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Recent U.S. fusion development strategy reports all recommend that the U.S. should pursue innovative science and technology to enable construction of a fusion pilot plant (FPP) that produces net electricity from fusion at low capital cost. Compact tokamaks have been proposed as a means of potentially reducing the capital cost of a FPP. However, compact steady-state tokamak FPPs face the challenge of integrating a high fraction of self-driven current with high core confinement, plasma pressure, and high divertor parallel heat flux. This integration is sufficiently challenging that a dedicated sustained-high-power-density (SHPD) tokamak facility is proposed by the U.S. community as the optimal way to close this integration gap. Performance projections for the steady-state tokamak FPP regime are presented and a preliminary SHPD device with substantial flexibility in lower aspect ratio (A = 2–2.5), shaping, and divertor configuration to narrow gaps to an FPP is described.

A spherical tokamak (ST) configuration is attractive as the basis for a fusion nuclear science facility (FNSF), due to its small size and relatively low cost. However, the compactness of the ST also exacerbates the power and particle handling problems anticipated in next-step devices, since local fluxes are higher and less space is available for optimizing plasma-facing components. On the other hand, novel divertor geometries that have recently been developed such as the snowflake and super-X divertors can be especially effective at reducing heat fluxes in an ST, helping to meet the exhaust challenge. Here, we present an analysis of the power and particle handling requirements of a candidate ST-FNSF, based on 0-D exhaust projections as well as 2-D edge plasma modeling using the SOLPS code. Both conventional and novel divertor geometries are considered. These show that, for reasonable assumptions on cross-field transport, operating points can be identified that are consistent with both core plasma operation and power and particle exhaust requirements, and these operating points are more easily accessible with novel divertors.

Tokamaks are sensitive to deviations from axisymmetry as small as δB /B0 ∼ 0.01% with significant degradation or improvement in performance. Ideal Perturbed Equilibrium Code (IPEC) has been developed in order to understand plasma response to such a small non-axisymmetric perturbation in tokamaks. IPEC has been benchmarked in cylindrical limits, with CAS3D code, and has been validated through Resonant Field Amplification (RFA) measurements and error field correction results. The calculations of perturbed equilibria, such as IPEC in tokamaks, can be used to decompose the distributions of external magnetic perturbations on a control surface by their importance to critical plasma properties. The coupling between the external field at the control surface and the resonant field driving islands at the rational surfaces determine the most important distribution of external magnetic field on the control surface for island opening and plasma locking. It has been shown that the most important distribution of external magnetic field changes little across various plasma profiles and parameters, indicating the possibility of robust error field corrections. The described IPEC method can be also applied to general 3D geometry if a relevant perturbed equilibrium code such as CAS3D is utilized (© 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

The National Spherical Torus Experiment (NSTX) produces plasmas with toroidal aspect ratio as low as 1.25, which can be heated by up to 6 MW high-harmonic fast waves and up to 7 MW of deuterium neutral beam injection. Using new poloidal field coils, plasmas with cross-section elongation up to 2.7, triangularity 0.8, plasma currents Ip up to 1.5 MA and normalized currents Ip/aBT up to 7.5 MA/mT have been achieved. A significant extension of the plasma pulse length, to 1.5 s at a plasma current of 0.7 MA, has been achieved by exploiting the bootstrap and NBI-driven currents to reduce the dissipation of poloidal flux. Inductive plasma startup has been supplemented by coaxial helicity injection (CHI) and the production of persistent current on closed flux surfaces by CHI has now been demonstrated in NSTX. The plasma response to magnetic field perturbations with toroidal mode numbers n = 1 or 3 and the effects on the plasma rotation have been investigated using three pairs of coils outside the vacuum vessel. Recent studies of both MHD stability and of transport benefitted from improved diagnostics, including measurements of the internal poloidal field using the motional Stark effect (MSE). In plasmas with a region of reversed magnetic shear in the core, now confirmed by the MSE data, improved electron confinement has been observed.

