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Fast 2-D cameras examine a variety of important aspects of the plasma edge and in-vessel components on the National Spherical Torus Experiment (NSTX). Four Phantom and two Miro visible-light cameras manufactured by Vision Research are used on NSTX for edge studies. Each camera can take several gigabytes (GBs) of data during each plasma pulse. Timely access to this amount of data can itself be a challenge, but analysing all these data using manual frame-by-frame examination is not practical. This paper describes image analysis, database techniques, and visualization methods used to organize the fast camera data and to facilitate physics insights from it. An example is presented of analysing and characterizing the size, movement and dynamics of coherent plasma structures (typically referred to as “blobs”) near the plasma edge. Software tools that generate statistics of blob speed, shape, amplitude, size, and orientation are described. The characteristics of emitted blobs affect plasma confinement and heat loads on plasma facing components, and are thus of particular interest to future machines like ITER.

NSTX experiments have explored lithium evaporated on a graphite divertor and other plasma-facing components in both L- and H- mode confinement regimes heated by high-power neutral beams. Improvements in plasma performance have followed these lithium depositions, including a reduction and eventual elimination of the HeGDC time between discharges, reduced edge neutral density, reduced plasma density, particularly in the edge and the SOL, increased pedestal electron and ion temperature, improved energy confinement and the suppression of ELMs in the H-mode. However, with improvements in confinement and suppression of ELMs, there was a significant secular increase in the effective ion charge Zeff and the radiated power in H-mode plasmas as a result of increases in the carbon and medium-Z metallic impurities. Lithium itself remained at a very low level in the plasma core

A gas puff imaging diagnostic is used in the National Spherical Tokamak Experiment [M. Ono, et al., Nucl. Fusion 40, 557 (2000)] to study the edge turbulence and intermittency present during H-mode discharges. In the case of low power Ohmic H-modes the suppression of turbulence/blobs is maintained through the duration of the (short lived) H-modes. Similar quiescent edges are seen during the early stages of H-modes created with the use of neutral beam injection. Nevertheless, as time progresses following the L-H transition, turbulence and blobs reappear although at a lower level than that typically seen during L-mode confinement. It is also seen that the time-averaged SOL emission profile broadens, as the power loss across the separatrix increases. These broad profiles are characterized by a large level of fluctuations and intermittent events.

Steady-state handling of divertor heat flux is a critical issue for present and future conventional and spherical tokamaks with compact high power density divertors. A novel “snowflake” divertor (SFD) configuration that takes advantage of magnetic properties of a second-order poloidal null has been predicted to have a larger plasma-wetted area and a larger divertor volume, in comparison with a standard first-order poloidal X-point divertor configuration. The SFD was obtained in 0.8 MA, 4–6 MW NBI-heated H-mode discharges in NSTX using two divertor magnetic coils. The SFD led to a partial detachment of the outer strike point even in low-collisionality scrape-off layer plasma obtained with lithium coatings in NSTX. Significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with good core confinement and MHD properties.

Calibrated amounts of 40 μm lithium dust and 10 μm tungsten powder have been dropped from above into the SOL of the National Spherical Torus Experiment (NSTX) to benchmark modeling of dust dynamics and transport. By combining the output from two visible-range fast cameras, 3-D trajectories are reliably obtained and have resulted in the generation of several hundred individual particle tracks. Particles are observed to undergo a variety of accelerations both parallel and perpendicular to the magnetic field, as well as abrupt large-angle changes in direction. All tracks obtained to date display particle motion that is constrained to within a few centimeters of the last closed flux surface. The 3-D trajectories are presented and compared to the location of the last closed flux surface as determined by EFIT.

The two-dimensional fluid turbulence code SOLT is employed to study the role of midplane turbulence on the scrape-off-layer (SOL) heat flux width of tokamak plasmas. The physics simulated includes curvature-driven-interchange modes, sheath losses, and perpendicular turbulent diffusive and convective (blob) transport. Midplane SOL profiles of density, temperature and parallel heat flux are obtained from the simulation and compared with experimental results from the National Spherical Torus Experiment (NSTX) to study the scaling of the heat flux width with power and plasma current. It is concluded that midplane turbulence is the main contributor to the SOL width for the low power ELM-free H-mode discharges studied, while additional physics is required to explain the plasma current scaling of the SOL width observed experimentally in higher power discharges. Additional simulations predict a transition to a convectively-dominated SOL at critical values of power and connection length.

Major developments on the Mega Amp Spherical Tokamak (MAST) have enabled important advances in support of ITER and the physics basis of a spherical tokamak (ST) based component test facility (CTF), as well as providing new insight into underlying tokamak physics. For example, L–H transition studies benefit from high spatial and temporal resolution measurements of pedestal profile evolution (temperature, density and radial electric field) and in support of pedestal stability studies the edge current density profile has been inferred from motional Stark effect measurements. The influence of the q-profile and E × B flow shear on transport has been studied in MAST and equilibrium flow shear has been included in gyro-kinetic codes, improving comparisons with the experimental data. H-modes exhibit a weaker q and stronger collisionality dependence of heat diffusivity than implied by IPB98(y,2) scaling, which may have important implications for the design of an ST-based CTF. ELM mitigation, an important issue for ITER, has been demonstrated by applying resonant magnetic perturbations (RMPs) using both internal and external coils, but full stabilization of type-I ELMs has not been observed. Modelling shows the importance of including the plasma response to the RMP fields. MAST plasmas with q > 1 and weak central magnetic shear regularly exhibit a long-lived saturated ideal internal mode. Measured plasma braking in the presence of this mode compares well with neo-classical toroidal viscosity theory. In support of basic physics understanding, high resolution Thomson scattering measurements are providing new insight into sawtooth crash dynamics and neo-classical tearing mode critical island widths. Retarding field analyser measurements show elevated ion temperatures in the scrape-off layer of L-mode plasmas and, in the presence of type-I ELMs, ions with energy greater than 500 eV are detected 20 cm outside the separatrix. Disruption mitigation by massive gas injection has reduced divertor heat loads by up to 70%.

