Advanced Tokamak Regimes Research Articles

Top Launch for Higher Off-axis Electron Cyclotron Current Drive Efficiency

Efficient off-axis current drive is crucial for economic, steady-state tokamak fusion power plants. “Top Launch” electron cyclotron current drive (ECCD) is one promising method for driving strong off-axis current to achieve the desired broad current profile for the Advanced Tokamak regime. New simulations show that by launching the electron cyclotron (EC) waves downwards (or upwards) nearly parallel to the resonance plane with a large toroidal steering, higher current drive efficiency can be obtained at large radii owing to the large Doppler shift, wave-particle interactions on HFS of the plasma, and the long absorption path. Implementations on CFETR and DIII-D are simulated identifying clear benefits and optimal configurations. The design of a prototype test being installed on DIII-D is presented, which will experimentally validate whether top launch ECCD can be an improved efficiency current drive technique for future fusion reactors.

High-beta, steady-state hybrid scenario on DIII-D

The potential of the hybrid scenario (first developed as an advanced inductive scenario for high fluence) as a regime for high-beta, steady-state plasmas is demonstrated on the DIII-D tokamak. These experiments show that the beneficial characteristics of hybrids, namely safety factor ⩾1 with low central magnetic shear, high stability limits and excellent confinement, are maintained when strong central current drive (electron cyclotron and neutral beam) is applied to increase the calculated non-inductive fraction to ≈100% (≈50% bootstrap current). The best discharges achieve normalized beta of 3.4, IPB98(y,2) confinement factor of 1.4, surface loop voltage of 0.01 V, and nearly equal electron and ion temperatures at low collisionality. A 0D physics model shows that steady-state hybrid operation with Qfus ~ 5 is feasible in FDF and ITER. The advantage of the hybrid scenario as an advanced tokamak regime is that the external current drive can be deposited near the plasma axis where the efficiency is high; additionally, good alignment between the current drive and plasma current profiles is not necessary as the poloidal magnetic flux pumping self-organizes the current density profile in hybrids with an tearing mode.

Physical program and diagnostics of the T-15 upgrade tokamak (brief overview)

Kurchatov Institute is upgrading now the T-15 tokamak to the machine with D-shaped plasma and copper magnetic system, capable for realizing lower and upper single-null and double-null magnetic configurations. The heating and current drive (CD) system consisting of the neutral beam injection (NBI), electron cyclotron resonance heating (ECRH/CD), electron Bernstein waves (EBW) heating and CD, ion cyclotron resonance heating (ICRH/CD), helicon and Lower Hybrid (LH) waves heating and CD is aiming to provide an effective heating of both electrons and ions, and on- and off-axis CD. The main research topics foreseen are the features of the confinement at high magnetic field and low aspect ratio, Advanced Tokamak regimes, steady-state operation, effects of turbulence with an emphasis on the role of the radial electric field Er, Geodesic Acoustic Modes (GAM) and Zonal Flows (ZF) in transport and confinement (including plasma self-organization, profile resiliency, influence of the q-profile), investigations of MHD effects and disruptions, Alfvén Eigenmodes (AE) and fast particles. Extended set of advanced diagnostics with identical equipment located at two toroidal positions will contribute to the 3D reconstruction of various types of the plasma structures like quasicoherent modes and long-range correlations.

Progress towards steady-state regimes in Alcator C-Mod

Recent progress on lower hybrid current drive (LHCD) experiment and simulation towards steady-state (t ⩾ 3–5 × τr, where τr is the current relaxation time) regimes on Alcator C-Mod is presented. Highly non-inductive reversed shear plasmas are obtained with spontaneous generation of internal transport barriers at the density close to what is expected on ITER steady-state scenarios. Progress has been made to better understand and mitigate the unexpected degradation of LHCD efficiency at high densities, which poses an issue both for extending the non-inductive plasmas to advanced tokamak regimes (with a high bootstrap current fraction, fBS) on C-Mod and for predicting the performance of LHCD on future devices such as ITER. Several physics mechanisms that potentially contribute anomalous losses of LHCD power have been studied extensively. Numerical modelling of collisional absorption in cold scrape-off layer plasmas has been integrated into a ray-tracing code. The LHEAF full-wave code clarifies that full-wave effects move the power deposition profile closer to the separatrix than a calculation based on the WKB approximation, leading to a lower efficiency. Non-linear wave interactions were studied experimentally by LH wave spectral measurements using Langmuir probes, suggesting parametric decay instabilities to explain the remaining difference between experiments and simulations. An experimental demonstration of recovery of good LHCD efficiency at high densities is also reported. Improving the single pass absorption is proposed as a key to recover LHCD efficiency. A wave physics design of additional launcher (LH3) has been developed in order to demonstrate an improved LHCD performance by realizing high (∼80%) single pass absorption, in which the synergistic interaction in the velocity space with the existing launcher is maximized.

