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Predictions of the recently developed paleoclassical transport model are compared with data from many toroidal plasma experiments: electron heat diffusivity in DIII-D, C-Mod and NSTX ohmic and near-ohmic plasmas; transport modeling of DIII-D ohmic-level discharges and of the RTP ECH 'stair-step' experiments with eITBs at low order rational surfaces; investigation of a strong eITB in JT-60U; H-mode Te edge pedestal properties in DIII-D; and electron heat diffusivities in non-tokamak experiments (NSTX/ST, MST/RFP, SSPX/spheromak). The radial electron heat transport predicted by the paleoclassical model is found to agree with a wide variety of ohmic-level experimental results and to set the lower limit (within a factor {approx} 2) for the radial electron heat transport in most resistive, current-carrying toroidal plasmas -- unless it is exceeded by fluctuation-induced transport, which often occurs in the edge of L-mode plasmas and when the electron temperature is high ({approx}>T{sub e}{sup crit} {approx}B{sup 2/3}{bar {alpha}}{sup 1/2} keV) because then paleoclassical transport becomes less than gyro-Bohm-level anomalous transport.

In order to reduce recirculating power fraction to acceptable levels, the spherical torus concept relies on the simultaneous achievement of high toroidal β and high bootstrap fraction in steady state. In the last year, as a result of plasma control system improvements, the achievable plasma elongation on NSTX has been raised from κ ∼ 2.1 to κ ∼ 2.6 - approximately a 25% increase. This increase in elongation has led to a substantial increase in the toroidal β for long pulse discharges. The increase in β is associated with an increase in plasma current at nearly fixed poloidal β, which enables higher βtwith nearly constant bootstrap fraction. As a result, for the first time in a spherical torus, a discharge with a plasma current of 1 MA has been sustained for 1 s (0.8 s current flat-top). Data are presented from NSTX correlating the increase in performance with increased plasma shaping capability. In addition to improved shaping, H-modes induced during the current ramp phase of the plasma discharge have been used to reduce flux consumption and to delay the onset of MHD instabilities. Based on these results, a modelled integrated scenario, which has 100% non-inductive current drive with very high toroidal β, will also be discussed. The NSTX poloidal field coils are currently being modified to produce the plasma shape which is required for this scenario, which requires high triangularity (δ ∼ 0.8) at elevated elongation (κ ∼ 2.5). The other main requirement of steady state on NSTX is the ability to drive a fraction of the total plasma current with RF waves. The results of high harmonic fast wave heating and current drive studies as well as electron Bernstein wave emission studies will be presented. © 2006 IAEA, Vienna.

The major objective of the National Spherical Torus Experiment (NSTX) is to understand basic toroidal confinement physics at low aspect ratio and high βT in order to advance the spherical torus (ST) concept. In order to do this, NSTX utilizes up to 7.5 MW of neutral beam injection, up to 6 MW of high harmonic fast waves (HHFWs), and it operates with plasma currents up to 1.5 MA and elongations of up to 2.6 at a toroidal field up to 0.45 T. New facility, and diagnostic and modelling capabilities developed over the past two years have enabled the NSTX research team to make significant progress towards establishing this physics basis for future ST devices. Improvements in plasma control have led to more routine operation at high elongation and high βT (up to ~40%) lasting for many energy confinement times. βT can be limited by either internal or external modes. The installation of an active error field (EF) correction coil pair has expanded the operating regime at low density and has allowed for initial resonant EF amplification experiments. The determination of the confinement and transport properties of NSTX plasmas has benefitted greatly from the implementation of higher spatial resolution kinetic diagnostics. The parametric variation of confinement is similar to that at conventional aspect ratio but with values enhanced relative to those determined from conventional aspect ratio scalings and with a BT dependence. The transport is highly dependent on details of both the flow and magnetic shear. Core turbulence was measured for the first time in an ST through correlation reflectometry. Non-inductive start-up has been explored using PF-only and transient co-axial helicity injection techniques, resulting in up to 140 kA of toroidal current generated by the latter technique. Calculated bootstrap and beam-driven currents have sustained up to 60% of the flat-top plasma current in NBI discharges. Studies of HHFW absorption have indicated parametric decay of the wave and associated edge thermal ion heating. Energetic particle modes, most notably toroidal Alfven eigenmodes and fishbone-like modes result in fast particle losses, and these instabilities may affect fast ion confinement on devices such as ITER. Finally, a variety of techniques has been developed for fuelling and power and particle control.

A broadly based study of the fusion engineering and plasma science conditions of a Component Test Facility (CTF), using the Spherical Torus or Spherical Tokamak (ST) configuration, have been carried out. The chamber systems testing conditions in a CTF are characterized by high fusion neutron fluxes {gamma}{sub n} > 4.4 x 10{sup 13} n/s/cm{sup 2}, over size scales > 10{sup 5} cm{sup 2} and depth scales > 50 cm, delivering > 3 accumulated displacement per atom (dpa) per year. The desired chamber conditions can be provided by a CTF with R{sub 0} 1.2 m, A = 1.5, elongation {approx} 3, I{sub p} {approx} 9 MA, B{sub T} {approx} 2.5 T, producing a driven fusion burn using 36 MW of combined neutral beam and RF power. Relatively robust ST plasma conditions are adequate, which have been shown achievable without active feedback manipulation of the MHD modes. The ST CTF will test the single-turn, copper alloy center leg for the toroidal field coil without an induction solenoid and neutron shielding, and require physics data on solenoid-free plasma current initiation, ramp-up, and sustainment to multiple MA level. A new systems code that combines the key required plasma and engineering science conditions of CTFmore » has been prepared and utilized as part of this study. The results show high potential for a family of lower-cost CTF devices to suit a variety of fusion engineering science test missions.« less