Experiments in NSTX have now unambiguously demonstrated the coupling of toroidal plasmas produced by the method of transient Coaxial Helicity Injection (CHI) to inductive sustainment and ramp-up of the toroidal current. The coupled discharges have ramped up to 700 kA and transitioned into H-mode with low inductance typical of the type of discharges needed for long-pulse operation, demonstrating the compatibility of the CHI startup method to conventional inductive operation used since the start of tokamak research. The method was first demonstrated on the smaller concept exploration device HIT-II at the University of Washington. These new results that were obtained on a machine built with mainly conventional components and on a size scale closer to a Component Test Facility, demonstrate that CHI is a viable solenoid-free plasma startup method for future STs and Tokamaks.

Filamentary structures are observed in the scrape-off layer of the National Spherical Torus Experiment during ELMs. While the primary filaments correspond to a direct result of the ELM event, the ‘secondary’ filaments which occur generally later but still within 1 ms of the ELM onset are observed to have the same characteristics as inter-ELM filaments (or blobs): poloidal auto-correlation lengths of ∼4 cm, broadband frequency and poloidal wave number spectra and radial velocities of 1–2 km/s. At the same time, no MHD modes are observed during the phase in which secondary filaments are present.

Small electrodes were installed in the outer-midplane edge of NSTX to attempt to control the local width of the scrape-off layer (SOL) by creating an outward E(pol)xB flow. When the applied voltage between electrodes was 90 V, the density between these electrodes increased by a factor of 3-10 over a radial width of similar to 4 cm. Thus a local control of the SOL plasma density was obtained. (C) 2009 Elsevier B.V. All rights reserved.

Coaxial Helicity Injection (CHI) has been successfully used in the National Spherical Torus Experiment (NSTX) for a demonstration of closed flux current generation without the use of the central solenoid. The favorable properties of the Spherical Torus (ST) arise from its very small aspect ratio. However, small aspect ratio devices have very restricted space for a substantial central solenoid. Thus methods for initiating the plasma current without relying on induction from a central solenoid are essential for the viability of the ST concept. CHI is a promising candidate for solenoid-free plasma startup in a ST. The method has now produced closed flux current up to 160 kA verifying the high current capability of this method in a large ST built with conventional tokamak components.

The transport of dust particles in plasmas may play a significant role in the performance of next step fusion devices. Highly mobile incandescent dust particles are observed on NSTX for the majority of the discharges using fast visible cameras. Particles are most often born in the divertor region during events such as ELMs or disruptions. Particles born on the midplane are most often deflected by the plasma boundary and remain outside the scrape off layer. The dynamics of the dust trajectories can be quite complex exhibiting a large variation in both speed (10–200 m/s) and direction. Particles may have constant velocities or exhibit various degrees of acceleration or deceleration. Abrupt reversals in direction are sometimes observed while some of the larger particles are seen to break apart during mid-flight. 3D trajectories of the dust particles have been derived from measurements of dust trajectories taken simultaneously from two observations points with two fast cameras.

The fast-evolving structure of MARFEs and ELMs is observed in the National Spherical Torus Experiment (NSTX) using a fast-framing camera. It is shown that Type V ELMs, the small ELM regime in NSTX, is characterized by a poloidally propagating ionization front close to the separatrix position. The ELMs also interact with the main chamber MARFE observed in NSTX coupling its movement and partial burn-through with that of the ELM cycle. During this cycle a toroidally localized MARFE precursor is seen propagating upward along field lines at speed of the order of 15 km/s. This interaction between ELM activity and MARFEs may be responsible for slow heat pulse propagation speeds seen in the divertor region.

Animations from fast 2D camera data are facilitating the investigation of spatially distributed phenomena in high-temperature plasmas. The National Spherical Torus Experiment (NSTX) now has six fast camera systems, and more are expected to be added. Image capture rates vary between 1000 and 500,000 frames per second. Archiving and retrieving this data is a challenge for data repositories and networks. For example, if all camera data had been archived during the 2004 run, the total amount of data from NSTX (300MB per pulse) would have doubled, and this year, one new camera alone can acquire 2GB per pulse. The paper will describe the storage strategies, and compare some data compression techniques used for NSTX. Tools which animate camera data, synchronized with displays from other plasma diagnostics and simulations, allow scientists to gain insights and observe correlations that would be difficult with conventional tools. A labor-saving technique is described for archiving fast camera images from a vendor's system into MDSplus and to AVI files. Examples of specific analysis and display techniques are presented. Future challenges are also discussed.

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.