In 2003, the National Spherical Torus Experiment (NSTX) plasma control system was used for plasma shape control using real-time equilibrium reconstruction (using the rtEFIT code). rtEFIT is now in routine use for plasma boundary control. More recently, the system has been upgraded to support feedback control of the resistive wall mode (RWM). This paper describes the hardware and software improvements that were made in support of these physics requirements. The control computer is an eight processor embedded system. The real-time data acquisition system now acquires 352 channels of data at 5 kHz for each NSTX plasma discharge. The latency for the data acquisition, which uses the Front Panel Data Port (FPDP) protocol, is measured to be ∼8 μs. A Stand-Alone Digitizer (SAD), designed at PPPL, along with an FPDP Input Multiplexing Module (FIMM) allows for simple modular upgrades. An interface module was built to interface between the FPDP output of the NSTX control system and the legacy Power Conversion Link (PCLink) used for communicating with the PPPL power supplies (first used for TFTR). Additionally a module has been built for communicating with the switching power amplifiers (SPA) recently installed on NSTX. In addition to the hardware developments, the control software on the NSTX control system has been upgraded. The device driver software for the hardware described above will be discussed, as well as the new control algorithms that have been developed to control the switching power supplies for RWM control. A reliable mode detection algorithm for RWM feedback is currently under development.

Saturated internal kink modes have been observed in many of the highest toroidal β discharges of the National Spherical Torus Experiment (NSTX). These modes often cause rotation flattening in the plasma core, can degrade fast-ion confinement and in some cases contribute to the complete loss of plasma angular momentum and stored energy. Characteristics of the modes are measured using soft x-ray, kinetic profile and magnetic diagnostics. Toroidal flows approaching Alfvenic speeds, island pressure peaking and enhanced viscous and diamagnetic effects associated with high-β may contribute to mode nonlinear stabilization. These saturation mechanisms are investigated for NSTX parameters and compared with experimental data.

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.

A simple analog circuit that measures the National Spherical Torus Experiment (NSTX) axisymmetric eddy current distribution has been designed and constructed. It is based on simple circuit model of the NSTX vacuum vessel that was calibrated using a special axisymmetric eddy current code which was written so that accuracy was maintained in the vicinity of the current filaments [J. Menard, J. Fusion Tech. (to be published)]. The measurement and the model have been benchmarked against data from numerous vacuum shots and they are in excellent agreement. This is an important measurement that helps give more accurate equilibrium reconstructions.

Ion absorption of the high harmonic fast wave in a spherical torus [Y.-K. M. Peng et al., Nucl. Fusion 26, 769 (1986)] is of critical importance to assessing the viability of the wave as a means of heating and driving current. Analysis of recent National Spherical Torus Experiment [M. Ono et al., Nucl. Fusion 40, 557 (2000)] shots has revealed that under some conditions when neutral beam and rf power are injected into the plasma simultaneously, a fast ion population with energy above the beam injection energy is sustained by the wave. In agreement with modeling, these experiments find the rf-induced fast ion tail strength and neutron rate at lower B-fields to be less enhanced, likely due to a larger β profile, which promotes greater off-axis absorption where the fast ion population is small. Ion loss codes find the increased loss fraction with decreased B insufficient to account for the changes in tail strength, providing further evidence that this is a rf interaction effect. Though greater ion absorption...

One of the goals of the National Spherical Torus Experiment (NSTX) is to investigate the physics of global mode stabilization in a low aspect ratio device. NSTX has a major radius R0 = 0.86 m, a midplane half-width of 0.7 m, and an on-axis vacuum toroidal field B0 0.6 T and has reached a plasma current Ip = 1.5 MA. Experiments have established the wall-stabilized MHD operating space of the machine. The maximum βt and βN have reached 35% and 6.5%, respectively, with βN reaching 9.5li. Collapses in plasma toroidal rotation and βt have been correlated with violation of the n = 1 ideal MHD beta limit, β Nn o-wall, computed by the DCON stability code using timeevolving EFIT reconstructions of experimental discharges. The resistive wall mode (RWM) was observed over a wide range of βN when β Nn o-wall was exceeded. Plasma toroidal rotation damping during the RWM was rapid and global. Damping rates were more than five times larger than caused by low toroidal mode number rotating modes alone, which displayed a slower, diffusive rotation damping away from the rational surface. The rotation damping rate and dynamics depend on the applied toroidal field and the computed minimum value of the safety factor. The computed RWM perturbed field structure from experimental plasma reconstructions has been input to the VALEN feedback analysis code for quantitative comparison of experimental and theoretical RWM growth rates and to analyse the effectiveness of various active feedback stabilization designs. The computed RWM n = 1 mode growth rate, which depends on plasma equilibrium parameters such as βN and pressure profile peaking, agrees well with experimental growth rates in different operating regimes. Increasing βN in the ST initially improves mode coupling to the stabilizing wall; however, at the highest βN values reached, the ideal with-wall beta limit, βN wall, is approached, the effectiveness of the passive stabilizing plates is reduced, and the computed RWM growth rate approaches ideal MHD growth rates. Several active mode control designs were considered and evaluated. The most effective configuration is computed to provide stabilization at βN up to 94% of the ideal with-wall limit.

