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Trapped electron mode (TEM) is the main source of turbulence predicted for the unique operation regime of a flat temperature profile under low-recycling conditions in the LTX-β tokamak, while ion temperature gradient driven turbulence may also occur with gas fueling from the edge. To investigate mainly TEM scale density fluctuations, a high spatial and time resolution 2D beam emission spectroscopy (BES) diagnostic is being developed. Apart from spatially localized density turbulence measurement, BES can provide turbulence flow and flow shear dynamics. This BES system will be realized using an avalanche photodiode-based camera and narrow band interference filter. The system can acquire data at 2 MHz. Simulations with the Simulation of Spectra (SOS) code indicate that a high signal to noise ratio can be achieved with the proposed system. This will enable sampling the density fluctuations at this high time resolution. The design considerations and system optimization using the SOS code are presented.

Pure lithium (Li) surfaces are difficult to maintain in fusion devices due to rapid oxide formation, therefore, parameterizing and understanding the mechanisms of hydrogen (H, D) retention in lithium oxide (Li2O) in addition to pure Li is crucial for Li plasma-facing material applications. To compare H retention in Li and Li2O films, measurements were made as a function of surface temperature (90–520 K) under ultrahigh vacuum (UHV) conditions using temperature programmed desorption (TPD). In both cases, the total retention dropped with surface temperature, from 95% at 90 K to 35% at 520 K Li2O films retained H in similar amounts as pure Li. Molecular Dynamics (MD) modeling was used to elucidate the mechanisms of H retention, and results were consistent with experiments in terms of both retention fraction and the drop of retention with temperature.

An 8-channel, high

The characteristics of lithium-coated plasma-facing components (PFCs) have been correlated with plasma performance on the lithium tokamak experiment-β (LTX-β). Previous experiments on LTX showed that the application of lithium to PFCs was needed to achieve higher performance discharges with flat electron temperature profiles and high edge temperatures. Samples that match the LTX-β PFCs were exposed to plasmas before and after PFCs were coated with lithium and transferred under vacuum to a surface analysis station. Measurements using x-ray photoelectron spectroscopy (XPS) revealed that the primary surface constituent was lithium oxide. Earlier XPS analysis of lithium-coated PFCs on LTX was only able to show the presence of surface oxygen. The new XPS data from LTX-β have sufficient resolution to clearly identify lithium compounds for the first time, and enable them to be correlated with how lithium-coated PFCs can reduce impurities and retain hydrogen to reduce recycling.

We measured the spectral emission of plasmas from the National Spherical Tokamak Experiment in the extreme ultraviolet region, typically dominated by M-shell iron lines. Although we found that most of the significant emission in the 170–270 A region emanates from iron, there are also some strong lines of carbon present. We show that the carbon lines are not produced by electron-impact excitation, and we speculate that they are formed instead by charge exchange.

The Lithium Tokamak eXperiment (LTX) is a spherical tokamak with the unique capability of studying the low-recycling regime by coating nearly 90% of the first wall with lithium in either solid or liquid form. Several grams of lithium are evaporated onto the plasma-facing side of the first wall. Without lithium coatings, the plasma discharge is limited to less than 5 ms and only 10 kA of plasma current, and the first wall acts as a particle source. With cold lithium coatings, plasma discharges last up to 20 ms with plasma currents up to 70 kA. The lithium coating provides a low-recycling first wall condition for the plasma and higher fueling rates are required to realize plasma densities similar to that of pre-lithium walls. Traditional puff fueling, supersonic gas injection, and molecular cluster injection (MCI) are used. Liquid lithium experiments will begin in 2012.

Developing a reactor compatible divertor and managing the associated plasma material interaction (PMI) has been identified as a high priority research area for magnetic confinement fusion. Accordingly on NSTX-U, the PMI research has received a strong emphasis. Moreover, with ˜15 MW of auxiliary heating power, NSTX-U will be able to test the PMI physics with the peak divertor plasma facing component (PFC) heat loads of up to 40-60 MW/m2.