H-mode operation plays a crucial role in National Spherical Torus Experiment (NSTX) research, allowing higher beta limits due to reduced plasma pressure peaking, and long pulse operation due to high bootstrap current fraction. Here, new results are presented in the areas of edge localized modes (ELMs), H-mode pedestal physics and power threshold studies. ELMs of several types as reported by higher aspect ratio tokamaks have been observed: (1) large, Type I ELMs, (2) intermediate-sized Type III ELMs and (3) tiny ELMs. Many high performance discharges in NSTX have the tiny ELMs (recently termed Type V), which have some differences as compared with small-magnitude ELM types in the published literature. A divertor multifaceted axisymmetric radiation from the edge (MARFE) on the inboard leg provides an effective light source to examine the effect of the ELMs on the divertor plasma; it is clear that only the large ELMs burn through the MARFE. The time difference between observation of the ELM flux at the outer and inner targets is substantially longer for the smallest ELMs as compared with the large ELMs. In addition, the visible light patterns show 'finger-like' striations during the tiny ELMs. H-mode pedestal studies have commenced, with the observation that the pedestal contains between 25% and 33% of the total stored energy, and the NSTX pedestal energy agrees reasonably well with a recent international multi-machine scaling. A power threshold identity experiment between NSTX and the Mega-Amp Spherical Tokamak shows comparable loss power at the L?H transition in balanced double-null discharges. Both machines require more power for the L?H transition as the balance is shifted toward lower-single null. High-field side gas fuelling enables more reliable H-mode access in NSTX, but does not always lead to a lower power threshold, e.g. with a reduction of the duration of early heating.

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.

The two-dimensional radial vs poloidal structure and motion of edge turbulence in the National Spherical Torus Experiment (NSTX) were measured using high-speed imaging of the visible light emission from a localized neutral gas puff. Edge turbulence images are shown and analysed for Ohmic, L- and H-mode plasma conditions. The two-dimensional images often show regions of strong localized light emission known as 'blobs', which move both poloidally and radially at a typical speed of ?105?cm?s?1, and sometimes show spatially periodic features.

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...

Global magnetohydrodynamic (MHD) stability limits in the National Spherical Torus Experiment (NSTX) have increased significantly recently due to a combination of device and operational improvements. First, more routine H-mode operation with broadened pressure profiles allows access to higher normalized β and lower internal inductance. Second, the correction of a poloidal field coil induced error-field has largely eliminated locked tearing modes during normal operation and increased the maximum achievable β. As a result of these improvements, peak β values have reached (not simultaneously) βT = 35%, βN = 6.4, βN = 4.5, βN/li = 10, and βP = 1.4. High βP operation with reduced tearing activity has allowed a doubling of discharge pulse-length to just over 1 s with sustained periods of βN≈6 above the ideal no-wall limit and near the with-wall limit. Details of the β-limit scalings and β-limiting instabilities in various operating regimes are described.

The National Spherical Torus Experiment (NSTX) is currently being upgraded to operate at twice the toroidal field and plasma current (up to 1 T and 2 MA), with a second, more tangentially aimed neutral beam (NB) for current and rotation control, allowing for pulse lengths up to 5 s. Recent NSTX physics analyses have addressed topics that will allow NSTX-Upgrade to achieve the research goals critical to a Fusion Nuclear Science Facility. These include producing stable, 100% non-inductive operation in high-performance plasmas, assessing plasma–material interface (PMI) solutions to handle the high heat loads expected in the next-step devices and exploring the unique spherical torus (ST) parameter regimes to advance predictive capability. Non-inductive operation and current profile control in NSTX-U will be facilitated by co-axial helicity injection (CHI) as well as radio frequency (RF) and NB heating. CHI studies using NIMROD indicate that the reconnection process is consistent with the 2D Sweet–Parker theory. Full-wave AORSA simulations show that RF power losses in the scrape-off layer (SOL) increase significantly for both NSTX and NSTX-U when the launched waves propagate in the SOL. Toroidal Alfven eigenmode avalanches and higher frequency Alfven eigenmodes can affect NB-driven current through energy loss and redistribution of fast ions. The inclusion of rotation and kinetic resonances, which depend on collisionality, is necessary for predicting experimental stability thresholds of fast growing ideal wall and resistive wall modes. Neutral beams and neoclassical toroidal viscosity generated from applied 3D fields can be used as actuators to produce rotation profiles optimized for global stability. DEGAS-2 has been used to study the dependence of gas penetration on SOL temperatures and densities for the MGI system being implemented on the Upgrade for disruption mitigation. PMI studies have focused on the effect of ELMs and 3D fields on plasma detachment and heat flux handling. Simulations indicate that snowflake and impurity seeded radiative divertors are candidates for heat flux mitigation in NSTX-U. Studies of lithium evaporation on graphite surfaces indicate that lithium increases oxygen surface concentrations on graphite, and deuterium–oxygen affinity, which increases deuterium pumping and reduces recycling. In situ and test-stand experiments of lithiated graphite and molybdenum indicate temperature-enhanced sputtering, although that test-stand studies also show the potential for heat flux reduction through lithium vapour shielding. Non-linear gyro kinetic simulations have indicated that ion transport can be enhanced by a shear-flow instability, and that non-local effects are necessary to explain the observed rapid changes in plasma turbulence. Predictive simulations have shown agreement between a microtearing-based reduced transport model and the measured electron temperatures in a microtearing unstable regime. Two Alfven eigenmode-driven fast ion transport models have been developed and successfully benchmarked against NSTX data. Upgrade construction is moving on schedule with initial physics research operation of NSTX-U planned for mid-2015.