The confinement of auxiliary heated national spherical torus experiment discharges is discussed. From a database analysis, it is found that the energy confinement time in NBI heated plasmas with either L- or H-modes edge is up to 2.5 times the values predicted by the ITER97L scaling. A high power NBI heated H-mode discharge is discussed in detail. TRANSP calculations based on the kinetic profile measurements reproduce well the magnetically determined stored energy, but overestimate the measured neutron rate by 30%. Power balance calculations reveal that the ion thermal transport is above or near neoclassical levels, and significantly below the electron thermal transport, which constitutes the main power loss channel. Perturbative impurity injection techniques indicate the particle diffusivity is slightly above the neoclassical level in discharges with L-mode edge. High-harmonic fast-wave (HHFW) bulk electron heating is described and thermal transport is discussed. Thermal ion transport is found to be above the neoclassical level, and thermal electron transport remains the main loss mechanism. Evidence of an electron thermal internal transport barrier obtained with HHFW heating is presented. A description of H-mode discharges obtained during HHFW heating is made.

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 (

High harmonic fast wave (HHFW) heating has been proposed as a particularly attractive means for plasma heating and current drive in the high beta plasmas that are achievable in spherical torus (ST) devices. The National Spherical Torus Experiment (NSTX) [M. Ono, S. M. Kaye, S. Neumeyer et al., in Proceedings of the 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque, 1999 (IEEE, Piscataway, NJ, 1999), p. 53] is such a device. An rf heating system has been installed on the NSTX to explore the physics of HHFW heating, current drive via rf waves and for use as a tool to demonstrate the attractiveness of the ST concept as a fusion device. To date, experiments have demonstrated many of the theoretical predictions for HHFW. In particular, strong wave absorption on electrons over a wide range of plasma parameters and wave parallel phase velocities, wave acceleration of energetic ions, and indications of current drive for directed wave spectra have been observed. In addition HHFW heating has been used to ...

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 boundary physics program presently focuses on edge power and particle flow optimization and control in β⩾25% long pulse L- and H-mode plasmas with high harmonic fast wave heating power up to 6 MW and neutral beam injection power up to 5 MW, with the emphasis on the edge physics implications resulting from the low aspect ratio geometry. To address the particle flux measurements in the divertor and plasma scrape-off layer (SOL), two spectrally filtered one-dimensional charge coupled device cameras have been fielded. The cameras utilize 2048 pixel 12-bit Dalsa CL-C6 arrays, f=85 mm lenses, and Δλ=1.5 nm bandpass interference filters. Both cameras provide mm spatial resolution, sub-ms temporal resolution and are spatially and photometrically calibrated. Midplane SOL and divertor brightness profiles of C III and deuterium species have been obtained in the L- and H-mode phases of center stack limited and diverted plasmas. The equilibria reconstructed by the EFIT code are ...

The behavior of large scale magnetohydrodynamic modes was investigated in the National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 (2000)] during reconnection events using combined analysis of magnetic probe signals, and soft x-ray data. The comparison of mode dynamics during precursor and disruption stages in T-11M (V. S. Vlasenkov et al., in Proceedings of the 6th International Conference on Plasma Physics and Controlled Nuclear Fusion Research, Berchtesgaden, 6–13 October 1976, p. 95) (small circular plasma), Tokamak Fusion Test Reactor [R. J. Hawryluk et al., Plasma Phys. Controlled Fusion 33, 1509 (1991)] (large circular plasma), and NSTX (large spherical plasma) was done. The analysis shows that the sequence of events of minor and major internal reconnection events (IRE) in NSTX is essentially similar to that for disruptions in moderate aspect ratio tokamaks. The main feature of disruption dynamics apparently affected by the small aspect ratio in NSTX appears in the relatively slow thermal quench event (5–10 times longer compared to ordinary tokamaks), which precedes the major IRE. The coincidence of the electron and neutron quench times during the major IRE leads us to the conclusion that the fast ions and hot electrons leave the center of a plasma column simultaneously, i.e., convectively.