An overview of the research capabilities and the future plans on the MA-class National Spherical Torus Experiment (NSTX) at Princeton is presented. NSTX research is exploring the scientific benefits of modifying the field line structure from that in more conventional aspect ratio devices, such as the tokamak. The relevant scientific issues pursued on NSTX include energy confinement, MHD stability at high beta, non-inductive sustainment, solenoid-free start-up, and power and particle handling. In support of the NSTX research goal, research tools are being developed by the NSTX team. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a high beta Demo device based on the ST, are being considered. For these, it is essential to develop high performance (high beta and high confinement), steady-state (non-inductively driven) ST operational scenarios and an efficient solenoid-free start-up concept. We will also briefly describe the Next-Step-ST (NSST) device being designed to address these issues in fusion-relevant plasma conditions.

Recent progress(1) in plasma science of the Spherical Tokamak (or Spherical Torus, ST)(2) has indicated relatively robust plasma conditions in a broad number of topical area including strong shaping, stability limits, energy confinement, self-driven current, and sustainment. This progress has enabled an extensive update of the plasma science and fusion engineering conditions of a Component Test Facility (CTF)(3), which is potentially a necessary step in the development of practical fusion energy. The chamber systems testing conditions in a CTF are characterized by high fusion neutron fluxes n > 4.4 1013 n/s/cm2, over sizescale > 105 cm2 and depth-scale > 50 cm, delivering > 3 accumulated displacement per atom (dpa) per year(4). Such chamber conditions are calculated to be achievable in a CTF with R0 = 1.2 m, A = 1.5, elongation ~ 3, Ip ~ 9 MA, BT ~ 2.5 T, producing a driven fusion burn using 36 MW of combined neutral beam and RF power. The ST CTF will test the life time of single-turn, copper alloy center leg for the toroidal field coil without an induction solenoid and neutron shielding, and require physics data on solenoid-free plasma current initiation, ramp-up, and sustainment to multiple MA level. A newmore » systems code that combines the key required plasma and engineering science conditions of CTF has been prepared and utilized as part of this study. The results show high potential for a family of relatively low cost CTF devices to suit a range of fusion engineering science test missions.« less

A major research goal of the national spherical torus experiment is establishing long-pulse, high beta, high confinement operation and its physics basis. This research has been enabled by facility capabilities developed during 2001 and 2002, including neutral beam (up to 7?MW) and high harmonic fast wave (HHFW) heating (up to 6?MW), toroidal fields up to 6?kG, plasma currents up to 1.5?MA, flexible shape control, and wall preparation techniques. These capabilities have enabled the generation of plasmas with of up to 35%. Normalized beta values often exceed the no-wall limit, and studies suggest that passive wall mode stabilization enables this for H mode plasmas with broad pressure profiles. The viability of long, high bootstrap current fraction operations has been established for ELMing H mode plasmas with toroidal beta values in excess of 15% and sustained for several current relaxation times. Improvements in wall conditioning and fuelling are likely contributing to a reduction in H mode power thresholds. Electron thermal conduction is the dominant thermal loss channel in auxiliary heated plasmas examined thus far. HHFW effectively heats electrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive is obtained by comparision of the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. Studies of emissions from electron Bernstein waves indicate a density scale length dependence of their transmission across the upper hybrid resonance near the plasma edge that is consistent with theoretical predictions. A peak heat flux to the divertor targets of 10?MW?m?2 has been measured in the H mode, with large asymmetries being observed in the power deposition between the inner and outer strike points. Non-inductive plasma startup studies have focused on coaxial helicity injection. With this technique, toroidal currents up to 400?kA have been driven, and studies to assess flux closure and coupling to other current drive techniques have begun.

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 (?), 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 ?T of up to 35% with a near unity central ?T have been obtained. NSTX will be exploring advanced regimes where ?T 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.

Impurity injection experiments were performed in the National Spherical Torus Experiment [NSTX, M. Ono et al., Nucl. Fusion 40, 557 (2000)] for a first assessment of low-Z impurity transport in a low field, low-aspect ratio device. Short neon puffs were injected in beam heated, co-rotating L-mode (low confinement) discharges and the radial penetration of the successive neon charge states has been monitored with arrays of filtered ultrasoft x-ray diodes. Time-dependent modeling of the neon emissivity in several spectral bands indicates a core diffusion coefficient in the neoclassical range (

H-modes are routinely obtained in the National Spherical Torus Experiment (NSTX) and have become a standard operational scenario. L–H transitions triggered by NBI heating have been obtained over a wide parameter range in Ip, Bt, and e in either lower-single-null (LSN) or double-null (DN) diverted discharges. Edge localized modes are observed in both configurations but the characteristics differ between DN and LSN, which also have different triangularities (δ). An H-mode duration of 500 ms was obtained in LSN, with a total pulse length of ~1 s. Preliminary power threshold studies indicate that the L–H threshold is between 600 kW and 1.2 MW, depending on the target parameters. Gas injector fuelling from the centre stack (i.e. the high toroidal field side) has enabled routine H-mode access, and comparisons with low-field side (LFS) fuelled H-mode discharges show that the LFS fuelling delays the L–H transition and alters the pre-transition plasma profiles. Gas puff imaging and reflectometry show that the H-mode edge is usually more quiescent than the L-mode edge. Divertor infrared camera measurements indicate up to 70% of available power flows to the divertor targets in quiescent H-mode discharges.

