an off-axis maximum of the current density profile (reversed magnetic shear in the core), able to form internal transport barriers, to increase the confinement of the plasma. The physics of advanced tokamak discharges is described, together with an overview of recent results from different tokamak experiments. International collaboration between experiments aims to provide a better understanding, control and optimization of these plasmas. The ability to explore advanced scenarios in ITER is very desirable, in order to verify the results obtained in experiments today and to demonstrate the potential to significantly increase the economic attractiveness of the tokamak.

Advanced Control Techniques and High Performance Discharges on DIII-D

The advancement of plasma control techniques has enabled significant progress to be made toward the scientific understanding and realization of Advanced Tokamak operation on DIII–D. The Advanced Tokamak features fully noninductive current drive, operation at high plasma pressure and high energy confinement time. These features require efficient current drive systems, simultaneous control of plasma current and pressure profiles, and active feedback control of plasma instabilities. A number of key systems on DIII-D have been developed to provide this control capability. A versatile electron cyclotron heating and current drive system is routinely providing in excess of 2 MW of power for pulse lengths from 2 to 5 s. This system has been used to provide offaxis current drive, direct electron heating and pressure profile modification, and stabilization of the Neoclassical Tearing Mode instability. A combination of control of magnetic error fields, neutral beam induced plasma rotation, and active feedback stabilization using both external and internal nonaxisymmetric coil systems has been used to stabilize the Resistive Wall Mode at high values of plasma pressure. Control of the ELM instability has recently been demonstrated using the newly installed internal coil system. The higher speed and expanded realtime diagnostic capability of our recently upgraded plasma control system permits these various control techniques to be simultaneously integrated to achieve our high performance discharges. This has resulted in fully noninductively driven plasmas with βN = 3.5 and βT = 3.6% sustained for up to 1 s. Upgrades and facility modifications to further enhance our control and scientific capabilities including rotation of a neutral beamline, expanded EC system power, and installation of a new lower divertor are discussed.

Recent Technological Progress for Advanced Tokamak Research in JT-60U and JFT-2M

For steady state advanced tokamak research with long pulse operations, JT-60U tokamak discharge, NBI and RF heating injection durations have been extended from 15 s to 65 s and from 10 s to 30 s respectively mainly by means of modifying their control systems and using derated power levels. In addition, technological issues for their long pulse injections with the heating systems have been solved as follows. The ion source of the negative ion NBI system was modified to increase gas conductance in the accelerator, which reduced the heat load to the grounded grid due to stripping loss to a level that enables operations of 2 MW for 30 s. A new method of controlling the anode voltage has been developed for sustaining the oscillation condition of a gyrotron in the electron cyclotron (EC) system. With this method, the EC injection duration has reached 16 s at 0.4 MW. To avoid serious damage of the LH launcher, a heat-resistant carbon grill LH antenna was implemented on the original stainless steel grill. To date, the advanced tokamak operations have been extended to βN = 2.1 for 20 s. In JFT-2M, high βN plasmas had been investigated with the vacuum vessel covered with ferritic steels. βN of ~3.5 was obtained with rwall/a~1.3-1.6 without serious influence of ferromagnetic walls (rwall is distance of the wall from a plasma center and a is minor radius of a plasma). This encourages the utilization of ferric steel as a structural material for future reactors.

Nonlinear Methods for Current Limit Constraint Satisfaction in Tokamak Plasma Shape Control

The DIII-D tokamak is capable of supporting a wide variety of plasma equilibria because of its relatively large number of coils and their proximity to the plasma. To support its advanced tokamak mission, the DIII–D experimental program continues to push the envelope of this capability, frequently encountering limits imposed by allowable currents in poloidal shaping coils. Violation of current constraints is presently dealt with by operator adjustment of control targets and gains between plasma discharges. At the same time, demands for more precise and stable control have motivated efforts to develop and install advanced multivariable algorithms for control of plasma shape in DIII-D and other devices. There is currently no way to ensure respect of nonlinear current constraints in a multivariable linear controller design and no practical way to manually tune these fully coupled controllers between discharges after installation. Various linear minimization schemes can be implemented to encourage currents to remain within limits, but adherence to these limits cannot be guaranteed by linear methods alone. In this paper, we describe ongoing efforts to provide methods that guarantee currents will not exceed preset limits, and that simultaneously achieve the best obtainable quality of control subject to current limit constraints.