Physics of resistive wall modes

The advanced tokamak regime is a promising candidate for steady-state tokamak operation which is desirable for a fusion reactor. This regime is characterized by a high bootstrap current fraction and a flat or reversed safety factor profile, which leads to operation close to the pressure limit. At this limit, an external kink mode becomes unstable. This external kink is converted into the slowly growing resistive wall mode (RWM) by the presence of a conducting wall. Reduction of the growth rate allows one to act on the mode and to stabilize it. There are two main factors which determine the stability of the RWM. The first factor comes from external magnetic perturbations (error fields, resistive wall, feedback coils, etc). This part of RWM physics is the same for tokamaks and reversed field pinch configurations. The physics of this interaction is relatively well understood and based on classical electrodynamics. The second ingredient of RWM physics is the interaction of the mode with plasma flow and fast particles. These interactions are particularly important for tokamaks, which have higher plasma flow and stronger trapped particle effects. The influence of the fast particles will also be increasingly more important in ITER and DEMO which will have a large fraction of fusion born alpha particles. These interactions have kinetic origins which make the computations challenging since not only particles influence the mode, but also the mode acts on the particles. Correct prediction of the ‘plasma–RWM’ interaction is an important ingredient which has to be combined with external field's influence (resistive wall, error fields and feedback) to make reliable predictions for RWM behaviour in tokamaks. All these issues are reviewed in this paper.

Lower hybrid current drive at high density in the multi-pass regime

Assessing the performance of lower hybrid current drive (LHCD) at high density is critical for developing non-inductive current drive systems on future steady-state experiments. Excellent LHCD efficiency has been observed during fully non-inductive operation (η=2.0−2.5×1019 AW–1m–2 at n¯e=0.5×1020 m–3) on Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] under conditions (ne, magnetic field and topology, and LHCD frequency) relevant to ITER [S. Shiraiwa et al., Nucl. Fusion 51, 103024 (2011)]. To extend these results to advanced tokamak regimes with higher bootstrap current fractions on C-Mod, it is necessary to increase n¯e to 1.0−1.5×1020 m−3. However, the number of current-carrying, non-thermal electrons generated by LHCD drops sharply in diverted configurations at densities that are well below the density limit previously observed on limited tokamaks. In these cases, changes in scrape off layer (SOL) ionization and density profiles are observed during LHCD, indicating that significant power is transferred from the LH waves to the SOL. Fokker-Planck simulations of these discharges utilizing ray tracing and full wave propagation codes indicate that LH waves in the high density, multi-pass absorption regime linger in the plasma edge, and SOL region, where absorption near or outside the LCFS results in the loss of current drive efficiency. Modeling predicts that non-thermal emission increases with stronger single-pass absorption. Experimental data show that increasing Te in high density LH discharges results in higher non-thermal electron emission, as predicted by the models.

Core transport properties in JT-60U and JET identity plasmas

The paper compares the transport properties of a set of dimensionless identity experiments performed between JET and JT-60U in the advanced tokamak regime with internal transport barrier, ITB. These International Tokamak Physics Activity, ITPA, joint experiments were carried out with the same plasma shape, toroidal magnetic field ripple and dimensionless profiles as close as possible during the ITB triggering phase in terms of safety factor, normalized Larmor radius, normalized collision frequency, thermal beta, ratio of ion to electron temperatures. Similarities in the ITB triggering mechanisms and sustainment were observed when a good match was achieved of the most relevant normalized profiles except the toroidal Mach number. Similar thermal ion transport levels in the two devices have been measured in either monotonic or non-monotonic q-profiles. In contrast, differences between JET and JT-60U were observed on the electron thermal and particle confinement in reversed magnetic shear configurations. It was found that the larger shear reversal in the very centre (inside normalized radius of 0.2) of JT-60U plasmas allowed the sustainment of stronger electron density ITBs compared with JET. As a consequence of peaked density profile, the core bootstrap current density is more than five times higher in JT-60U compared with JET. Thanks to the bootstrap effect and the slightly broader neutral beam deposition, reversed magnetic shear configurations are self-sustained in JT-60U scenarios. Analyses of similarities and differences between the two devices address key questions on the validity of the usual assumptions made in ITER steady scenario modelling, e.g. a flat density profile in the core with thermal transport barrier? Such assumptions have consequences on the prediction of fusion performance, bootstrap current and on the sustainment of the scenario.