Author(s): Menard, JE; Allain, JP; Battaglia, DJ; Bedoya, F; Bell, RE; Belova, E; Berkery, JW; Boyer, MD; Crocker, N; Diallo, A; Ebrahimi, F; Ferraro, N; Fredrickson, E; Frerichs, H; Gerhardt, S; Gorelenkov, N; Guttenfelder, W; Heidbrink, W; Kaita, R; Kaye, SM; Kriete, DM; Kubota, S; Leblanc, BP; Liu, D; Lunsford, R; Mueller, D; Myers, CE; Ono, M; Park, JK; Podesta, M; Raman, R; Reinke, M; Ren, Y; Sabbagh, SA; Schmitz, O; Scotti, F; Sechrest, Y; Skinner, CH; Smith, DR; Soukhanovskii, V; Stoltzfus-Dueck, T; Yuh, H; Wang, Z; Waters, I; Ahn, JW; Andre, R; Barchfeld, R; Beiersdorfer, P; Bertelli, N; Bhattacharjee, A; Brennan, D; Buttery, R; Capece, A; Canal, G; Canik, J; Chang, CS; Darrow, D; Delgado-Aparicio, L; Domier, C; Ethier, S; Evans, T; Ferron, J; Finkenthal, M; Fonck, R; Gan, K; Gates, D; Goumiri, I; Gray, T; Hosea, J; Humphreys, D; Jarboe, T; Jardin, S; Jaworski, MA; Koel, B; Kolemen, E; Ku, S; La Haye, RJ; Levinton, F; Luhmann, N; Maingi, R; Maqueda, R; McKee, G; Meier, E; Myra, J; Perkins, R | Abstract: The National Spherical Torus Experiment (NSTX) has undergone a major upgrade, and the NSTX Upgrade (NSTX-U) Project was completed in the summer of 2015. NSTX-U first plasma was subsequently achieved, diagnostic and control systems have been commissioned, the H-mode accessed, magnetic error fields identified and mitigated, and the first physics research campaign carried out. During ten run weeks of operation, NSTX-U surpassed NSTX record pulse-durations and toroidal fields (TF), and high-performance ∼1 MA H-mode plasmas comparable to the best of NSTX have been sustained near and slightly above the n = 1 no-wall stability limit and with H-mode confinement multiplier H98y,2 above 1. Transport and turbulence studies in L-mode plasmas have identified the coexistence of at least two ion-gyro-scale turbulent micro-instabilities near the same radial location but propagating in opposite (i.e. ion and electron diamagnetic) directions. These modes have the characteristics of ion-temperature gradient and micro-tearing modes, respectively, and the role of these modes in contributing to thermal transport is under active investigation. The new second more tangential neutral beam injection was observed to significantly modify the stability of two types of Alfven eigenmodes. Improvements in offline disruption forecasting were made in the areas of identification of rotating MHD modes and other macroscopic instabilities using the disruption event characterization and forecasting code. Lastly, the materials analysis and particle probe was utilized on NSTX-U for the first time and enabled assessments of the correlation between boronized wall conditions and plasma performance. These and other highlights from the first run campaign of NSTX-U are described.

Gas puff imaging (GPI) is a diagnostic of plasma turbulence which uses a puff of neutral gas at the plasma edge to increase the local visible light emission for improved space-time resolution of plasma fluctuations. This paper reviews gas puff imaging diagnostics of edge plasma turbulence in magnetic fusion research, with a focus on the instrumentation, diagnostic cross-checks, and interpretation issues. The gas puff imaging hardware, optics, and detectors are described for about 10 GPI systems implemented over the past ∼15 years. Comparison of GPI results with other edge turbulence diagnostic results is described, and many common features are observed. Several issues in the interpretation of GPI measurements are discussed, and potential improvements in hardware and modeling are suggested.

Research in NSTX has been conducted to establish spherical torus plasmas to be used for high ?, auxiliary heated experiments. This device has a major radius R0 = 0.86?m and a midplane halfwidth of 0.7?m. It has been operated with toroidal magnetic field B0 ? 0.3?T and Ip ? 1.0?MA. The evolution of the plasma equilibrium is analysed between discharges with an automated version of the EFIT code. Limiter, double null and lower single null diverted configurations have been sustained for several energy confinement times. The plasma stored energy reached 92?kJ (?t = 17.8%) with neutral beam heating. A plasma elongation in the range 1.6 ? ? ? 2.0 and a triangularity in the range 0.25 ? ? ? 0.45 have been sustained, with values of ? = 2.6 and ? = 0.6 being reached transiently. The reconstructed magnetic signals are fitted to the corresponding measured values with low errors. Aspects of the plasma boundary, pressure and safety factor profiles are supported by measurements from non-magnetic diagnostics. Plasma densities have reached 0.8 and 1.2 times the Greenwald limit in deuterium and helium plasmas, respectively, with no clear limit encountered. Instabilities including sawteeth and reconnection events, characterized by Mirnov oscillations, and a perturbation of the Ip, ? and li evolutions, have been observed. A low q limit was observed and is imposed by a low toroidal mode number kink instability.

The use of fast 2D cameras on NSTX continues to grow. There are 6 cameras with the capability of taking up to 1–2 gigabytes (GBs) of data apiece during each plasma shot on the National Spherical Torus Experiment (NSTX). Efficient storage and retrieval of this data remains a challenge. Performance comparisons are presented for reading data stored in MDSplus, using both compressed data and segmented records, and direct access I/O with different read sizes. Encouragingly, fast 2D camera data provides considerable insight into plasma complexities, such as small-scale turbulence and particle transport. The last part of this paper is an example of more recent uses: dual cameras looking at the same region of the plasma from different angles, which can provide trajectories of incandescent particles in 3D. A laboratory simulation of the 3D trajectories is presented, as well as corresponding data from NSTX plasma where glowing dust particles can be followed.