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/.

Neoclassical tearing modes (NTMs) can lead to disruption and loss of confinement. Previous analysis of these modes used large aspect ratio, low β (plasma pressure/magnetic pressure) approximations to determine the effect of NTMs on tokamak plasmas. A more accurate tool is needed to predict the onset of these instabilities. As a follow-up to recent theoretical work, a code has been written which computes the tearing mode island growth rate for arbitrary tokamak geometry. It calls PEST-3 [A. Pletzer et al., J. Comput. Phys. 115, 530 (1994)] to compute Δ′, the resistive magnetohydrodynamic (MHD) matching parameter. The code also calls the FLUXGRID routines in NIMROD [A. H. Glasser et al., Plasma Phys. Controlled Fusion 41, A747 (1999)] for Dnc, DI, and DR [C. C. Hegna, Phys. Plasmas 6, 3980 (1999); A. H. Glasser et al., Phys. Fluids 18, 875 (1975)], which are the bootstrap current driven term and the ideal and resistive interchange mode criterion, respectively. In addition to these components, the NIMROD rou...

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.

Boronization of NSTX has allowed access to higher density higher confinement plasmas. A glow discharge with 4 mtorr helium and 10% deuterated trimethyl boron deposited 1.7 g of boron on the plasma facing surfaces. Ion beam analysis of witness coupons showed a B + C areal density of 1018 cm-2 corresponding to a film thickness of 100 nm. Subsequent ohmic discharges showed oxygen emission lines reduced by a factor of 15, carbon emission reduced by a factor of two and copper reduced to undetectable levels. After boronization, the plasma current flat-top time increased by 70% enabling access to higher density higher confinement plasmas.

This paper examines a method for real-time control of non-inductively sustained scenarios in NSTX-U by using TRANSP, a time-dependent integrated modeling code for prediction and interpretive analysis of tokamak experimental data, as a simulator. The actuators considered for control in this work are the six neutral beam sources and the plasma boundary shape. To understand the response of the plasma current, stored energy, and central safety factor to these actuators and to enable systematic design of control algorithms, simulations were run in which the actuators were modulated and a linearized dynamic response model was generated. A multi-variable model-based control scheme that accounts for the coupling and slow dynamics of the system while mitigating the effect of actuator limitations was designed and simulated. Simulations show that modest changes in the outer gap and heating power can improve the response time of the system, reject perturbations, and track target values of the controlled values.

Electron cyclotron emission (ECE) has been employed as a standard electron temperature profile diagnostic on many tokamaks and stellarators, but most magnetically confined plasma devices cannot take advantage of standard ECE diagnostics to measure temperature. They are either “overdense,” operating at high density relative to the magnetic field (e.g., ωpe≫Ωce in a spherical torus) or they have insufficient density and temperature to reach the blackbody condition (τ>2). Electron Bernstein waves (EBWs) are electrostatic waves that can propagate in overdense plasmas and have a high optical thickness at the electron cyclotron resonance layers as a result of their large kperp. In this article we report on measurements of EBW emission on the CDX-U spherical torus, where B0∼2 kG, 〈ne〉∼1013 cm−3 and Te≈10–200 eV. Results are presented for electromagnetic measurements of EBW emission, mode converted near the plasma edge. The EBW emission was absolutely calibrated and compared to the electron temperature profile me...

The five ARIES designs, which correspond to five different tokamak operating modes, are reviewed and compared. Physics figures of merit are introduced that quantify the major parameters of a tokamak design in a physics operating space. The five operating modes are compared to one another and to the existing tokamak data base in terms of these physics parameters. While the steady-state first stability design [ARIES-I] and the pulsed first stability design [PULSAR] are closest, no design has yet been completely prototyped in existing tokamaks.