Fire, A Test Bed for ARIES-RS/AT Advanced Physics and Plasma Technology

The overall vision for FIRE is to develop and test the fusion plasma physics and plasma technologies needed to realize capabilities of the ARIES-RS/AT power plant designs. The mission of FIRE is to attain, explore, understand and optimize a fusion dominated plasma which would be satisfied by producing DT fusion plasmas with nominal fusion gains ~10, self-driven currents of ≈ 80%, fusion power ~ 150 - 300 MW and pulse lengths up to 40 s. Achieving these goals will require the deployment of several key fusion technologies under conditions approaching those of ARIES-RS/AT. The FIRE plasma configuration with strong plasma shaping, a double null pumped divertor and all metal plasma facing components is a 40% scale model of the ARIES-RS/AT plasma configuration. “Steady-state” advanced tokamak modes in FIRE with high β, high bootstrap fraction and 100% non-inductive current drive are suitable for testing the physics of the ARIES-RS/AT operating modes. The development of techniques to handle power plant relevant exhaust power while maintaining low tritium inventory is a major objective for a burning plasma experiment. The FIRE H-modes and AT-modes result in fusion power densities from 3 - 10 MWm-3 and neutron wall loading from 2 - 4 MW m-2 which are at the levels expected from the ARIES-RS/AT design studies.

Nonrigid, Linear Plasma Response Model Based on Perturbed Equilibria for Axisymmetric Tokamak Control Design

Tokamak control design relies on an accurate linear model of the plasma response, which can often dominate the local field variations in regions under active feedback control. For example, when fluxes at selected points on the plasma boundary are regulated in DIII–D [1], the plasma response to a change in a coil current gives rise to a flux change which can be larger than and opposite to the flux change caused by the coil alone.In the past, rigid plasma models have been used for linear stability and shape control design. In a rigid model, the plasma current profile is considered fixed and moves rigidly in response to control coils to maintain radial and vertical force balance. In a nonrigid model, however, changes in the plasma shape and current profile are taken into account. Such models are expected to be important for future advanced tokamak control design. The present work describes development of a nonrigid plasma response model for high-accuracy multivariable control design and provides comparisons of model predictions against DIII-D experimental data. The linear perturbed plasma response model is calculated rapidly from an existing equilibrium solution.

Integrated Plasma Control in Next-Generation Devices Using DIII-D Modeling and Simulation Approaches

An extensive set of software tools for integrated plasma control, developed and validated on the DIII-D tokamak, has been applied to several nextgeneration fusion device designs including KSTAR, EAST, and ITER. These devices will require elements of integrated plasma control in order to achieve high reliability advanced tokamak or burning plasma operation. Plasma Control Systems (PCS) based on the DIII–D PCS [1] have been designed for each of these devices. The integrated plasma control approach uses validated physics models to design controllers for plasma shape and both axisymmetric and nonaxisymmetric MHD instabilities [2] and confirms control performance by operating actual machine control hardware and software against detailed tokamak system simulations. The physics-based models include conductors, diagnostics, power supplies, and both linear and nonlinear plasma models. These models can be implemented in the detailed control simulations to verify event handling and demonstrate functioning of control action under realistic hardware (CPU and network) conditions. Results of simulations are shown, illustrating control performance characteristics produced by each device design, engineering choices, and control system algorithms and hardware. Such simulations allow confirmation of performance prior to actual implementation on an operating device.

Advanced tokamak research at the DIII-D National Fusion Facility in support of ITER

Fusion energy research aims to develop an economically and environmentally sustainable energy system. The tokamak, a doughnut shaped plasma confined by magnetic fields generated by currents flowing in external coils and the plasma, is a leading concept. Advanced Tokamak (AT) research in the DIII-D tokamak seeks to provide a scientific basis for steady-state high performance operation. This necessitates replacing the inherently pulsed inductive method of driving plasma current. Our approach emphasizes high pressure to maximize fusion gain while maximizing the self-driven bootstrap current, along with external current profile control. This requires integrated, simultaneous control of many characteristics of the plasma with a diverse set of techniques. This has already resulted in noninductive conditions being maintained at high pressure on current relaxation timescales. A high degree of physical understanding is facilitated by a closely coupled integrated modelling effort. Simulations are used both to plan and interpret experiments, making possible continued development of the models themselves. An ultimate objective is the capability to predict behaviour in future AT experiments. Analysis of experimental results relies on use of the TRANSP code via the FusionGrid, and our use of the FusionGrid will increase as additional analysis and simulation tools are made available.