Experiments in NSTX have now unambiguously demonstrated the coupling of toroidal plasmas produced by the technique of CHI to inductive sustainment and ramp-up of the toroidal plasma current. This is an important step because an alternate method for plasma startup is essential for developing a fusion reactor based on the spherical torus concept. Elimination of the central solenoid would also allow greater flexibility in the choice of the aspect ratio in tokamak designs now being considered. The transient CHI method for spherical torus startup was originally developed on the HIT-II experiment at the University of Washington.

Fast cameras already installed on the National Spherical Torus Experiment (NSTX) have be equipped with near-infrared (NIR) filters in order to measure the surface temperature in the lower divertor region. Such a system provides a unique combination of high speed (> 50 kHz) and wide fi eld-of-view (> 50% of the divertor). Benchtop calibrations demonstrated the system's ability to measure thermal emission down to 330 oC. There is also, however, signi cant plasma light background in NSTX. Without improvements in background reduction, the current system is incapable of measuring signals below the background equivalent temperature (600 - 700 oC). Thermal signatures have been detected in cases of extreme divertor heating. It is observed that the divertor can reach temperatures around 800 oC when high harmonic fast wave (HHFW) heating is used. These temperature profiles were fi t using a simple heat diffusion code, providing a measurement of the heat flux to the divertor. Comparisons to other infrared thermography systems on NSTX are made.

The Coaxial Slow Source (CSS) is a device in which 'annular' field reversed configurations (FRCs) (small aspect ratio, elongated, toroidal plasmas with poloidal field only) are formed in the space between coaxial coils carrying toroidal currents. The device is constructed so that the plasma can be translated into a simple cylindrical chamber and re-formed as a conventional FRC. Formation of FRCs on slow time-scales (50-250 mu s) at low loop voltage (87.5-700 V) has been demonstrated. The paper presents initial results from the new parallel configuration device CSSP in which the inner and outer coils are connected in parallel to the capacitor banks. The parallel coil arrangement has significantly reduced the contact of the plasma with the walls and thus the impurity content. Configurations with confinement times of 1o-30 mu s, densities of 3*1021 m-3 and temperatures of 3-20 eV are typical. Results are presented on formation, energy balance, a nondestructive tilt instability and dynamics associated with magnetic tearing

A lithium particle dropper was installed on NSTX during the 2008 campaign. Though the primary purpose of the dropper was to study the effects of Li wall conditioning, this experimental configuration also afforded a unique opportunity to study the interaction of dust of a known size and composition with a reactor grade plasma. Li powder was dropped into NSTX at a rate of ∼ 1–35 mg/sec attaining a velocity of ∼ 5 m/s at the plasma boundary. The individual particles rapidly become incandescent due to their interaction with the plasma. Two fast visible cameras, spatially separated by up to 60 degrees, simultaneously viewed the individual particles. Data from the two cameras was used to reconstruct the first detailed 3-D trajectory information of dust in a plasma with a known size, composition and velocity using a 3-D tracking code [1]. These data will be used to further constrain the comparisons to the dust transport codes such as DUSTT used to predict dust behavior in future fusion reactors. Although Li is not presently considered a reactor material, its mass and thus its behavior should be close to that of Be which is a first wall material on ITER.

The National Spherical Torus Experiment (NSTX) produces plasmas, with toroidal aspect ratio as low as 1.25 and plasma currents up to 1.5 MA, which can be heated by up to 6 MW High‐Harmonic Fast Waves and up to 7 MW of deuterium Neutral Beam Injection. With these capabilities, NSTX has already made considerable progress in advancing the scientific understanding of high performance plasmas needed for low‐aspect‐ratio reactor concepts and for ITER. In transport and turbulence research on NSTX, the role of magnetic shear is being elucidated in discharges in which electron energy transport barriers are observed. Scaling studies indicate a weaker dependence on plasma current than at conventional aspect ratio and a significant dependence on toroidal field (BT).

A linear array of four small biased electrodes was installed in NSTX in an attempt to control the width of the scrape-off layer by creating a strong local poloidal electric field. The set of electrodes was separated poloidally by a 1 cm gap between electrodes and were located slightly below the midplane of NSTX, 1 cm behind the rf antenna, and oriented so that each electrode is facing approximately normal to the magnetic field. Each electrode can be independently biased to +/-100 V. Present power supplies limit the current on two electrodes to 30 A and the other two to 10 A each. The effect of local biasing was measured with a set of Langmuir probes placed between the electrodes and another set extending radially outward from the electrodes, and also by the gas puff imaging diagnostic located 1 m away along the magnetic field lines intersecting the electrodes. Two fast cameras were also aimed directly at the electrode array. The hardware and controls of the biasing experiment will be presented and the initial effects on local plasma parameters will be discussed.

Here, the structure and motion of discrete plasma blobs (a.k.a. filaments) in the edge and scrape-off layer of NSTX is studied for representative Ohmic and H-mode discharges. Individual blobs were tracked in the 2D radial versus poloidal plane using data from the gas puff imaging diagnostic taken at 400 000 frames s-1. A database of blob amplitude, size, ellipticity, tilt, and velocity was obtained for ~45 000 individual blobs. Empirical relationships between various properties are described, e.g. blob speed versus amplitude and blob tilt versus ellipticity. The blob velocities are also compared with analytic models.