Electron density and temperature in the CDX-U low aspect ratio tokamak are too low to allow observation of fast magnetohydrodynamic activity using soft x-ray continuum emission. However, spectroscopic measurements show that extreme ultraviolet (XUV) line emission of intrinsic impurities is bright enough to observe such activity. In addition, a fast monitoring system for local temperature changes in the plasma core is required for planned auxiliary heating experiments. We present a spectrally resolved tomographic system for fast imaging of O VI 2s–np, 2p–nd (n⩾3), C V 1s2–1s2p and C VI 1s–2p XUV transitions. Using this emission, we can study both core and edge MHD activity, while the C VI to C V intensity ratio can indicate local changes in electron temperature. To achieve maximal throughput together with the needed spectral resolution, we use arrays of surface barrier diodes filtered with bandpass elemental filters. Using M edge filters (Zr, Pd, and Ag), we achieve both good discrimination between the abo...

Elimination of the central solenoid would simplify the engineering design of a Fusion Nuclear Science Facility (FNSF) and tokamak based devices. The method of Transient Coaxial Helicity Injection (CHI) has successfully demonstrated formation of a high-quality closed flux plasma in NSTX and will be used as the front end of the start-up method for a full demonstration of non-inductive current start-up, followed by non-inductive current ramp-up using neutral beams in the NSTX-U device that is now under construction at the Princeton Plasma Physics Laboratory (PPPL). CHI is implemented by driving current along open field lines that connect the lower inner and outer divertor plates of a Spherical Torus (ST). The engineering system requirements and the design of the CHI system on NSTX-U are described.

The design features developed for the Spherical Tokamak (ST) in the PPPL pilot plant study was used as the starting point in developing designs to meet the mission of a Fusion Nuclear Science Facility (FNSF) considering a range of machine sizes based on the influence of tritium consumption and maintenance strategies. The compact nature of a steady state operated ST device for this mission pushes operating conditions and places challenges in the design of components, device maintenance and the integration of supports and services. This paper reviews the general arrangement, design details and maintenance strategy of the ST-FNSF device core for a 1.6-m and 1.0-m device; operating points which bracket the region between purchasing and breeding tritium.

The maximum achievable tritium breeding ratio (TBR), the dose to the insulator of the Cu coils, and the radial build definition are among numerous design issues investigated in detail for a Fusion Nuclear Science Facility (FNSF) based on the spherical tokamak (ST) concept. The ongoing PPPL study is considering a range of machine sizes with 1-2.2 m major radius. Preliminary shielding analysis for the PF coils of the intermediate size machine (R~1.7 m) indicated excessive dose to the cyanate ester/epoxy organic insulator, suggesting a more radiation resistant ceramic insulator such as MgO. The 3-D analysis predicts a TBR of ~1 when the details of the dual-cooled LiPb blanket and outboard penetrations are included in the model. Potential means to increase the TBR were investigated. Understanding the impact of various device sizes on the TBR is another important ongoing research activity to determine the threshold in device size for achieving T self-sufficiency.

This paper discusses disruption rates, disruption causes, and disruptivity statistics in the high- βN National Spherical Torus Experiment (NSTX) [M. Ono, et al. Nuclear Fusion 40, 557 (2000)]. While the overall disruption rate is rather high, configurations with high βN , moderate q*, strong boundary shaping, sufficient rotation, and broad pressure and current profiles are found to have the lowest disruptivity; active n=1 control further reduces the disruptivity. The disruptivity increases rapidly for q* 1.25 is generally acceptable for avoiding the onset of core rotating n=1 kink/tearing modes; when EPM and ELM disturbances are present, the required qmin for avoiding those modes is raised to ~1.5. The current ramp and early flat-top phase of the discharges are prone to n=1 core rotating modes locking to the wall, leading to a disruption. Small changes to the discharge fueling during this phase can often mitigate the rotation damping associated with these modes and eliminate the disruption. The largest stored energy disruptions are those that occur at high current when a plasma current rampdown is initiated incorrectly.

