Conventional Aspect Ratio Tokamaks Research Articles

Collective fast ion instability-induced losses in National Spherical Tokamak Experiment

A wide variety of fast ion driven instabilities are excited during neutral beam injection (NBI) in the National Spherical Torus Experiment (NSTX) [Nucl. Fusion 40, 557 (2000)] due to the large ratio of fast ion velocity to Alfvén velocity, Vfast∕VAlfvén, and high fast ion beta. The ratio Vfast∕VAlfvén in ITER [Nucl. Fusion 39, 2137 (1999)] and NSTX is comparable. The modes can be divided into three categories: chirping energetic particle modes (EPM) in the frequency range 0 to 120kHz, the toroidal Alfvén eigenmodes (TAE) with a frequency range of 50kHz to 200kHz, and the compressional and global Alfvén eigenmodes (CAE and GAE, respectively) between 300kHz and the ion cyclotron frequency. Fast ion driven modes are of particular interest because of their potential to cause substantial fast ion losses. In all regimes of NBI heated operation we see transient neutron rate drops, correlated with bursts of TAE or fishbone-like EPMs. The fast ion loss events are predominantly correlated with the EPMs, although losses are also seen with bursts of multiple, large amplitude TAE. The latter is of particular significance for ITER; the transport of fast ions from the expected resonance overlap in phase space of a “sea” of large amplitude TAE is the kind of physics expected in ITER. The internal structure and amplitude of the TAE and EPMs has been measured with quadrature reflectometry and soft x-ray cameras. The TAE bursts have internal amplitudes of ñ∕n=1% and toroidal mode numbers 2<n<7. The EPMs are core localized, kink-like modes similar to the fishbones in conventional aspect ratio tokamaks. Unlike the fishbones, the EPMs can be present with q(0)>1 and can have a toroidal mode number n>1. The range of the frequency chirp can be quite large and the resonance can be through a fishbone-like precessional drift resonance, or through a bounce resonance.

Progress towards high-performance, steady-state spherical torus

Research on the spherical torus (or spherical tokamak) (ST) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect ratio devices, such as the conventional tokamak. The ST experiments are being conducted in various US research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium sized ST research facilities: PEGASUS at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (β), non-inductive sustainment, Ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values βT of up to 35% with a near unity central βT have been obtained. NSTX will be exploring advanced regimes where βT up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta poloidal regime, where discharges with a high non-inductive fraction (∼60% bootstrap current+NBI current drive) were sustained over the resistive skin time. Research on radio-frequency (RF) based heating and current drive utilizing high harmonic fast wave and electron Bernstein wave is also pursued on NSTX, PEGASUS, and CDX-U. For non-inductive start-up, the coaxial helicity injection, developed in HIT/HIT-II, has been adopted on NSTX to test the method up to Ip ∼ 500 kA. In parallel, start-up using a RF current drive and only external poloidal field coils are being developed on NSTX. The area of power and particle handling is expected to be challenging because of the higher power density expected in the ST relative to that in conventional aspect-ratio tokamaks. Due to its promise for power and particle handling, liquid lithium is being studied in CDX-U as a potential plasma-facing surface for a fusion reactor.

H-mode threshold and dynamics in the National Spherical Torus Experiment

Edge parameters play a critical role in high confinement mode (H-mode) access, which is a key component of discharge optimization in present day toroidal confinement experiments and the design of next generation devices. Because the edge magnetic topology of a spherical torus (ST) differs from a conventional aspect ratio tokamak, H-modes in STs exhibit important differences compared with tokamaks. The dependence of the National Spherical Torus Experiment (NSTX) [C. Neumeyer et al., Fusion Eng. Des. 54, 275 (2001)] edge plasma on heating power, including the low confinement mode (L-mode) to H-mode (L-H) transition requirements and the occurrence of edge-localized modes (ELMs), and on divertor configuration is quantified. Comparisons between good L-modes and H-modes show greater differences in the ion channel than the electron channel. The threshold power for the H-mode transition in NSTX is generally above the predictions of a recent International Tokamak Experimental Reactor (ITER) [ITER Physics Basis Editors, Nucl. Fusion 39, 2175 (1999)] scaling. Correlations of transition and ELM phenomena with turbulent fluctuations revealed by gas puff imaging and reflectometry are observed. In both single-null and double-null divertor discharges, the density peaks off-axis, sometimes developing prominent “ears” which can be sustained for many energy confinement times, τE, in the absence of ELMs. A wide variety of ELM behavior is observed, and ELM characteristics depend on configuration and fueling.