This paper describes the range of variations in edge and SOL turbulence observed using a gas puff imaging (GPI) diagnostic in NSTX discharges. The database consists of 140 shots including Ohmic, L-mode, and H-mode plasmas measured during steady-state conditions (e.g. without ELMs). Turbulence quantities were evaluated using both cross-correlation analysis and blob tracking. Relative fluctuation levels varied from –1.0, correlation times were –40 μs, correlation lengths were –10 cm, and turbulence velocities were km s−1 and km s−1 outwards. These variations were evaluated with respect to both the global and local edge plasma parameters, and compared with simplified theoretical models.

Summary form only given. Future toroidal magnetic confinement fusion plasma devices such as the Component Test Facility (CTF) require toroidal current drive. Present devices rely on a solenoid to drive current, however, for devices with a high neutron flux, it is impractical to provide the necessary shielding. An alternative method must be provided both to provide toroidal current. A new method of non-inductive startup, referred to as transient coaxial helicity injection (transient CHI), was successfully developed on the Helicity Injected Torus (HIT-II) experiment. In this method, a plasma current is produced by discharging a capacitor bank between coaxial electrodes in the presence of toroidal and poloidal magnetic fields chosen such that the plasma rapidly expands into the chamber. When the injected current is rapidly decreased, magnetic reconnection occurs near the injection electrodes, with the toroidal plasma current forming closed flux surfaces. In HIT-II CHI produced 100 kA of current on closed flux surfaces using 30 kA of injector current. Transient CHI has been successfully used on the National Spherical Torus Experiment (NSTX) for an unambiguous proof-of-principle demonstration of closed-flux current generation without the use of the central solenoid. When coupled with the HIT-II results, this demonstrates very favorable scaling to a larger device. The current multiplication factor between the injected current and the achieved toroidal current is 60 in NSTX, compared to six in HIT-II. Detailed experimental measurements made on NSTX include, fast time-scale visible imaging of the entire process. Experimental results from these experiments will be presented

Coaxial Helicity Injection (CHI), a non-inductive method to initiate plasma and generate toroidal plasma current, is being investigated in the National Spherical Torus Experiment (NSTX). The center stack and outer vacuum vessel are separated by insulating gaps at the top and bottom of the slim central column so that a high voltage (up to 2 kV) can be applied between them from a pulsed power supply or a capacitor bank to initiate an arc discharge. In the presence of a suitable poloidal magnetic field, the discharge is initiated at the lower gap (the injector gap) and because of the strong toroidal field develops a helical structure resulting in substantial toroidal plasma current being driven. In NSTX, up to 390 kA of toroidal current has been generated for an injected current of 25 kA. The early investigations of CHI however frequently developed arcs across the insulator at the top of the machine (the absorber gap), which terminated the desired discharge. This arcing greatly restricted the operational space available for CHI studies. During 2002, the absorber region was modified to suppress these arcs. The new design includes a new ceramic insulator on the high field side of the absorber region with a much longer tracking distance between conducting elements at the different potentials. Furthermore, two new coils were installed near the absorber to provide the ability to minimize the poloidal field connecting the center stack and outer vacuum vessel. During the subsequent experimental campaign, CHI operation was less prone to arcing in the absorber and those arcs that did occur did not terminate the main discharge

H-mode operation plays a crucial role in NSTX research, allowing higher beta limits due to reduced plasma pressure peaking, and long pulse operation due to high bootstrap current fraction. Here, new results are presented in the areas of edge localized modes (ELMs), H-mode pedestal physics, L-H turbulence, and power threshold studies, ELMS of several other types (as observed in conventional aspect ratio tokamaks) are often observed: 1) large, Type I ELMs, 2) "medium" Type II/III ELMs, and 3) giant ELMs which can reduce stored energy by up to 30% in certain conditions. In addition, many high performance discharges in NSTX have tiny ELMs (newly termed Type V), which have some differences as compared with ELM types in the published literature. The H-mode pedestal typically contains between 25-33% of the total stored energy, and the NSTX pedestal energy agrees reasonably well with a recent international multi-machine scaling. We find that the L-H transition occurrs on a ~ 100 μsec timescale as viewed by a gas puff imaging diagnostic, and that intermittent quiescent periods precede the final transition. A power threshold identity experiment between NSTX and MAST shows comparable loss power at the L-H transition in balanced double-null discharges. Both machines require more power for the L-H transition as the balance is shifted toward lower single null. High field side gas fueling enables more reliable H-mode access, but does not always lead to a lower power threshold e.g. with a reduction of the duration of early heating. Finally the edge plasma parameters just before the L-H transition were compared with theories of the transition. It was found that while some theories can separate well-developed L- and H-mode data, they have little predictive value. The National Spherical Torus Experiment (NSTX) is a medium-size, low aspect ratio spherical torus(1) with both neutral beam (NBI) and radio-frequency (RF) auxiliary heating, and main parameters: R=0.86m, a=0.67m, R/a ! 1.26, B t ≤ 0.6 T, I p ≤ 1.5 MA, P NBI < 7.4 MW, P RF < 6 MW. An assessment of the attractiveness of the spherical torus (ST) concept in the areas of high beta (β) stability, turbulence and transport, start-up and current drive, and boundary physics is the main research goal of NSTX. This paper describes H-mode physics research in NSTX with the primary focus on imaging of edge-localized modes (ELMs) with new diagnostics; an overview of recent NSTX results is given in (2). I. Edge localized modes Several ELM types from the higher aspect ratio tokamaks have been observed on NSTX. The large ELMs with stored energy drop between 5-15% were observed at high normalized beta (β N )

The 2-D structure of edge plasma turbulence has been measured in the National Spherical Torus Experiment (NSTX) by viewing the emission of the Da spectral line of deuterium. Images have been made at framing rates of up to 250,000 frames/sec using an ultra-high speed CCD camera developed by Princeton Scientific Instruments. A sequence of images showing the transition between L-mode and H-mode states is shown.