This paper describes techniques for measuring halo currents, and their associated toroidal peaking, in the National Spherical Torus Experiments [M. Ono et al., Nucl. Fusion 40, 557 (2000)]. The measurements are based on three techniques: (1) measurement of the toroidal field created by the poloidal halo current, either with segmented Rogowski coils or discrete toroidal field sensors, (2) the direct measurement of halo currents into specially instrument tiles, and (3) small Rogowski coils placed on the mechanical supports of in-vessel components. For the segmented Rogowski coils and discrete toroidal field detectors, it is shown that the toroidal peaking factor inferred from the data is significantly less than the peaking factor of the underlying halo current distribution, and a simple model is developed to relate the two. For the array of discrete toroidal field detectors and small Rogowski sensors, the compensation steps that are used to isolate the halo current signal are described. The electrical and mechanical design of compact under-tile resistive shunts and mini-Rogowski coils is described. Example data from the various systems are shown.

The lithium tokamak experiment (LTX) is a modest-sized spherical tokamak (R(0)=0.4 m and a=0.26 m) designed to investigate the low-recycling lithium wall operating regime for magnetically confined plasmas. LTX will reach this regime through a lithium-coated shell internal to the vacuum vessel, conformal to the plasma last-closed-flux surface, and heated to 300-400 °C. This structure is highly conductive and not axisymmetric. The three-dimensional nature of the shell causes the eddy currents and magnetic fields to be three-dimensional as well. In order to analyze the plasma equilibrium in the presence of three-dimensional eddy currents, an extensive array of unique magnetic diagnostics has been implemented. Sensors are designed to survive high temperatures and incidental contact with lithium and provide data on toroidal asymmetries as well as full coverage of the poloidal cross-section. The magnetic array has been utilized to determine the effects of nonaxisymmetric eddy currents and to model the start-up phase of LTX. Measurements from the magnetic array, coupled with two-dimensional field component modeling, have allowed a suitable field null and initial plasma current to be produced. For full magnetic reconstructions, a three-dimensional electromagnetic model of the vacuum vessel and shell is under development.

Transient Coaxial Helicity Injection (CHI) started discharges in NSTX have attained peak currents up to 300 kA and when these discharges are coupled to induction, it has produced up to 200 kA additional current over inductive-only operation. CHI in NSTX has shown to be energetically quite efficient, producing a plasma current of about 10 A/Joule of capacitor bank energy. In addition, for the first time, the CHI produced toroidal current that couples to induction continues to increase with the energy supplied by the CHI power supply at otherwise similar values of the injector flux, indicating the potential for substantial current generation capability by CHI in NSTX and in future toroidal devices. __________________________________________________

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).

Alfv́n cascade (AC) modes are observed in the National Spherical Torus Experiment [M. Ono, Nucl. Fusion 40, 557 (2000)] reversed shear plasmas over a wide range (up to ∼25% on axis, or ∼11% at minimum q) of Β (ratio of kinetic pressure to magnetic pressure). At low Β, the AC mode spectrum shows characteristics similar to conventional tokamaks. At higher Β, distinct Β and ∇Β effects are observed in the spectrum, including a significant reduction in the relative size of the frequency sweep and a toroidal mode number dependence in the minimum mode frequency. AC mode structure is obtained using reflectometry. Fast-ion loss associated with AC mode activity is observed. AC mode polarization at the plasma edge is consistent with expectation. Magnetohydrodynamic (MHD) spectroscopy is shown to be usable to determineqminat both low Β and high Β. Observed AC mode structure and frequency are found to be consistent with calculations for the same plasma conditions and geometry using the linear, ideal MHD hybrid kinetic code NOVA-K [C. Z. Cheng, Phys. Rep. 211, 1 (1992)]. © 2008 American Institute of Physics.

The coupling of Alfven Cascade (AC) modes or reversed-shear Alfven eigenmodes (rsAE) to Geodesic Acoustic Modes (GAM) implies that the range of the AC frequency sweep is reduced as the electron β is increased. This model provides an explanation for the otherwise surprising absence of AC modes in reverse shear NSTX plasmas, given the rich spectrum of beam-driven instabilities typically seen in NSTX. In experiments done at very low β to investigate this prediction, AC modes were seen, and as the βe was increased from shot to shot, the range of the AC frequency sweep was reduced, in agreement with this theoretical prediction.