It is shown here that microturbulence can be stabilized in the presence of steep temperature and density profiles. Indeed in high β plasmas, pressure profile gradients are associated with high |β′|=−∂β/∂ρ, where β=P/(B2/2μ0) and ρ the square root of the toroidal flux normalized to its edge value. It is shown here that high values of |β′| have a stabilizing influence on drift modes. This may form the basis for a positive feedback loop in which high core beta values lead to improved confinement, and to further increase in β. A gyrokinetic electromagnetic flux tube code, GS2 [M. Kotschenreuther, G. Rewoldt, and W. M. Tang, Comput. Phys. Commun. 88, 128 (1995)], is used for analyzing the microstability. In high β spherical tokamak plasmas, high |β′| rather than low aspect ratio is a source of stabilization. Therefore, the effect of high |β′| should be stabilizing in the plasmas of the National Spherical Torus Experiment [Y.-K. Peng, M. G. Bell, R. E. Bell et al., Phys. Plasmas 7, 1681 (2000)].

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 Edi...

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.

The National Spherical Torus Experiment (NSTX) at Princeton University, Princeton, NJ, is designed to assess the potential of the low-aspect-ratio spherical torus concept for magnetic plasma confinement. The plasma has been heated by up to 7 MW of neutral beam injection (NBI) at an injection energy of 100 keV and up to 6 MW of high harmonic fast wave (HHFW) at 30 MHz. NSTX has achieved /spl beta//sub T/ of 32%. A variety of MHD phenomena have been observed to limit /spl beta/. NSTX has now begun addressing /spl tau//sub E/ scaling, /spl beta/ limits, and current drive issues. During the NBI heating experiments, a broad T/sub i/ profile with T/sub i/ up to 2 keV, T/sub i/>T/sub e/ and a large toroidal rotation were observed. Transport analysis suggests that the impurity ions have diffusivities approaching neoclassical. For L-Mode plasmas, /spl tau//sub E/ is up to two times the ITER97L L-Mode scaling and exceeds the ITER98pby2 H-Mode scaling in some cases. Transitions to H-Mode have been observed which result in an approximate doubling of /spl tau//sub E/ after the transition in some conditions. During HHFW heating, T/sub e/>T/sub i/ and T/sub e/ up to 3.5 keV were observed. Current drive has been studied using both coaxial helicity injection with up to 390 kA of toroidal current and HHFW. HHFW has produced H-modes with significant bootstrap current fraction at low I/sub p/, high q, and high /spl beta//sub p/.

Research on the stability of spherical torus plasmas at and above the no-wall beta limit is being addressed on the National Spherical Torus Experiment [M. Ono et al., Nucl. Fusion 40, 557 (2000)], that has produced low aspect ratio plasmas, R/a∼1.27 at plasma current exceeding 1.4 MA with high energy confinement (TauE/TauE_ITER89P>2). Toroidal and normalized beta have exceeded 25% and 4.3, respectively, in q∼7 plasmas. The beta limit is observed to increase and then saturate with increasing li. The stability factor βN/li has reached 6, limited by sudden beta collapses. Increased pressure peaking leads to a decrease in βN. Ideal stability analysis of equilibria reconstructed with EFIT [L. L. Lao et al., Nucl. Fusion 25, 1611 (1985)] shows that the plasmas are at the no-wall beta limit for the n=1 kink/ballooning mode. Low aspect ratio and high edge q theoretically alter the plasma stability and mode structure compared to standard tokamak configurations. Below the no-wall limit, stability calculations show the perturbed radial field is maximized near the center column and mode stability is not highly effected by a nearby conducting wall due to the short poloidal wavelength in this region. In contrast, as beta reaches and exceeds the no-wall limit, the mode becomes strongly ballooning with long poloidal wavelength at large major radius and is highly wall stabilized. In this way, wall stabilization is more effective at higher beta in low aspect ratio geometry. The resistive wall mode has been observed in plasmas exceeding the ideal no-wall beta limit and leads to rapid toroidal rotation damping across the plasma core.

For the first time, data from eight different tokamaks have been combined in an international database for internal transport barriers (ITBs). An analysis of the data for the formation of an ITB with dominant ion heating shows a clear dependence of the threshold power on the minor radius and line-averaged electron density for the formation of ion ITBs. The dependence of ITB formation on the toroidal magnetic field is weak. For the formation of ITBs with dominant electron heating, the database is smaller, but for the threshold power a strong increase with plasma size and a weak toroidal field dependence could also be identified. Based on these results, an expression for the power required to form an ITB is given using global variables only. These results give a basis for the analysis of the database using local values (like magnetic shear) and a detailed comparison with theory-based models.

The dynamic evolution of the first National Spherical Torus Experiment (NSTX) H-modes will be discussed. The H-modes were obtained in lower-single null divertor discharges with various forms of plasma heating. The exact timing of divertor formation and also the NBI power level affects both whether or not the discharge exhibits ELMs and the duration of the ELM-free phase. ELM-free discharges had energy confinement as high as 120 ms, whereas the few discharges with ELMs had confinement times ∼50-70 ms. Buildup of a steep edge density gradient and formation of ears' on the density profile were observed within a few ms of the L-H transition, yielding broader density and pressure profiles. The L-H transition was marked by a decrease in edge visible light and simultaneous increase in electron Bernstein wave emission, reflecting a steepening of the edge density gradient. Gas puff imaging of He-i light during the H-mode phase showed rapid formation of a narrow emission layer ∼2 cm wide in the H-mqde phase, which returned within 20 microseconds at termination of the H-mode phase to a broader turbulent emission layer. All of the H-modes were terminated by an MHD reconnection event. The first power threshold (P th ) study showed the neutral beam injection power (P b ) component of P th to be ≤0.84 MW, higher than that predicted by the ITER database scaling.