Research on the Spherical Torus (or Spherical Tokamak) 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 Spherical Tours (ST) experiments are being conducted in various U.S. research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium-size 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 U.S., 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 (B), noninductive 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 bT of up to 35% with the near unity central betaT have been obtained. NSTX will be exploring advanced regimes where bT up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for noninductive sustainment in NSTX is the high beta-poloidal regime, where discharges with a high noninductive fraction ({approx}60% bootstrap current + neutral-beam-injected current drive) were sustained over the resistive skin time. Research on radio-frequency-based heating and current drive utilizing HHFW (High Harmonic Fast Wave) and EBW (Electron Bernstein Wave) is also pursued on NSTX, Pegasus, and CDX-U. For noninductive start-up, the Coaxial Helicity Injection (CHI), developed in HIT/HIT-II, has been adopted on NSTX to test the method up to Ip {approx} 500 kA. In parallel, start-up using radio-frequency 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.

The two-dimensional radial versus poloidal structure and motion of edge turbulence in NSTX (National Spherical Torus Experiment) were measured by using high-speed imaging of the visible light emission from a localized neutral gas puff. Edge turbulence images are shown and analyzed for Ohmic, L-mode (low-confinement mode) and H-mode (high-confinement mode) plasma conditions. Typical edge turbulence poloidal correlation lengths as measured using this technique are = 4 {+-} 1 cm and autocorrelation times are 40 {+-} 20 {micro}sec in all three regimes. The relative fluctuation level is typically smaller in H-mode than in L-mode, and transitions from H- to L-mode and can occur remarkably quickly (=30 {micro}sec). The two-dimensional images often show localized regions of strong light emission which move both poloidally and radially through the observed region at a typical speed of =10{sup 5} cm/sec, and sometimes show spatially coherent modes.

The National Spherical Torus Experiment (NSTX) has produced toroidal plasmas at low aspect ratio (A = R/a = 0.86 m/0.68 m approximately equal to 1.3, where R is the major radius and a is the minor radius of the torus) with plasma currents of 1.4 MA. The rapid development of the machine has led to very exciting physics results during the first full year of physics operation. Pulse lengths in excess of 0.5 sec have been obtained with inductive current drive. Up to 4 MW of High Harmonic Fast Wave (HHFW) heating power has been applied with 6 MW planned. Using only 2 MW of HHFW heating power clear evidence of electron heating is seen with HHFW, as observed by the multi-point Thomson scattering diagnostic. A noninductive current drive concept known as Coaxial Helicity Injection (CHI) has driven 260 kA of toroidal current. Neutral-beam heating power of 5 MW has been injected. Plasmas with beta toroidal (= 2 mu(subscript ''0'') /B(superscript ''2'') = a measure of magnetic confinement efficiency ) of 22% have been achieved, as calculated using the EFIT equilibrium reconstruction code. Beta-limiting phenomena have been observed, and the maximum beta toroidal scales with I(subscript ''p'')/aB(subscript ''t''). High frequency (>MHz) magnetic fluctuations have been observed. High-confinement mode plasmas are observed with confinement times of >100 msec. Beam-heated plasmas show energy confinement times in excess of those predicted by empirical scaling expressions. Ion temperatures in excess of 2.0 keV have been measured, and power balance suggests that the power loss from the ions to the electrons may exceed the calculated classical input power to the ions.

The National Spherical Torus Experiment (NSTX) is a proof-of-principle experiment designed to study the physics of Spherical Tori (ST), i.e., low-aspect-ratio toroidal plasmas. Important issues for ST research are whether the high-eta stability and reduced transport theoretically predicted for this configuration can be realized experimentally. In NSTX, the commissioning of a digital real-time plasma control system, the provision of flexible heating systems, and the application of wall conditioning techniques were instrumental in achieving routine operation with good confinement. NSTX has produced plasmas with R/a {approx} 0.85 m/0.68 m, A {approx} 1.25, Ip * 1.1 MA, BT = 0.3-0.45 T, k * 2.2, d * 0.5, with auxiliary heating by up to 4 MW of High Harmonic Fast Waves, and 5 MW of 80 keV D0 Neutral Beam Injection (NBI). The energy confinement time in plasmas heated by NBI has exceeded 100 ms and a toroidal beta (bT = 2m0 /BT02, where BT0 is the central vacuum toroidal magnetic field) up to 22% has be en achieved. HHFW power of 2.3 MW has increased the electron temperature from an initial 0.4 keV to 0.9 keV both with and without producing a significant density rise in the plasma. The early application of both NBI and HHFW heating has slowed the penetration of the inductively produced plasma current, modifying the current profile and, thereby, the observed MHD stability.