高周波による核融合プラズマ制御の進展 3.低域混成波を用いた電流分布制御と先進トカマク/定常プラズマ運転への応用

Non-inductive current drive (CD) by radio-frquency waves in the lower hybrid range of frequency (LHRF) is one of the most powerful schemes to drive a plasma current and control the current profile of a tokamak plasma externally. This paper reviews recent achievements of CD by LHRF (LHCD) in tokamak experiments with a special emphasis on steady state tokamak operation and profile control in an advanced tokamak.

ITB formation with counter ECCD and LHCD and Suprathermal ECCD experiments on FTU in ITER relevant conditions

In this paper the results of the experiments of combined injection of LH waves and EC waves performed in the FTU tokamak are reported. Such experiments were mainly devoted to study and to control the access to the advanced tokamak scenarios in plasma conditions close to ITER parameters (Bt≈5T or higher, ne,line≈1020 m−3). Two different absorption mechanisms were used for the EC waves. In the first one the cold resonance absorption of EC waves launched with a toroidal angle was used to induce small modification of the current profile mainly maintained by LH waves. In the second one the absorption through Doppler shift due to the fast electron tails generated by LHCD was used.

Advanced tokamak research in DIII-D

Advanced tokamak (AT) research in DIII-D seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and high poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles and active magnetohydrodynamic stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization by plasma rotation and active feedback with non-axisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these conditions, bootstrap supplies most of the current. Steady-state operation requires replacing the remaining inductively driven current, mostly located near the half radius, with non-inductive external sources. In DIII-D this current is provided by ECCD, and nearly stationary AT discharges have been sustained with little remaining inductive current. Fast wave current drive is being developed to control the central magnetic shear. Density control, with divertor cryopumps, of AT discharges with ELMing H-mode edges facilitates high current drive efficiency at reactor relevant collisionalities. An advanced plasma control system allows integrated control of these elements. Close coupling between modelling and experiment is key to understanding the separate elements, their complex nonlinear interactions, and their integration into self-consistent high performance scenarios. This approach has resulted in fully non-inductively driven plasmas with βN ≤ 3.5 and βT ≤ 3.6% sustained for up to 1 s, which is approximately equal to one current relaxation time. Progress in this area, and its implications for next-step devices, will be illustrated by results of these and other recent experiment and simulation efforts.

Formation criteria and positioning of internal transport barriers in ASDEX Upgrade

Advanced tokamak discharges on ASDEX Upgrade can generate an internal transport barrier (ITB) during the current ramp-up phase early in the discharge. Formation of the ITB has become more reliable with the discovery that a low density is necessary for it to form. These ITBs form in very low or negative central shear regions. There is no clear evidence of integer q magnetic surfaces triggering the ITB phase. The appearance of an integer q surface (usually q = 2) often leads to a second ITB (i.e. another steepening in the ion temperature gradient) inside the first ITB.Investigation of possible turbulence suppression mechanisms suggests that something is missing from the generally accepted model of sufficient E × B sheared flow suppressing the turbulence since the E × B shearing rate was not greater than, or comparable with, the maximum linear growth rate in about half of the ITB discharges analysed. Thus, it does not explain the reduced radial transport observed in these discharges.

Present Status of the KSTAR Superconducting Magnet System Development

The mission of Korea Superconducting Tokamak Advanced Research (KSTAR) project is to develop an advanced steady-state superconducting tokamak for establishing a scientific and technological basis for an attractive fusion reactor. Because one of the KSTAR mission is to achieve a steady-state operation, the use of superconducting coils is an obvious choice for the magnet system. The KSTAR superconducting magnet system consists of 16 Toroidal Field (TF) coils and 14 Poloidal Field (PF) coils. Internally-cooled Cable-In-Conduit Conductors (CICC) are put into use in both the TF and PF coil systems. The TF coil system provides a field of 3.5 T at the plasma center and the PF coil system is able to provide a flux swing of 17 V-sec. The major achievement in KSTAR magnet-system development includes the development of CICC, the development of a full-size TF model coil, the development of a coil system for background magnetic-field generation, the construction of a large-scale superconducting magnet and CICC test facility. TF and PF coils are in the stage of fabrication to pave the way for the scheduled completion of KSTAR by the end of 2006.