The quiescent double barrier (QDB) regime is a high performance regime recently identified in DIII-D and characterized by a double transport barrier structure (core and edge) that can be maintained for several seconds, often limited only by the pulse length capabilities of the DIII-D hardware. The QDB regime has been sustained for up to 25 τE with fusion performance of up to βNH89≈7. The edge barrier is ELM-free, but modulated by low frequency MHD activity that allows density control via an external cryopump. The core barrier is similar to those seen in previous internal transport barrier experiments, but is maintained without complete stabilization of turbulence. Instead, the turbulence correlation lengths become very short so as to minimize the transport length scales. The two barriers are separated by a region of high transport that is a consequence of a zero-crossing in the E×B shearing rate. These discharges typically possess highly peaked density profiles. This has several implications: narrow bootstrap current profile, reduced beta limit and increased impurity retention. We will report on studies of each of these issues.

Rapidly developing diagnostic, operational, and analysis capability is enabling the first detailed local physics studies to begin in high β plasmas of the National Spherical Torus Experiment (NSTX). These studies are motivated in part by the observation of energy confinement times in neutral-beam-heated discharges that are favourable with respect to predictions from the ITER-89P scaling expression. For plasmas heated with neutral beam injection (NBI), analysis based on profile measurements suggests that electron heat conduction is the dominant thermal loss channel. Cases where early analysis indicates that ion thermal conduction may be exceptionally low is motivating studies of possible sources of ion heating not presently accounted for by classical collisional processes. Gyrokinetic microstability studies indicate that long wavelength turbulence with kθρi~0.1-1 may be stable or suppressed by E×B shear in these plasmas, while modes with kθρi~50 may be robust. High harmonic fast wave (HHFW) heating efficiently heats electrons on NSTX, and studies have begun using it to assess transport in the electron channel. Regarding edge transport, H-mode transitions occur with either NBI or HHFW heating. The power required for L- to H-mode transitions far exceeds that expected from empirical ELM-free H-mode scaling laws derived from moderate aspect ratio devices. Finally, initial fluctuation measurements made with two techniques are permitting the first characterizations of edge turbulence.

Substantial progress has been made towards both understanding and control of internal transport barriers (ITBs) on DIII-D, resulting in the discovery of a new sustained high performance operating mode termed the quiescent double barrier (QDB) regime. The QDB regime combines core transport barriers with a quiescent ELM-free H mode edge (termed QH mode), giving rise to separate (double) core and edge transport barriers. The core and edge barriers are mutually compatible and do not merge, resulting in broad core profiles with an edge pedestal. The QH mode edge is characterized by ELM-free behaviour with continuous multiharmonic MHD activity in the pedestal region and has provided density and radiated power control for longer than 3.5 s (25τE) with divertor pumping. QDB plasmas are long pulse high performance candidates, having maintained a βN H89 product of 7 for five energy confinement times (Ti ≤ 16 keV, βN ≤ 2.9, H89 ≤ 2.4 τE ≤ 150 ms, DD neutron rate Sn ≤ 4 × 1015 s-1). The QDB regime has only been obtained in counter-NBI discharges (injection antiparallel to the plasma current) with divertor pumping. Other results include successful expansion of the ITB radius using (separately) both impurity injection and counter-NBI, and the formation of ITBs in the electron thermal channel using both ECH and strong negative central shear (NCS) at high power. These results are interpreted within a theoretical framework in which turbulence suppression is the key to ITB formation and control, and a decrease in core turbulence is observed in all cases of ITB formation.

A new method of actively modifying the plasma-wall interaction was tested on the Tokamak Fusion Test Reactor. A laser was used to introduce a directed lithium aerosol into the discharge scrape-off layer. The lithium introduced in this fashion ablated and migrated preferentially to the limiter contact points. This allowed the plasma-wall interaction to be influenced in situ and in real time by external means. Significant improvement in energy confinement and fusion neutron production rate as well as a reduction in the plasma Zeff have been documented in a neutral beam heated plasma. The introduction of a metallic aerosol into the plasma edge increased the internal inductance of the plasma column and also resulted in prompt heating of core electrons in ohmic plasmas. Preliminary evidence also suggests that the introduction of an aerosol leads to both edge poloidal velocity shear and edge electric field shear.

Experiments on the DIII-D tokamak have identified a new sustained high-performance operating mode, termed the quiescent double barrier (QDB) regime. The QDB regime combines internal transport barriers (ITBs) with a quiescent, edge localized mode (ELM)-free H-mode edge, termed QH-mode, giving rise to separate core and edge transport barriers. These double barriers have been maintained for {>}3.5 s (~25τE), demonstrating a long-pulse, quasi-steady-state capability. The combination of core ITBs and edge H-mode temperature pedestals results in high-performance plasmas; a βN H89 product of 7 has been maintained for 10 τE, other peak (non-simultaneous) parameters include Ti≤17 keV, βN≤2.9% m T MA-1, H89≤2.6, β≤3.8%, τE≤ 160 ms, and DD neutron rate Sn≤5.5×1015 s-1. These results address a major issue with tokamak plasmas: how to sustain long-pulse, high-performance H-mode plasmas without ELMs, yet retaining the density and impurity control hitherto provided by ELMs. In these QDB plasmas ELMs are replaced by continuous benign MHD activity in the edge, which enhances particle transport. A signature of operation with a QH-mode edge appears to be very large radial electric fields in the edge and scrape-off layer (SOL). In the core, simulations and modelling replicate many of the features of the observed transport and fluctuation behaviour, including the ion temperature profile and turbulence correlation lengths. Slow high-Z impurity accumulation (τ≥500 ms) is observed in the centre of many QDB plasmas, and is the subject of ongoing analysis. To date the QDB regime has only been obtained in plasmas with counter-NBI (injection anti-parallel to the plasma current), and with divertor cryopumping to control the density.