FAST: Feasibility Analysis for a Completely Superconducting Magnet System

FAST (Fusion Advanced Studies Torus), the Italian proposal for a European satellite facility to ITER, is a compact tokamak ( R <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">0</sub> = 1.82 m, a= 0.64 m, triangularity δ = 0.4) able to investigate non linear dynamics effects of alpha-particle behavior in burning plasmas and to test technical solutions for the first wall/divertor directly relevant for ITER and DEMO (e.g.: full-tungsten wall and divertor and advanced liquid metal divertor). The machine is designed to operate with Deuterium plasmas in a dimensionless parameter range close to that of ITER and to access advanced tokamak regimes with long pulse duration with respect to the current diffusion time. It foresees a maximum magnetic field on plasma axis of 8.5 T and a maximum plasma current of 8 MA. In the present design phase, the feasibility of a superconducting solution for the magnet system is being investigated by ENEA. It consists of 18 Toroidal Field, 6 Poloidal Field and 6 Central Solenoid module coils, all of which wound by Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn and NbTi Cable-In-Conduit Conductors. All the main aspects driving the magnets' design, from mechanical to neutronic and thermal analyses, are here presented and discussed.

Saturated ideal modes in advanced tokamak regimes in MAST

MAST plasmas with a safety factor above unity and a profile with either weakly reversed shear or broad low-shear regions, regularly exhibit long-lived saturated ideal magnetohydrodynamic (MHD) instabilities. The toroidal rotation is flattened in the presence of such perturbations and the fast ion losses are enhanced. These ideal modes, distinguished as such by the notable lack of islands or signs of reconnection, are driven unstable as the safety factor approaches unity. This could be of significance for advanced scenarios, or hybrid scenarios which aim to keep the safety factor just above rational surfaces associated with deleterious resistive MHD instabilities, especially in spherical tokamaks which are more susceptible to such ideal internal modes. The role of rotation, fast ions and ion diamagnetic effects in determining the marginal mode stability is discussed, as well as the role of instabilities with higher toroidal mode numbers as the safety factor evolves to lower values.

Hard x-ray diagnostic for lower hybrid experiments on Alcator C-Mod

Alcator C-Mod’s lower hybrid current drive (LHCD) system allows the exploration of advanced tokamak (AT) regimes at densities relevant to ITER and fusion reactors. The location of the LHCD is critical to AT performance and may be inferred by measuring the nonthermal bremsstrahlung emission in the hard x-ray (HXR) region. A pinhole camera using an array of 32 CdZnTe detectors is used to image energies in the 20–200keV range. Detectors and pulse processing electronics are integrated into a compact and modular package making extensive use of printed circuit board and surface mount technology. The system also makes use of fast digitization and software signal processing techniques. An ambient environment of neutrons, gammas, and high rf power requires careful shielding. Shielding is studied using the neutron and photon transport code MCNP. The design of the diagnostic is presented along with background measurements in lieu of LHCD fast electrons. Background measurements are then compared to advanced modeling results to predict the power threshold for meaningful HXR data for a H-mode target plasma.

Advanced tokamak research on long time scales in JT-60 Upgrade

The duration of advanced tokamak plasma operation in JT-60 Upgrade (JT-60U) [A. Kitsunezaki et al., Fusion Sci. Technol. 42, 179 (2002)] has been extended on long time scales exceeding the current diffusion time (τR) and close to the wall saturation time. A very high bootstrap current fraction (fBS) of 75% was maintained for 7.4s (2.7τR), while a normalized beta of 2.5 was maintained for 15.5s (∼9.5τR). The current profile reaches stationary conditions in ∼τR for the small fBS regime, while a longer time is required for the large fBS regime. A plasma with a weak shear q profile, similar to requirements for the ITER steady-state operational scenario, was successfully maintained. The particle inventory in the wall was observed to saturate in repeated long-pulse (∼30s) H-mode discharges with edge localized modes. The analysis of neutral particles in the scrape-of-layer plasma indicates different time scales are involved for the wall saturation. Ferritic steel tiles have been installed in the vacuum vessel to reduce the toroidal field ripple towards extending the advanced tokamak regime further in the coming operational campaign.