Next-step-targeted experiments on the Mega-Amp Spherical Tokamak

Since its first physics campaign, the principal parameters on MAST (Mega-Amp Spherical Tokamak) [A. Sykes et al., Nuclear Fusion 41, 1423 (2001)] have been brought up towards their design values. Considerable advances have been made in a range of physics areas of direct relevance to the International Thermonuclear Experimental Reactor (ITER) [ITER Physics Basis, Nuclear Fusion 39, 2175 (1999)]. In this paper, results on H-mode access, global confinement and pedestal studies are presented and compared with conventional aspect ratio tokamak scalings. Physics and engineering requirements relevant to next step spherical tokamak devices are discussed, in particular the plasma formation, current ramp-up and sustainment, and plasma exhaust. Results of first experiments directly targeting these issues are presented: Plasma current up to 0.5 MA has been produced without use of the central solenoid flux, and current ramp-up and sustainment without use of the central solenoid flux has been demonstrated. Experiments on neutral beam heating and current drive (CD) demonstrate up to 50% bootstrap current fraction and good CD efficiency, and divertor power loading has been found to be tolerable and have a favorable outboard asymmetry.

Recent results from the National Spherical Torus Experiment

The National Spherical Torus Experiment (NSTX) is a low aspect-ratio fusion research facility whose research goal is to make a determination of the attractiveness of the spherical torus concept in the areas of high-β stability, confinement, current drive, and divertor physics. Remarkable progress was made in extending the operational regime of the device in FY 2002. In brief, βt of 34% and βN of 6.5 were achieved. H-mode became the main operational regime, and energy confinement exceeded conventional aspect-ratio tokamak scalings. Heating was demonstrated with the radiofrequency antenna, and signatures of current drive were observed. Current initiation with coaxial helicity injection produced discharges of 400 kA, and first measurements of divertor heat flux profiles in H-mode were made.

H-mode plasmas in the MAST spherical tokamak

Operating scenarios in the Mega-Ampère Spherical Tokamak (MAST) have been recently developed to access routinely H-mode plasmas over a wide range of plasma conditions. Gas refuelling at the inboard mid-plane has proved invaluable in facilitating L- to H-mode transitions. Already, a comprehensive range of H-mode behaviour has been observed providing similar and contrasting comparisons with conventional tokamaks. For example, similarities are found in H-mode edge and global evolutions, increases in confinement and ELM behaviour. In contrast, H-mode threshold powers are about an order of magnitude higher in MAST than those predicted by conventional aspect ratio tokamak scalings. At the highest βθ values obtained so far, transient confinement times of ~100 ms are achieved and calculations indicate that bootstrap currents may approach 50% of the plasma current.

Characteristics of the first H-mode discharges in the national spherical torus experiment.

We report observations of the first low-to-high ( L-H) confinement mode transitions in the National Spherical Torus Experiment. The H-mode energy confinement time increased over reference discharges transiently by 100-200%, as high as approximately 100 ms. This confinement time is approximately 2 times higher than predicted by a multimachine scaling. Thus the confinement time of spherical tori has been extended to a record high value, leading to an eventual revision of confinement scalings. Finally, the power threshold for H-mode access is >10x higher than predicted by an international scaling from conventional aspect-ratio tokamaks, which could lead to new understanding of H-mode transition dynamics.

Equilibrium reconstruction in the START tokamak

The computation of magnetic equilibria in the START spherical tokamak is more difficult than those in more conventional large aspect ratio tokamaks. This difficulty arises partly as a result of the use of induction compression to generate high current plasma, as this precludes the positioning of magnetic diagnostics close to the outboard side of the plasma. In addition, the effect of a conducting wall with a high, but finite, conductivity must be included. A method is presented for obtaining plasma equilibrium reconstructions based on the EFIT code. New constraints are used to relate isoflux surface locations deduced from radial profile measurements of electron temperature. A model of flux diffusion through the vessel wall is developed. It is shown that neglecting flux diffusion in the vessel wall can lead to a significant underestimate in the calculation of the plasma βt. Using a relatively sparse set of magnetic signals, βt can be obtained to within a fractional error of ±10%. Using constraints to relate isoflux surface locations, the principle involved in determining the internal q profile is demonstrated.