The main aim of the National Spherical Torus Experiment (NSTX) is to establish the fusion physics principles of the spherical torus (ST) concept. The NSTX device began plasma operations in February 1999 and the plasma current Ip was successfully brought up to the design value of 1 MA on 14 December 1999. The planned plasma shaping parameters, elongation κ = 1.6-2.2 and triangularity δ = 0.2-0.4, were achieved in inner wall limited, and single null and double null diverted configurations. The coaxial helicity injection (CHI) and high harmonic fast wave (HHFW) experiments were also initiated. CHI current of 27 kA produced up to 260 kA toroidal current without using an ohmic solenoid. With the injection of 2.3 MW of HHFW power, using 12 antennas connected to six transmitters, electrons were heated from a central temperature of 400 eV to 900 eV at a central density of 3.5 × 1013 cm-3, increasing the plasma energy to 59 kJ and the toroidal β, βT, to 10%. The NBI system commenced operation in September 2000. The initial results with two ion sources (PNBI = 2.8 MW) show good heating, producing a total plasma stored energy of 90 kJ corresponding to βT ≈ 18% at a plasma current of 1.1 MA.

A new sustained high-performance regime, combining discrete edge and core transport barriers, has been discovered in the DIII-D tokamak. Edge localized modes (ELMs) are replaced by a steady oscillation that increases edge particle transport, thereby allowing particle control with no ELM-induced pulsed divertor heat load. The core barrier resembles those usually seen with a low (L) mode edge, without the degradation often associated with ELMs. The barriers are separated by a narrow region of high transport associated with a zero crossing in the E x B shearing rate.

High-confinement (H-mode) operation is the choice for next-step tokamak devices based either on conventional or advanced tokamak physics. This choice, however, comes at a significant cost for both the conventional and advanced tokamaks because of the effects of edge localized modes (ELMs). ELMs can produce significant erosion in the divertor and can affect the beta limit and reduced core transport regions needed for advanced tokamak operation. Experimental results from DIII-D [J. L. Luxon et al., Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] this year have demonstrated a new operating regime, the quiescent H-mode regime, which solves these problems. We have achieved quiescent H-mode operation that is ELM-free and yet has good density and impurity control. In addition, we have demonstrated that an internal transport barrier can be produced and maintained inside the H-mode edge barrier for long periods of time (>3.5 s or >25 en...

The capability to inject deuterium pellets from the magnetic high field side (HFS) has been added to the DIII-D tokamak [J. L. Luxon and L. G. Davis, Fusion Technol. 8, 441 (1985)]. It is observed that pellets injected from the HFS lead to deeper mass deposition than identical pellets injected from the outside midplane, in spite of a factor of 4 lower pellet speed. HFS injected pellets have been used to generate peaked density profile plasmas [peaking factor (ne(0)/〈ne〉) in excess of 3] that develop internal transport barriers when centrally heated with neutral beam injection. The transport barriers are formed in conditions where Te∼Ti and q(0) is above unity. The peaked density profiles, characteristic of the internal transport barrier, persist for several energy confinement times. The pellets are also used to investigate transport barrier physics and modify plasma edge conditions. Transitions from L- to H-mode have been triggered by pellets, effectively lowering the H-mode threshold power by 2.4 MW. Pellets injected into H-mode plasmas are found to trigger edge localized modes (ELMs). ELMs triggered from the low field side (LFS) outside midplane injected pellets are of significantly longer duration than from HFS injected pellets.

We study the characteristics of self-generated zonal flows as observed in nonlinear global gyrokinetic simulations of toroidal ion temperature gradient (ITG) turbulence for typical parameters of DIII-D core plasmas. In addition, we discuss various possibilities for experimental measurements and the development of new diagnostics.

Transport phenomena are studied in Advanced Tokamak (AT) regimes in the DIII-D tokamak [Plasma Physics and Controlled Nuclear Fusion Research, 1986 (International Atomics Energy Agency, Vienna, 1987), Vol. I, p. 159], with the goal of developing understanding and control during each of three phases: Formation of the internal transport barrier (ITB) with counter neutral beam injection taking place when the heating power exceeds a threshold value of about 9 MW, contrasting to co-NBI injection, where Pthreshold

The National Spherical Torus Experiment (NSTX) is being built at the Princeton Plasma Physics Laboratory to test the fusion physics principles for the Spherical Torus (ST) concept at the MA level. The NSTX nominal plasma parameters are R {sub 0} = 85 cm, a = 67 cm, R/a greater than or equal to 1.26, B {sub T} = 3 kG, I {sub p} = 1 MA, q {sub 95} = 14, elongation {kappa} less than or equal to 2.2, triangularity {delta} less than or equal to 0.5, and plasma pulse length of up to 5 sec. The plasma heating/current drive (CD) tools are High Harmonic Fast Wave (HHFW) (6 MW, 5 sec), Neutral Beam Injection (NBI) (5 MW, 80 keV, 5 sec), and Coaxial Helicity Injection (CHI). Theoretical calculations predict that NSTX should provide exciting possibilities for exploring a number of important new physics regimes including very high plasma beta, naturally high plasma elongation, high bootstrap current fraction, absolute magnetic well, and high pressure driven sheared flow. In addition, the NSTX program plans to explore fully noninductive plasma start-up, as well as a dispersive scrape-off layer for heat and particle flux handling.