The Alcator C-Mod lower hybrid current drive experiment transmitter

Abstract Alcator C-Mod, is a high-field high-density, diverted, compact tokamak, which, in its present form uses inductive current drive and is heated with 5 MW of ICRF auxiliary power. C-Mod is in the process of being upgraded with a 4.6 GHz lower hybrid heating and current drive (LHCD) system. The purpose of the experiment is to develop and explore the potential of “Advanced Tokamak Regimes” under quasi-steady-state conditions. In this paper, an overview of the RF transmitter and the controls and protection systems for the Lower Hybrid Project is given. The transmitter will use 12 250 kW klystrons operating simultaneously which will result in a total nominal power at the klystrons of nearly 3 MW for a planned pulse width of 5 s. Active control system vector modulators provide phase and amplitude drive for each klystron, and I-Q detectors are used to monitor phase and amplitude. These feedback signals are used in digital controllers for closed-loop control of klystron phase and amplitude to preset values. An expected upgrade of four additional klystrons will result in a total nominal power of 4 MW. The transmitters have been tested to full power, and installation of the lower hybrid current drive experiment on the C-Mod Tokamak is expected in 2004/2005.

Integrated Plasma Control in DIII-D

The integrated plasma control approach provides a systematic method for designing plasma control algorithms with high reliability and for confirming their performance off-line prior to experimental implementation. This approach includes construction of plasma and system response models, validation of models against operating experiments, design of integrated controllers that operate in concert with one another as well as with supervisory modules, simulation of control action against off-line and actual machine control platforms, and iteration of the design-test loop to optimize performance. Using this approach, required levels of robustness to model uncertainties and off-normal events can be quantified and incorporated in the design process. The DIII-D digital plasma control system (PCS) enables application of this method by providing a flexible programming environment and an architecture for real-time parallel operation of a set of computers that executes the large set of control algorithms needed for exploration of the advanced tokamak regime. The present work describes the DIII-D PCS and the approach, benefits, and progress made in integrated plasma control as applied to the DIII-D tokamak, with implications for the International Thermonuclear Experimental Reactor design and other next-generation tokamaks.

High performance integrated plasma control in DIII-D

Abstract The DIII-D mission to explore the advanced tokamak (AT) regime places significant demands on the DIII-D plasma control system (PCS) [D.A. Humphreys, et al., Fusion Eng. Des. 66–68 (2003) 633], including simultaneous and accurate regulation of plasma shape, stored energy, and density, as well as coordinated suppression of magnetohydrodynamic (MHD) instabilities. The present work describes selected new solutions to key control problems in DIII-D AT operation. These include novel nonrigid, resistive linear plasma response models, nonlinear algorithms for avoidance of current limiting, and improved neoclassical tearing mode control algorithms.

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.

Status of and prospects for advanced tokamak regimes from multi-machine comparisons using the ‘International Tokamak Physics Activity’ database

Advanced tokamak regimes obtained in ASDEX Upgrade, DIII-D, FT-U, JET, JT-60U, TCV and Tore Supra experiments are assessed both in terms of their fusion performance and capability for ultimately reaching steady-state using data from the international internal transport barrier database. These advanced modes of tokamak operation are characterized by an improved core confinement and a modified current profile compared to the relaxed Ohmically driven one. The present results obtained in these experiments are studied in view of their prospect for achieving either long pulses (‘hybrid’ scenario with inductive and non-inductive current drive) or ultimately steady-state purely non-inductive current drive operation in next step devices such as ITER. A new operational diagram for advanced tokamak operation is proposed where the figure of merit characterizing the fusion performances and confinement, , is drawn versus the fraction of the plasma current driven by the bootstrap effect. In this diagram, present day advanced tokamak regimes have now reached an operational domain that is required in the non-inductive ITER current drive operation with typically 50% of the plasma current driven by the bootstrap effect (Green et al 2003 Plasma Phys. Control. Fusion 45 587). In addition, the existence domain of the advanced mode regimes is also mapped in terms of dimensionless plasmas physics quantities such as normalized Larmor radius, normalized collisionality, Mach number and ratio of ion to electron temperature. The gap between present day and future advanced tokamak experiments is quantitatively assessed in terms of these dimensionless parameters.