Engineering design of the National Spherical Torus Experiment

NSTX is a proof-of-principle experiment aimed at exploring the physics of the ‘spherical torus’ (ST) configuration, which is predicted to exhibit more efficient magnetic confinement than conventional large aspect ratio tokamaks, among other advantages. The low aspect ratio ( R/ a, typically 1.2–2 in ST designs compared to 4–5 in conventional tokamaks) decreases the available cross sectional area through the center of the torus for toroidal and poloidal field coil conductors, vacuum vessel wall, plasma facing components, etc., thus increasing the need to deploy all components within the so-called ‘center stack’ in the most efficient manner possible. Several unique design features have been developed for the NSTX center stack, and careful engineering of this region of the machine, utilizing materials up to their engineering allowables, has been key to meeting the desired objectives. The design and construction of the machine has been accomplished in a rapid and cost effective manner thanks to the availability of extensive facilities, a strong experience base from the TFTR era, and good cooperation between institutions.

Steady state operation of spherical tokamaks

For a fusion power plant to be economically and technologically attractive, it should be as compact as possible and capable of `steady state' operation. One approach is based upon the spherical tokamak (ST) concept. This configuration features many of the qualities of the conventional aspect ratio tokamak, such as good confinement and MHD stability, together with a number of highly promising features for the realization of a cost effective, steady state fusion power core. In particular, it allows high β capability due to the high I/B achievable at low A and strong, natural shaping of the configuration, together with the possibility of approaching 100% pressure driven currents through high natural elongation and access to second stability to ideal ballooning modes. The physics insights gained from experiments on START are discussed and the theoretical modelling work summarized, describing how the beneficial properties of STs can be combined in steady state ST power plant designs such as the Culham STPP concept. Where appropriate, we shall refer to the two new mega-amp machines MAST and NSTX, both designed for a plasma duration approaching the plasma current relaxation time and also other new, smaller, STs, designed to expand the knowledge gained so far.

Magneto-hydro-dynamic limits in spherical tokamaks

The operational limits observed in spherical tokamaks, notably the small tight aspect ratio tokamak (START) device [A. Sykes et al., Nucl. Fusion 32, 694 (1992)], are consistent with those found in conventional aspect ratio tokamaks. In particular the highest β achieved (∼40%) is consistent with an ideal magneto-hydro-dynamic (MHD) Troyon type limit, the upper limit on density is well described by the Greenwald density (πa2n̄e/Ip∼1) and the normalized current (Ip/aBt) is limited such that q95≳2. Stability calculations indicate scope for increasing both normalized β and normalized current beyond the values so far achieved, although wall stabilization is generally needed for low-n modes. In double null configurations current terminating disruptions occur at each of the operational boundaries, though the current quench tends to be slow at the density limit and disruptions at high β may be due to the low q. In early limiter START discharges, before the divertor coils were installed, disruptions rarely occurred. Instead internal reconnection events which have all the characteristics of a disruption except the current quench occurred. These various disruptive behaviors are explained in terms of a model in which helicity is conserved during the disruption. Due to the low toroidal field beam ions in START, and α particles in a ST power plant, are super-Alfvénic. This gives the possibility for toroidal Alfvén eigenmodes (TAEs) to occur and such modes are frequently observed in START neutral beam injection (NBI) discharges, but seem to be benign. The features of these observed TAEs are shown to be in agreement with MHD calculations.

The physics of spherical confinement systems

Spherical torus magnetic confinement systems, covering spheromaks and spherical tokamaks (STs), are reviewed. As well as being potentially very important for fusion, spherical tori research is enhancing our understanding of magnetic confinement systems with wider applications than fusion research. The studies contribute to the conventional tokamak, for example, ITER via a range of scalings, as well as to our understanding of `quiescent' plasmas and those subject to `turbulent magnetohydrodynamic (MHD) relaxation'. The theoretical and experimental properties are described, showing how these vary with configuration and contrasting them with the conventional aspect ratio tokamak. Topics covered include equilibrium, refuelling, helicity injection, influence of trapped particle fraction, plasma heating, confinement, stability (including pressure limits and energetic particle instabilities) and disruption resilience.

Fast wave current drive in low aspect ratio tokamaks

The magnetic field of a low aspect ratio tokamak (LART) deviates rather significantly from the approximately 1/R dependence of conventional tokamaks. The parameterization of current drive efficiency on the basis of conventional large aspect ratio tokamaks becomes inaccurate. The physics of high harmonic ion cyclotron fast wave current drive (FWCD) and low frequency FWCD in LARTs is examined using exact evaluations of the current drive efficiencies from the adjoint technique. The results indicate that the high harmonic ion cyclotron current drive cannot penetrate to the plasma core at the high beta values typical of LART reactors, and hence this scheme may only be applicable for off-axis steady state current drive, while the low frequency current drive penetrates well into the plasma core and is appropriate for on-axis current profile control. The implications for advanced scenarios of LART are discussed