In fast wave to ion Bernstein wave mode conversion experiments in DT supershot plasmas, localized efficient ion heating rather than electron heating was observed, which was due to Doppler broadened tritium cyclotron resonance overlap into the mode conversion region. The ion temperature heat pulse associated with RF power modulation in this regime could provide a diagnostic tool for measuring the local ion thermal conductivity in various confinement regimes. In direct launch ion Bernstein wave heating experiments, core power coupling was limited by the excitation of parasitic edge modes. However, a sheared poloidal flow was observed that is consistent in both magnitude and direction with theoretical models based on RF driven Reynolds stress. With the modest power coupled to the core (~360 kW), the magnitude of the shear in the observed flow was estimated to be a factor of 3-4 too low to trigger transport barrier formation through localized shear suppression of turbulence.

Evidence is presented for the role of radial electric field shear in enhanced confinement regimes attained without sharp bifurcations or transitions. Temperature scans at constant density, created in the reheat phase following deuterium pellet injection into supershot plasmas in the Tokamak Fusion Test Reactor [J. D. Strachan, et al., Phys. Rev. Lett. 58, 1004 (1987)] are simulated using a physics-based transport model. The slow reheat of the ion temperature profile, during which the temperature nearly doubles, is not explained by relatively comprehensive models of transport due to Ion Temperature Gradient Driven Turbulence (ITGDT), which depends primarily on the (unchanging) electron density gradient. An extended model, including the suppression of toroidal ITGDT by self-consistent radial electric field shear, does reproduce the reheat phase. The extended reheat at constant density is observed in supershot but not L-Mode plasmas.

The article reports results of experiments to further determine the underlying physics behind the formation and development of internal transport barriers (ITBs) in the DIII-D tokamak. The initial ITB formation occurs when the neutral beam heating power exceeds a threshold value during the early stages of the current ramp in low density discharges. This region of reduced transport, made accessible by suppression of long wavelength turbulence by sheared flows, is most evident in the ion temperature and impurity rotation profiles. In some cases, reduced transport is also observed in the electron temperature and density profiles. If the power is near the threshold, the barrier remains stationary and encloses only a small fraction of the plasma volume. If, however, the power is increased, the transport barrier expands to encompass a larger fraction of the plasma volume. The dynamic behaviour of the transport barrier during the growth phase exhibits rapid transport events that are associated with both broadening of the profiles and reductions in turbulence and associated transport. In some but not all cases, these events are correlated with the safety factor q passing through integer values. The final state following this evolution is a plasma exhibiting ion thermal transport at or below neoclassical levels. Typically the electron thermal transport remains anomalously high. Recent experimental results are reported in which RF electron heating was applied to plasmas with an ion ITB, thereby increasing both the electron and the ion transport. Although the results are partially in agreement with the usual E × B shear suppression hypothesis, the results still leave questions that must be addressed in future experiments.

To study the applicability of artificially enhanced impurity radiation for mitigation of the plasma-limiter interaction in reactor regimes, krypton and xenon gases were injected into TFTR supershots and high internal inductance plasmas. At NBI powers PB ≥ 30 MW, carbon influxes (`blooms') were suppressed, leading to improved energy confinement and neutron production in both D and DT plasmas, and the highest DT fusion energy production (7.6 MJ) in a TFTR pulse. Comparisons of the measured radiated power profiles with predictions of the MIST impurity transport code have guided studies of highly radiative plasmas in ITER. The response of the electron and ion temperatures to greatly increased radiative losses from the electrons was used to study thermal transport mechanisms. A change in the radial electric field Er is associated with the improved confinement observed.

Perturbative experiments on TFTR have investigated the transport of multiple ion species in reverse shear (RS) plasmas. The profile evolutions of trace tritium and helium and intrinsic carbon indicate the formation of core particle transport barriers in enhanced reverse shear (ERS) plasmas. There is an order of magnitude reduction in the particle diffusivity inside the RS region. The diffusivities for these species in ERS plasmas agree with neoclassical theory.

The plasma dynamics of enhanced confinement regimes in the TFTR core and the DIII-D core and edge are compared in order to identify a common physics basis. Despite differences in transition timescale and location, as well as the sign of the radial electric field Er, observations suggest that E × B shear effects on turbulence induced transport play a dominant role in governing barrier dynamics in all cases. Fast confinement bifurcations are observed in the TFTR core enhanced reverse shear (ERS) regime and in the edge DIII-D H mode. Both show spontaneous Er shear layer formation prior to the confinement change and a negative Er well that persists as steep gradients form. These dynamics differ from those of DIII-D negative central shear (NCS) plasmas. There, slow transitions are observed when the applied torque from unidirectional beam injection is small, while faster development and more dramatic confinement improvements occur at higher applied torques. Unlike the H mode and ERS cases, the NCS core generally has a positive Er hill and no strong Er shear precursor. However, similarity experiments performed on TFTR indicate that ERS, L mode and NCS-like regimes can all be accessed in a continuous fashion by varying the E × B shear through changes in the applied torque at constant power. As in the DIII-D NCS case, core confinement in TFTR reverse shear plasmas improves slowly as co-rotation begins to dominate the determination of Er, no strong Er shear layer develops prior to that improvement, and the plasma possesses a positive Er hill. Reductions in transport with Er gradients of either sign are consistent with the picture of E × B shear suppression and decorrelation of turbulence. At fixed input power, intermediate levels of confinement improvement are achieved by varying the E × B shear with changes in the applied neutral beam torque. The data suggest that control over the plasma pressure profile in a reactor may be possible if an external source of E × B shear, such as might be applied with RF techniques, is used to modify the shear which otherwise occurs.