A method of coaxial helicity injection has successfully produced a closed flux current without the use of the central solenoid in the NSTX device, on a size scale closer to a spherical torus reactor, for a proof-of-principle demonstration of this concept. For the first time, a remarkable 60 times current multiplication factor was achieved. Grad-Shafranov plasma equilibrium reconstructions are used to verify the existence of closed flux current. In some discharges the generated current persists for a surprisingly long time approximately 400 ms.

A supersonic gas injector (SGI) has been developed for fueling and diagnostic applications on the National Spherical Torus Experiment (NSTX). It is comprised of a graphite converging-diverging Laval nozzle and a commercial piezoelectric gas valve mounted on a movable probe at a low field side midplane port location. Also mounted on the probe is a diagnostic package: a Langmuir probe, two thermocouples and five pick-up coils for measuring toroidal, radial, vertical magnetic field components and magnetic fluctuations at the location of the SGI tip. The SGI flow rate is up to 4times1021 particles/s, comparable to conventional NSTX gas injectors. The nozzle operates in a pulsed regime at room temperature and a reservoir gas pressure up to 0.33 MPa. The deuterium jet Mach number of about 4, and the divergence half-angle of 5deg - 25deg have been measured in laboratory experiments simulating NSTX environment. In initial NSTX experiments reliable operation of the SGI and all mounted diagnostics at distances 1-20 cm from the plasma separatrix has been demonstrated. The SGI has been used for fueling of ohmic and 2-4 MW NBI heated L- and H-mode plasmas. Fueling efficiency in the range 0.1-0.3 has been obtained from the plasma electron inventory analysis

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

During neutral beam injection (NBI) in the National Spherical Torus Experiment (NSTX), a wide variety of fast ion driven instabilities is excited by the large ratio of fast ion velocity to Alfven velocity, together with the relatively high fast ion beta, beta(sub)f. The fast ion instabilities have frequencies ranging from a few kilohertz to the ion cyclotron frequency. The modes can be divided roughly into three categories, starting with Energetic Particle Modes (EPM) in the lowest frequency range (0 to 120 kHz), the Toroidal Alfven Eigenmodes (TAE) in the intermediate frequency range (50 to 200 kHz) and the Compressional and Global Alfven Eigenmodes (CAE and GAE, respectively) from approximately equal to 300 kHz up to the ion cyclotron frequency. Each of these categories of modes exhibits a wide range of behavior, including quasi-continuous oscillation, bursting, chirping and, except for the lower frequency range, turbulence.

Quantitative benchmarks of computational physics codes against experiment are essential for the credible application of such codes. Fluctuation measurements can provide necessary critical tests of nonlinear gyrokinetic simulations, but such require extraordinary computational resources. Linear microstability calculations with the GS2 [1] gyrokinetic code have been carried out for tokamak and ST experiments which exhibit internal transport barriers (ITB) and good plasma confinement. Qualitative correlation is found for improved confinement before and during ITB plasmas on Alcator C-Mod [2] and NSTX [3] with weaker long wavelength microinstabilities in the plasma core regions. Mixing length transport models are discussed. The NSTX L-mode is found to be near marginal stability for kinetic ballooning modes. Fully electromagnetic, linear, gyrokinetic calculations of the Alcator C-Mod ITB during off-axis rf heating, following four plasma species and including the complete electron response show ITG/TEM microturbulence is suppressed in the plasma core and in the barrier region before barrier formation, without recourse to the usual requirements of velocity shear or reversed magnetic shear [4-5]. No strongly growing long or short wavelength drift modes are found in the plasma core but strong ITG/TEM and ETG drift wave turbulence is found outside the barrier region. Linear microstability analysis is qualitatively consistent with the experimental transport analysis, showing low transport inside and high transport outside the ITB region before barrier formation, without consideration of ExB shear stabilization. Calculations of the ITG mode instability threshold and the dependence of instability on temperature and density scaling length have been used to test the GS2 gyrokinetic model for ITG microturbulence on C-Mod [4-5]. Transport analysis at the ITB onset time in Hmode experiments [6] is in rough agreement with the linear mixing length model !/ . The mixing length estimates exceed experimental estimates of transport by less than a factor of two and are in agreement with experiment in the plasma core (see Table I). GS2