Lithium conditioning of plasma-facing surfaces has been implemented in National Spherical Torus Experiment (NSTX) leading to improvements in plasma performance such as reduced D recycling and a reduction in edge localized modes. Analysis of postmortem tiles and offline experiments along with atomistic modeling has identified interactions between Li-O-D and Li-C-D as chemical channels for deuterium retention in ATJ graphite. However, previous surface chemistry analysis of NSTX tiles were conducted postmortem (i.e., after a completed annual campaign), and it was not possible to correlate the performance of particular discharges with the state of the material surface at the time. Materials Analysis Particle Probe (MAPP) is the first in-vacuo surface analysis diagnostic directly integrated into a tokamak and capable of chemical surface analysis of plasma facing samples retrieved from the vessel in between discharges. It uses X-ray photoelectron spectroscopy, direct recoil spectroscopy, low energy ion surface spectroscopy, and thermal desorption spectroscopy to investigate the chemical functionalities between D and lithiated graphite at both the near surface (5-10 nm) and top surface layer (0.3-0.6 nm), respectively. MAPP will correlate plasma facing component surface chemistry with plasma performance and lead the way to improved understanding of plasma-surface interactions and their effect on global plasma performance. Remote operation and data acquisition, integrated into NSTX diagnostic and interlocks, make MAPP an advanced PMI diagnostic with stringent engineering constraints.

The Materials Analysis and Particle Probe (MAPP) is a compact in vacuo surface science diagnostic, designed to provide in situ surface characterization of plasma facing components in a tokamak environment. MAPP has been implemented for operation on the Lithium Tokamak Experiment at Princeton Plasma Physics Laboratory (PPPL), where all control and analysis systems are currently under development for full remote operation. Control systems include vacuum management, instrument power, and translational/rotational probe drive. Analysis systems include onboard Langmuir probes and all components required for x-ray photoelectron spectroscopy, low-energy ion scattering spectroscopy, direct recoil spectroscopy, and thermal desorption spectroscopy surface analysis techniques.

Use of a large-area liquid lithium surface as a limiter has significantly improved the plasma performance in the Current Drive Experiment-Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory. Previous CDX-U experiments with a partially-covered toroidal lithium limiter tray have shown a decrease in impurities and the recycling of hydrogenic species. Improvements in loading techniques have permitted nearly full coverage of the tray surface with liquid lithium. Under these conditions, there was a large drop in the loop voltage needed to sustain the plasma current. The data are consistent with simulations that indicate more stable plasmas having broader current profiles, higher temperatures, and lowered impurities with liquid lithium walls. As further evidence for reduced recycling with a liquid lithium limiter, the gas puffing had to be increased by up to a factor of eight for the same plasma density achieved with an empty toroidal tray limiter.

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.

Flux surface shape information can be used to constrain the current profile for reconstruction of the plasma equilibrium. One method of inferring flux surface shape relies on plasma x-ray emission; however, deviations from the flux surfaces due to impurity and density asymmetries complicate the interpretation. Electron isotherm surfaces should correspond well to the plasma flux surfaces, and equilibrium constraint modeling using this isotherm information constrains the current profile. The KFIT code is used to assess the profile uncertainty and to optimize the number, location and SNR required for the Te detectors. As Te imaging detectors we consider tangentially viewing, vertically spaced, linear gas electron multiplier arrays operated in pulse height analysis (PHA) mode and multifoil soft x-ray arrays. Isoflux coordinate sets provided by Te measurements offer a strong constraint on the equilibrium reconstruction in both a stacked horizontal array configuration and a crossed horizontal and vertical beam ...