Improving confinement and β limits simultaneously in long pulse ELMy H mode discharges is investigated. The product βN H98y serves as a useful figure of merit for performance, where βN ≡ β/(I/aB) and H98y is the ratio of the thermal confinement time to the most recent ELMy H mode confinement scaling established by the ITER confinement database working group. In discharges with q0 ≈ 1 (no sawteeth) and discharges with qmin > 1.5 and negative central magnetic shear, βN ≈ 2.9 and H98y ≈ 1.4 are sustained for up to 2 s. Although peaked profiles are observed, steep internal transport barriers are not present. Further increases in βN in these discharges are limited by neoclassical tearing modes (NTMs) in the positive shear region. In another recently developed regime, βN ≈ 3.8 and H98y ≈ 2 have been sustained during large infrequent ELMs in non-sawtoothing discharges with q0 ≈ 1. This level of performance is similar to that obtained in ELM free regimes such as VH mode. The limitation on βN and pulse length in these discharges is also the onset of NTMs.

A case for substantial loss of fast ions degrading the performance of tokamak fusion test reactor plasmas [Phys. Plasmas 2, 2176 (1995)] with reversed magnetic shear (RS) is presented. The principal evidence is obtained from an experiment with short (40–70 ms) tritium beam pulses injected into deuterium beam heated RS plasmas [Phys. Rev. Lett. 82, 924 (1999)]. Modeling of this experiment indicates that up to 40% beam power is lost on a time scale much shorter than the beam–ion slowing down time. Critical parameters which connect modeling and experiment are: The total 14 MeV neutron emission, its radial profile, and the transverse stored energy. The fusion performance of some plasmas with internal transport barriers is further deteriorated by impurity accumulation in the plasma core.

The mission of the National Spherical Torus Experiment (NSTX) is to prove the principles of spherical torus physics by producing high-beta toroidal plasmas that are non-inductively sustained, and whose current profiles are in steady-state. NSTX will be one of the first ultra low a[P(input) up to 11 MW] in order to produce high-beta toroidal (25 to 40%), low collisionality, high bootstrap fraction (less than or equal to 70%) discharges. Both radio-frequency (RF) and neutral-beam (NB) heating and current drive will be employed. Built into NSTX is sufficient configurational flexibility to study a range of operating space and the resulting dependences of the confinement, micro- and MHD stability, and particle and power handling properties. NSTX research will be carried out by a nationally based science team.

Anomalous beam-ion loss has been observed in an experiment with short tritium beam pulses injected into deuterium-beam-heated Tokamak Fusion Test Reactor plasmas (P{sub NBI}=15 thinspthinspMW) with reversed magnetic shear (RS). Comparisons of the measured total 14thinspthinspMeV neutron emission, the neutron flux along eight radial locations, and the perpendicular plasma stored energy with predictions from an extensive set of TRANSP simulations suggest that about 40{percent} beam power is lost on a time scale much shorter than the tritium beam pulse length {Delta}t=70 thinspthinspms. In contrast with recent results [K. Tobita {ital et al.,} Nucl.thinspthinspFusion {bold 37}, 1583 (1997)] from RS experiments at JT-60U, we were not able to show conclusively that magnetic field ripple is responsible for this anomaly. {copyright} {ital 1999} {ital The American Physical Society}

Spontaneous improvements of plasma confinement during auxiliary heating have been observed in many tokamaks when the q profile has been modified from its normal resistive equilibrium so that q is greater than 1 and the magnetic shear is reduced or reversed in a region near the magnetic axis. The effects on the overall plasma confinement result from the formation in the plasma interior of transport barriers, regions where the thermal and particle transport coefficients are substantially reduced. These internal barriers are sometimes tied to unique magnetic surfaces, such as the surface where the shear reverses. The reduction in transport appears to result from the suppression of turbulence by sheared plasma flow, which has now been measured in TFTR. Extensions of the theory for turbulence suppression show that this underlying paradigm may also explain other regimes of improved core confinement. The excitement generated by these discoveries must be tempered by the realization that transport and stability to pressure-driven MHD instabilities are intimately linked in these plasmas through the bootstrap current and the effect of the resulting current profile on the transport. Thus the development of control tools and strategies is essential if these improved regimes of confinement are to be exploited to improve the prospects for fusion energy production.

A bifurcation in the core poloidal rotation of carbon impurity ions in the Tokamak Fusion Test Reactor (TFTR) has been observed prior to the transport bifurcation associated with enhanced reverse shear plasmas. In a narrow radial region of the plasma, the impurity ion poloidal rotation reverses direction. This poloidal flow is associated with the establishment of a large negative radial electric field with strong shear. The measured poloidal velocities before, during, and after this precursor differ from neoclassical predictions.