Design aspects of a MSE diagnostic for ITER

The motional Stark effect (MSE) diagnostic is unique in its ability to measure the current profile and will be essential in the International Thermonuclear Experimental Reactor (ITER) for detailed analysis of Advanced Tokamak and other types of discharges. However, design of a MSE diagnostic for ITER presents many unique challenges. Among these is optical analysis for the convoluted optical path, required for effective neutron shielding, that employs several reflective optics arranged to form a labyrinth. The geometry of the diagnostic has been laid out and the expected Doppler shifts and channel resolution calculated. A model of the optical train has also been developed based on the Mueller matrix formalism. Unfolding the pitch angle for this complicated geometry is not straightforward and possible methods are evaluated. The CORSICA code is used to model a variety of ITER discharges including start-up, Ip-ramp, and reverse shear. The code also incorporates a synthetic MSE diagnostic that can be used to evaluate different viewing locations and optimize channel locations for the above discharges. Simulation of the optical emission spectrum is also under way.

Design integration of liquid surface divertors

Abstract The US Enabling Technology Program in fusion is investigating the use of free flowing liquid surfaces facing the plasma. We have been studying the issues in integrating a liquid surface divertor into a configuration based upon an advanced tokamak, specifically the ARIES-RS configuration. The simplest form of such a divertor is to extend the flow of the liquid first wall into the divertor and thereby avoid introducing additional fluid streams. In this case, one can modify the flow above the divertor to enhance thermal mixing. For divertors with flowing liquid metals (or other electrically conductive fluids) MHD (magneto-hydrodynamics) effects are a major concern and can produce forces that redirect flow and suppress turbulence. An evaluation of Flibe (a molten salt) as a working fluid was done to assess a case in which the MHD forces could be largely neglected. Initial studies indicate that, for a tokamak with high power density, an integrated Flibe first wall and divertor does not seem workable. We have continued work with molten salts and replaced Flibe with Flinabe, a mixture of lithium, sodium and beryllium fluorides, that has some potential because of its lower melting temperature. Sn and Sn–Li have also been considered, and the initial evaluations on heat removal with minimal plasma contamination show promise, although the complicated 3D MHD flows cannot yet be fully modeled. Particle pumping in these design concepts is accomplished by conventional means (ports and pumps). However, trapping of hydrogen in these flowing liquids seems plausible and novel concepts for entrapping helium are also being studied.

Transport and Stability Studies in Negative Central Shear Advanced Tokamak Plasmas

Achieving high performance for long duration is a key goal of advanced tokamak research around the world. To this end, tokamak experiments are focusing on obtaining (a) a high fraction of well-aligned noninductive plasma current, (b) internal transport barriers (ITBs) in the ion and electron transport channels over a wide radial region with transport approaching neoclassical values, and (c) control of resistive wall modes and neoclassical tearing modes that limit the achievable beta. A current profile that yields a negative central magnetic shear (NCS) in the core is consistent with this focus; NCS is conducive for obtaining ITBs, a high degree of bootstrap current alignment, and reaching the second stability region for ideal ballooning modes, while being stable to ideal kink modes at high beta with wall stabilization and neoclassical tearing modes in the core NCS region. Much progress has been made in obtaining advanced performance in several tokamaks through an increasing understanding of the stability and transport properties of tokamak plasmas. Radio-frequency and neutral beam current drive scenarios are routinely developed and implemented in experiments to access new advanced regimes and control plasma profiles. Short-duration and sustained ITBs have been obtained in the ion and electron channels. The formation of an ITB is attributable to the stabilization of ion and electron temperature gradient and trapped electron modes by the negative shear and by the enhanced E × B flow shear rate and rarefaction of resonant surfaces near the rational qmin values. The progress in understanding the underlying physics in such plasmas and the development of techniques and technology would be of interest in stellarator efforts.