A prototype pulsed supersonic gas injector (SGI) has been developed for the National Spherical Torus Experiment (NSTX). Experiments in NSTX will explore the compatibility of the supersonic gas jet fueling with the H-mode plasma edge, edge localized mode control, edge magnetohydrodynamic stability, radio frequency heating scenarios, and start-up scenarios with a fast plasma density ramp up. The diagnostic applications include localized impurity gas injections for transport and turbulence experiments and edge helium spectroscopy for edge Te and ne profile measurements. Nozzle and gas injector design considerations are presented and four types of supersonic nozzles are discussed. The prototype SGI operates at room temperature. It is comprised of a small graphite Laval nozzle coupled to a modified commercial piezoelectric valve and mounted on a movable vacuum feedthrough. The critical properties of the SGI jet—low divergence, high density, and sharp boundary gradient, achievable only at M>1, have been demonst...

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

An innovative fast system for x-ray imaging has been developed at ENEA-Frascati, Italy, in collaboration with INFN-Pisa, Italy. It is based on a pinhole camera coupled to a micropattern gas detector (MPGD) having a gas electron multiplier (GEM) as amplifying stage. This detector (2.5 cm×2.5 cm active area) is equipped with a two dimensional readout printed circuit board with 144 pixels (12×12). It is able to get x-ray images of the plasma at very high framing rate (up to 100 kHz) in a selectable x-ray energy range, with different magnifications or views of the plasma. The system has been tested successfully on the Frascati Tokamak Upgrade (FTU) during Summer 2001, with a one-dimensional perpendicular view of the plasma. In collaboration with ENEA, the Johns Hopkins University (JHU) and Princeton Plasma Physics (PPPL), this system has been setup and calibrated in the x-ray energy range 3–8 keV and it has been installed, with a tangential view, on the NSTX experiment at PPPL. Time resolved, two-dimensional ...

We develop an integrated instrumental and computational package for the diagnosis of impurity content, Zeff profile, particle transport, and magnetohydrodynamics activity in magnetic fusion experiment plasmas. The package includes broadband filtered arrays of absolute (AXUV) photodiodes, a transmission grating imaging spectrometer measuring up to 20 chords across the discharge and having a few angstrom spectral resolution, together with an atomic physics database coupled with an impurity transport code. The atomic physics database is based on ab initio computations with the Hebrew University Lawrence Livermore atomic code. The package is designed for the diagnostic of sub-keV plasmas having predominantly low-Z impurities (C, B, and O) together with trace metals. A preliminary version is being tested on the National Spherical Torus Experiment spherical torus at the Princeton Plasma Physics Laboratory, using the ultrasoft x-ray imaging system and a grazing incidence spectrometer. Representative results from...

The Johns Hopkins Plasma Spectroscopy Group is developing a transmission grating based imaging spectrometer for the ultrasoft x-ray [(USXR), 10–300 A] range. The spectrometer will be integrated into an impurity diagnostic package for magnetic fusion experiments, which provides time and space resolved information about radiation losses, Zeff profiles, and particle transport. The spectrometer has a simple layout, consisting of collimating and space resolving slits, a transmission grating, and a two-dimensional imaging USXR detector. We tested two types of detectors, a CsI coated multichannel plate and a phosphor P45 coated fiber optic plate, both with intensified charge-coupled-device image readout. The performance of the 5000 1/mm, 3:1 bar to open area ratio transmission grating has been evaluated in the laboratory using Kα lines from a Manson source and the emission from a Penning discharge. A prototype spectrometer equiped with the first type detector and optimized for 6 A spectral resolution has been te...

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/.

Results on the interaction between plasma in the current drive experiment-upgrade (CDX-U) spherical torus and a liquid lithium limiter are reported. It is observed that macroscopic lithium droplets detach from the limiter head and fall towards the plasma core. However, no disruptions occurred during these discharges despite the fact that relatively large-scale blobs are observed entering the confined plasma. A multi-tip Langmuir probe measures the edge plasma properties. It is found that the average density and temperature and their fluctuations are unaffected by the presence of lithium within experimental error.