Measurements of the radial electric field have been obtained on the Tokamak Fusion Test Reactor utilizing the motional Stark effect diagnostic. A large negative excursion in the radial electric field occurs before the transition to the enhanced reversed magnetic shear mode. The electric field is localized to a narrow spatial region and is not observed in discharges without the transition to improved confinement. Concomitant with the radial electric field excursion is a change in the measured impurity poloidal rotation velocity and bursts of magnetic fluctuations. {copyright} {ital 1998} {ital The American Physical Society}

Perturbative experiments on the Tokamak Fusion Test Reactor [Phys. Plasmas 4, 1736 (1997)] (TFTR) have investigated transport in reverse shear plasmas. On TFTR, reverse magnetic shear plasmas bifurcate into two states with different transport properties: reverse shear (RS) and enhanced reverse shear (ERS) with improved core confinement. Measurements of the 14 MeV t(d,n)α neutrons and charge-exchange recombination radiation spectra are used to infer the trace tritium and helium profiles, respectively. The profile evolution indicate the formation of core particle transport barriers in ERS plasmas. The transport barrier is manifested by an order-of-magnitude reduction in the particle diffusivity (DT,DHe) and a smaller reduction in the pinch within the reverse shear region. The low diffusivities are consistent with neoclassical predictions. Furthermore, DT and DHe≈χeff, the effective thermal diffusivity. Although the measured coefficients imply no helium ash accumulation, the situation is uncertain in a react...

The phenomenology of transport barrier formation is reviewed with a focus on physics that may be common to the edge and the core. To this end, the framework of E x B velocity shear reduction of turbulence-induced fluxes, applied to the edge for some time, is studied in light of measurements of core bifurcation dynamics and recent tests of causality. Also, the possible role of the magnetic shear structure in facilitating core barrier formation is examined. Experimental and theoretical challenges for developing predictive capability for reactor-relevant conditions are highlighted by recent observations of spontaneous electric field shear generation far removed from edge effects, and efforts to characterize the plasma edge at and across the L-H transition.

Impurity poloidal rotation velocities have been measured in the core of TFTR plasmas using a new spectroscopic diagnostic. Two types of transition to enhanced confinement in reversed shear plasmas are examined. A bifurcation in carbon poloidal rotation is observed to occur before the transition to enhanced confinement for one of these types, while other measured plasma parameters remain constant. A narrow radial region with reversed poloidal rotation and rotational shear is established 60-100 ms before the transition and is associated with a large negative radial electric field.

Tritium and helium gas-puff experiments have been conducted on TFTR to investigate transport in reverse shear plasmas. Small amounts of tritium and helium gas are puffed into the equilibrium phase of these plasmas. Measurements of the 14 MeV neutron fusion product and the charge-exchange spectrum of helium reveal the transport details of tritium and helium, respectively. The inferred tritium and helium density profiles indicate the formation of a core transport barrier for each species in reverse shear plasmas with enhanced confinement.

A large “notch,” or non-monotonic feature, appears in measured toroidal velocity profiles of the carbon impurity in the Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Fusion 26, 11 (1984)], centered near the radius of strongest ion temperature gradient. This is explained as a consequence of radial momentum transport dominated by anomalous diffusion together with parallel heat friction on the impurity ions arising from the hydrogenic neoclassical parallel heat flow. The toroidal velocity profile of the hydrogenic species is predicted to be monotonic, from measurements of the impurity toroidal velocity, consistent with the anomalous radial diffusion of toroidal momentum. This supports a neoclassical calculation of the radial electric field for near-balanced beam injection. In supershot plasmas [Phys. Rev. Lett. 58, 1004 (1987)], a well structure in the radial electric field profile is found in the enhanced confinement region. An associated shear layer separates the core, where the local confine...

Short wavelength fluctuations with k ∼ ω pe c and kϱi ∼ 5 are detected by microwave scattering in TFTR plasmas with reversed magnetic shear. They propagate in the ion diamagnetic drift direction with a frequency below the ion diamagnetic drift frequency in the frame of reference where Er = 0. The variation of the fluctuation amplitude in the plasma core correlates with the variation of the local electron heat conductivity, suggesting that these fluctuations may be the cause of anomalous electron heat transport in these plasmas.

The scaling of the electron thermal diffusivity, {chi}{sub e}, with relative gyro radius, {rho}*, has been measured on TFTR by comparing nearly identical ICRF-heated discharges which differ only in hydrogenic isotopic content. Contrary to the gyro-Bohm scaling ({chi}{sub e} {approximately} {chi}{sub B}{rho}*, where {chi}{sub B} is the Bohm diffusivity) observed on DIII-D when {rho}* was varied through a scan of magnetic field strength, {chi}{sub e} is found to scale inversely with {rho}*. Hence, global energy confinement is 8--11% higher in deuterium-tritium plasmas than in deuterium only plasmas, with the higher stored energy attributed almost entirely to the electrons.

The roles of turbulence stabilization by sheared E×B flow and Shafranov shift gradients are examined for Tokamak Fusion Test Reactor [D. J. Grove and D. M. Meade, Nucl. Fusion 25, 1167 (1985)] enhanced reverse-shear (ERS) plasmas. Both effects in combination provide the basis of a positive-feedback model that predicts reinforced turbulence suppression with increasing pressure gradient. Local fluctuation behavior at the onset of ERS confinement is consistent with this framework. The power required for transitions into the ERS regime are lower when high power neutral beams are applied earlier in the current profile evolution, consistent with the suggestion that both effects play a role. Separation of the roles of E×B and Shafranov shift effects was performed by varying the E×B shear through changes in the toroidal velocity with nearly steady-state pressure profiles. Transport and fluctuation levels increase only when E×B shearing rates are driven below a critical value that is comparable to the fastest line...