Chapter 9: Performance-Limiting MHD Activity and Possibilities for Its Stabilization in ASDEX Upgrade

Performance-limiting magnetohydrodynamic (MHD) instabilities on ASDEX Upgrade are discussed. In the conventional H-mode scenario, the main MHD performance limitation is found to be the neoclassical tearing mode (NTM). The onset β of NTMs in ASDEX Upgrade scales with the poloidal ion gyroradius, in agreement with theoretical expectations. At higher β values, NTMs occur in a more benign form, the frequently-interrupted-regime NTMs, which lead to a smaller confinement degradation than normal NTMs. Active control of NTMs by electron cyclotron current drive in the island has been demonstrated on ASDEX Upgrade. In advanced tokamak regimes with reversed shear, a variety of performance-limiting instabilities has been observed. The shear reversal zone can be unstable to double tearing modes or to infernal modes; both have been identified in ASDEX Upgrade. Due to the broad current profile in advanced tokamak discharges, the ideal external kink mode can be unstable at relatively low βN ≤ 2; this is a main limitation to strongly reversed shear discharges with peaked pressure profiles. Finally, it is shown that fast-particle-driven modes such as fishbones can also have beneficial effects, such as providing stationary current profiles or triggering internal transport barriers.

Technological development and progress of plasma performance on the JT-60U

The plasma performance exploration of JT-60U in advanced tokamak regimes based on high βp and reversed shear mode operation has been conducted intensively, by using 500 keV negative-ion based neutral beam injection (N-NBI) and 110 GHz electron cyclotron (EC) systems for plasma heating and current drive, and a repetitive centrifugal pellet injector for efficient core particle fueling. High power injection of 6.2 MW at 381 keV for 1.7 s and a long pulse injection of 10 s at 2.6 MW and 355 keV were achieved. In the EC system capable of generating 4 MW of rf power for 5 s, the maximum injected EC power reached 3.0 MW for 2.7 s so far. Pellet injection at a velocity of 100 m/s from the mid-plane in the HFS has also performed well. With application of N-NBI power of 5.7 MW at 402 keV, a high fusion triple product was increased to n i (0) τ E T i(0), of 3×10 20 keV s/m 3 under a full non-inductive current drive condition. By using steerable antennas of the EC system and a tracking technique of the MHD island, suppression of the neoclassical tearing mode in real time has been demonstrated successfully. High-magnetic-field-side pellet injection extended the plasma density up to 0.7 n GW with H 89p kept at ∼2.

Design of a Compact Lower Hybrid Coupler for Alcator C-Mod

Princeton Plasma Physics Laboratory and the Massachusetts Institute of Technology are preparing an experiment of current profile control using lower hybrid waves to produce and sustain advanced tokamak regimes in steady-state conditions in Alcator C-Mod. Unlike the Joint European Torus, ToreSupra, and JT60 couplers, the C-Mod lower hybrid coupler does not employ the now conventional multijunction design but will have similar characteristics, compactness, and internal power division while retaining full control of the antenna element phasing. This is achieved by using 3-dB vertical power splitters and a stack of laminated plates with the waveguides milled in them. Construction is simplified and allows easy control and maintenance of all parts. Many precautions are taken to avoid arcing. Special care is also taken to avoid the recycling of reflected power, which could affect the coupling and the launched n॥ spectrum. The results from C-Mod should allow further simplification in the designs of the coupler planned for KSTAR and ITER.

Physics basis and simulation of burning plasma physics for the fusion ignition research experiment (FIRE)

The FIRE design for a burning plasma experiment is described in terms of its physics basis and engineering features. Systems analysis indicates that the device has a wide operating space to accomplish its mission, both for the ELMy H-mode reference and the high bootstrap current/high-β advanced tokamak regimes. Simulations with 1.5D transport codes reported here both confirm and constrain the systems projections. Experimental and theoretical results are used to establish the basis for successful burning plasma experiments in FIRE.