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

Electron transport has been measured in the Current-Drive Experiment Upgrade (CDX-U) (T. Jones, Ph.D. thesis, Princeton University, 1995) using two separate perturbative techniques. Sawteeth at the q=1 radius (r/a∼0.15) induced outward-propagating heat pulses, providing time-of-flight information leading to a determination of χe as a function of radius. Gas modulation at the plasma edge introduced inward-propagating cold pulses, providing a complementary time-of-flight based χe profile measurement. This work represents the first localized measurement of χe in a spherical torus. Core (r/a

By reducing neutral recycling and providing a low-Z plasma-facing surface, lithium wall coatings have led to major performance improvements in TFTR1, CDX-U2, and NSTX3. The effects of lithium coatings on plasma performance are being studied in the Lithium Tokamak Experiment (LTX), a modest size (R=0.4 m, a=0.26 m) spherical tokamak. In LTX, lithium coatings can be applied to close-fitting stainless steel surfaces surrounding ∼90% of the plasma. The surfaces can be heated to ∼300°C, allowing plasma operations with partly or fully liquid lithium surfaces.

The importance of conditioning plasma-facing components (PFCs) has long been recognized as a critical element in obtaining high-performance plasmas in magnetic confinement devices. Lithium coatings, for example, have been used for decades for conditioning PFCs. Since the initial studies on the Tokamak Fusion Test Reactor (TFTR), experiments on devices with different aspect ratios and magnetic geometries like the National Spherical Torus Experiment (NSTX) continue to show the relationship between the lithium PFCs and good confinement and stability. While such results are promising, their empirical nature do not reflect the detailed relationship between the PFCs and the dynamic conditions that occur in the tokamak environment. A first step developing an understanding such complexity will be taken in the upgrade to NSTX, or the National Spherical Torus Experiment-Upgrade (NSTX-U) that is nearing completion. New measurement capabilities include the materials analysis and particle probe for in situ surface analysis of samples exposed to tokamak plasmas. The onion-skin modeling for edge analysis (OEDGE) suite of codes, for example, will be used to model the underlying mechanisms for such material migration in NSTX-U. This will lead to a better understanding of how plasma-facing surfaces evolve during a shot, and how the composition of the plasma-facing surface influences the discharge performance we observe. This paper will provide an overview of these capabilities, and highlight their importance for NSTX-U plans to transition from carbon to high- ${Z}$ PFCs.

Coaxial helicity injection (CHI) on the National Spherical Torus Experiment (NSTX) has produced 240 kA of toroidal current without the use of the central solenoid. Values of the current multiplication ratio (CHI produced toroidal current/injector current) up to 10 were obtained, in agreement with predictions. The discharges, which lasted for up to 200 ms, limited only by the programmed waveform, are more than an order of magnitude longer in duration than any CHI discharges previously produced in a spheromak or a spherical torus.

The mission of the National Spherical Torus Experiment (NSTX) is to extend the understanding of toroidal physics to low aspect ratio (R/a approximately equal to 1.25) in low collisionality regimes. NSTX is designed to operate with up to 6 MW of High Harmonic Fast Wave (HHFW) heating and current drive, 5 MW of Neutral Beam Injection (NBI) and Co-Axial Helicity Injection (CHI) for non-inductive startup. Initial experiments focused on establishing conditions that will allow NSTX to achieve its aims of simultaneous high-bt and high-bootstrap current fraction, and to develop methods for non-inductive operation, which will be necessary for Spherical Torus power plants. Ohmic discharges with plasma currents up to 1 MA and with a range of shapes and configurations were produced. Density limits in deuterium and helium reached 80% and 120% of the Greenwald limit respectively. Significant electron heating was observed with up to 2.3 MW of HHFW. Up to 270 kA of toroidal current for up to 200 msec was produced noninductively using CHI. Initial NBI experiments were carried out with up to two beam sources (3.2 MW). Plasmas with stored energies of up to 140 kJ and bt =21% were produced.

Coaxial helicity injection has been investigated on the National Spherical Torus Experiment (NSTX). Initial experiments produced 130 kA of toroidal current without the use of the central solenoid. The corresponding injector current was 20 kA. Discharges with pulse lengths up to 130 ms have been produced.