Operating Characteristics of the KSTAR Superconducting TF Coil

The Korea Superconducting Tokamak Advanced Research (KSTAR) device, a steady-state-capable advanced Tokamak, is to be built in Korea. The KSTAR device is made up of 16 toroidal field (TF) and 14 poloidal field (PF) superconducting coils. A simulation code for the transient operation of the KSTAR TF and PF coils is developed. For stable Tokamak operation, the operating characteristics of the KSTAR superconducting magnet system are studied for operation scenario. We calculated the temperature margin of superconducting cables, cryogenic flow parameters and the heat deposition in TF and PF coils. The effect of the 3D heat conduction and the eddy current loss in the supporting structure and vacuum chamber of KSTAR is also included.

Real-time identification of the resistive-wall-mode in DIII-D with Kalman filter ELM discrimination

Abstract The resistive-wall-mode (RWM) is a major performance-limiting instability in present-day tokamaks. Active control and stabilization of the mode will almost certainly be essential for the success of advanced tokamaks and for the economic viability of tokamak fusion reactors. High performance tokamak plasmas often experience edge-localized-modes (ELMs) which can interfere with RWM identification and control. If the RWM control scheme reacts to an ELM the RWM may be driven unstable instead of controlled. An algorithm for real-time identification of the RWM with discrimination of ELMs in the DIII-D tokamak has been developed using a combination of matched filter and Kalman filter methods. The algorithm has been implemented in DIII-D’s real-time plasma control system (PCS) and is available to drive active mode control schemes.

High performance advanced tokamak regimes in DIII-D for next-step experiments

Advanced Tokamak (AT) research in DIII-D [K. H. Burrell for the DIII-D Team, in Proceedings of the 19th Fusion Energy Conference, Lyon, France, 2002 (International Atomic Energy Agency, Vienna, 2002) published on CD-ROM] seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles, and active magnetohydrodynamic stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization via plasma rotation and active feedback with nonaxisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these conditions, bootstrap supplies most of the current. Steady-state operation requires replacing the remaining Ohmic current, mostly located near the half radius, with noninductive external sources. In DIII-D this current is provided by ECCD, and nearly stationary AT discharges have been sustained with little remaining Ohmic current. Fast wave current drive is being developed to control the central magnetic shear. Density control, with divertor cryopumps, of AT discharges with edge localized moding H-mode edges facilitates high current drive efficiency at reactor relevant collisionalities. A sophisticated plasma control system allows integrated control of these elements. Close coupling between modeling and experiment is key to understanding the separate elements, their complex nonlinear interactions, and their integration into self-consistent high performance scenarios. Progress on this development, and its implications for next-step devices, will be illustrated by results of recent experiment and simulation efforts.

Modeling of feedback and rotation stabilization of the resistive wall mode in tokamaks

Steady-state operation of the advanced tokamak reactor relies on maintaining plasma stability with respect to the resistive wall mode (RWM). Active magnetic feedback and plasma rotation are the two methods proposed and demonstrated for this purpose. A comprehensive modeling effort including both magnetic feedback and plasma rotation is needed for understanding the physical mechanisms of the stabilization and to project to future devices. For plasma with low rotation, a complete solution for the feedback issue is obtained by assuming the plasma obeys ideal magnetohydrodynamics (MHDs) and utilizing a normal mode approach (NMA) [M. S. Chu et al., Nucl. Fusion 43, 441 (2003)]. It is found that poloidal sensors are more effective than radial sensors and coils inside of the vacuum vessel more effective than outside. For plasmas with non-negligible rotation, a comprehensive linear nonideal MHD code, the MARS-F has been found to be suitable. MARS-F [Y. Q. Liu et al., Phys. Plasmas 7, 3681 (2000)] has been benchmarked in the ideal MHD limit against the NMA. The effect of rotation stabilization of the plasma depends on the plasma dissipation model. Broad qualitative features of the experiment are reproduced. Rotation reduces the feedback gain required for RWM stabilization. Reduction is significant when rotation is near the critical rotation speed needed for stabilization. The International Thermonuclear Experimental Reactor (ITER) [R. Aymar et al., Plasma Phys. Controlled Fusion 44, 519 (2002)] (scenario IV for advanced tokamak operation) may be feedback stabilized with β above the no wall limit and up to an increment of ∼50% towards the ideal limit. Rotation further improves the stability.