This paper describes a detailed examination of the effects of a relatively small pulsed deuterium gas puff on the edge plasma and edge turbulence in NSTX. This gas puff caused little or no change in the line-averaged plasma density or total stored energy, or in the edge density and electron temperature up to the time of the peak of the gas puff. The radial profile of the Dα light emission and the edge turbulence within this gas puff did not vary significantly over its rise and fall, implying that these gas puffs did not significantly perturb the local edge plasma or edge turbulence. These measurements are compared with modeling by DEGAS 2, UEDGE, and with simplified estimates for the expected effects of this gas puff.

Transport and turbulence profiles were directly evaluated using probes for the first time in the edge and scrape-off layer (SOL) of NSTX [Ono et al., Nucl. Fusion 40, 557 (2000)] in low (L) and high (H) confinement, low power (Pin∼ 1.3 MW), beam-heated, lower single-null discharges. Radial turbulent particle fluxes peak near the last closed flux surface (LCFS) at ≈4×1021 s−1 in L-mode and are suppressed to ≈0.2×1021 s−1 in H mode (80%–90% lower) mostly due to a reduction in density fluctuation amplitude and of the phase between density and radial velocity fluctuations. The radial particle fluxes are consistent with particle inventory based on SOLPS fluid modeling. A strong intermittent component is identified. Hot, dense plasma filaments 4–10 cm in diameter, appear first ∼2 cm inside the LCFS at a rate of ∼1×1021 s−1 and leave that region with radial speeds of ∼3–5 km/s, decaying as they travel through the SOL, while voids travel inward toward the core. Profiles of normalized fluctuations feature levels o...

The 2-D radial vs. poloidal structure of edge turbulence in the Alcator C-Mod tokamak [I.H. Hutchinson, R. Boivin, P.T. Bonoli et al., Nuclear Fusion 41(2001) 1391] was measured using fast cameras and compared with 3-D numerical simulations of edge plasma turbulence. The main diagnostic is Gas Puff Imaging (GPI), in which the visible D(subscript alpha) emission from a localized D(subscript 2) gas puff is viewed along a local magnetic field line. The observed D(subscript alpha) fluctuations have a typical radial and poloidal scale of approximately 1 cm, and often have strong local maxima (''blobs'') in the scrape-off layer. The motion of this 2-D structure motion has also been measured using an ultra-fast framing camera with 12 frames taken at 250,000 frames/sec. Numerical simulations produce turbulent structures with roughly similar spatial and temporal scales and transport levels as that observed in the experiment; however, some differences are also noted, perhaps requiring diagnostic improvement and/or additional physics in the numerical model.

Rapidly developing diagnostic, operational, and analysis capability is enabling the first detailed local physics studies to begin in high-beta plasmas of the National Spherical Torus Experiment (NSTX). These studies are motivated in part by energy confinement times in neutral-beam-heated discharges that are favorable with respect to predictions from the ITER-89P scaling expression. Analysis of heat fluxes based on profile measurements with neutral-beam injection (NBI) suggest that the ion thermal transport may be exceptionally low, and that electron thermal transport is the dominant loss channel. This analysis motivates studies of possible sources of ion heating not presently accounted for by classical collisional processes. Gyrokinetic microstability studies indicate that long wavelength turbulence with k(subscript ''theta'') rho(subscript ''i'') {approx} 0.1-1 may be suppressed in these plasmas, while modes with k(subscript ''theta'') rho(subscript ''i'') {approx} 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 [high-confinement mode] transitions occur with either NBI or HHFW heating. The power required for low-confinement mode (L-mode) to H-mode transitions far exceeds that expected from empirical edge-localized-mode-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.

The National Spherical Torus Experiment (NSTX) is a spherical tokamak with nominal plasma major radius R(subscript ''0'') = 0.85 m, minor radius a = 0.66 m, and aspect ratio A > 1.28. Typical discharge parameters are plasma current I (subscript ''p'') = 0.7-1.4 MA, toroidal magnetic field B(subscript ''t0'') = 0.25-0.45 Tesla at major radius R(subscript ''0''), elongation = 1.7-2.2, triangularity 0.3-0.5, line-average electron density = 2-5 x 10(superscript ''19'') m(superscript ''-3''), electron temperature T(subscript ''e'')(0) = 0.5-1.5 keV, and ion temperature T(subscript ''i'')(0) = 0.5-2 keV. The NSTX auxiliary heating systems can routinely deliver 4.5 MW of 80-keV deuterium neutral beams and 3 MW of 30-MHz high-harmonic fast-wave power. Kinetic profile diagnostics presently include a 10-channel, 30-Hz multipulse Thomson scattering system (MPTS), a 17-channel charge-exchange recombination spectroscopy (CHERS) system, a 48-chord ultra-soft X-ray (USXR) array, and a 15-chord bolometry array. Initial experiments utilizing auxiliary heating on NSTX have focused on MHD stability limits, confinement trends, studying H-mode characteristics, and performing initial power balance calculations.

We report observations of the first low-to-high (L-H) confinement mode transitions in the National Spherical Torus Experiment (NSTX). The H-mode energy confinement time increased over reference L-mode discharges transiently by 100-300%, as high as {approximately}150 ms. This confinement time is {approximately}1.8-2.3 times higher than predicted by a multi-machine ELM-free H-mode scaling. This achievement extends the H-mode window of fusion devices down to a record low aspect ratio (R/a) {approximately} 1.3, challenging both confinement and L-H power thresholds scalings based on conventional aspect ratio tokamaks.