Recent upgrades to CDX-U spherical torus diagnostics include two 10-channel AXUV diode arrays. The multilayer mirror (MLM) array measures the λ150 O VI brightness profile in the poloidal plane using the Mo/B4C synthetic multilayer structures as dispersive elements. The foil filter array has a tangential view and is equipped with interchangeable clear aperture, beryllium and titanium filters. This allows measurements of radiated power, O VI or C V radial distributions, respectively. The O VI and C V emissivity and the radiated power profiles are highly peaked. A Neoclassical impurity accumulation mechanism is considered as an explanation. For radiated power measurements in the Te⩽100 eV plasmas, photon energy dependent corrections must be used in order to account for nonlinear AXUV sensitivity in the range Ephot⩽20 eV. The arrays are also used for characterization of resistive MHD phenomena, such as the low m modes, saw-tooth oscillations and internal reconnection events. Based on the successful operation ...

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

A spectrally resolved ultrasoft x-ray imaging system, consisting of arrays of high resolution (

A fast camera system is being prepared for the TEXTOR tokamak for the investigation of magnetohydrodynamics (MHD) instabilities. The camera will consist of a scintillator screen for conversion of x rays into visible light, a fiberoptic cable for transport of the visible images out of the Textor magnetic field, an electrostatic image converter for amplification and demagnification of the images, and a fast charge-coupled device (CCD) camera for capturing and storing the image data. The purpose of our camera is two-dimensional imaging of magnetic islands and the structure of MHD instabilities. In order to increase the signal intensity and simultaneously reduce statistical noise, we maximize the diameter of the scintillator screen to 100 mm. For good time response, we use fast scintillators and phosphors, and plan to employ a 1 MHz framing CCD camera. MHD events having an amplitude larger than 1% can be observed with a time resolution of 100 kHz and a spatial resolution of 3 cm. In the future, the camera will be used on the Large Helical Device in Japan. (C) 1999 American Institute of Physics. [S0034-6748(99)62201-8].

A simple low velocity boron micro-pellet injector has been under development for Current Drive Experiment Upgrade (CDX-U) spherical torus edge and core impurity transport measurements, and wall conditioning. The injector consists of 16 barrels on a rotatable turret. Each barrel can be loaded with boron powder particles of diameters ranging from 1 to 40 μm diameter in amounts ranging from less than 0.25 mg to more than 2 mg. A selected barrel is manually rotated into firing position using a vacuum precision rotary/linear motion feedthrough. A piezoelectric valve gas feed system triggered by CDX-U discharge timing is used to control H2 or D2 propellant gas at a cylinder pressure of 5.8×10−3 Pa (40 psi) or less. The injector barrel-to-CDX-U plasma edge distance is 0.47 m. Initial low mass injections of neutral boron beams were performed into CDX-U plasmas at a velocity of 23 m/s. Measurements were obtained with a filtered gated charge coupled device TV camera, bolometry, visible spectroscopy, and ultrasoft ...

The goals of the diagnostic set for initial plasma operations on the National Spherical Torus Experiment (NSTX) are to provide information on the fundamental discharge parameters necessary to characterize NSTX plasmas, and guide NSTX operations for optimized performance. While most of these diagnostics involve relatively straightforward techniques requiring little or no research and development, and rely heavily on reusing existing equipment, a few pose considerable challenges due to the NSTX geometry. Magnetic diagnostics and visible discharge imaging will be the primary systems for plasma control and equilibrium. Confinement and transport studies will be supported by electron temperature, density, and impurity profile diagnostics. Scrapeoff layer and divertor physics will be investigated with spectroscopic and bolometric diagnostics, thermocouples, and edge Langmuir probes. X-ray fluctuation diagnostics and magnetic pickup coils will be provided for MHD stability analysis. The latter impose special difficulties for NSTX, because of the high temperature environment and the very limited space on the center stack. This article provides an overview of the diagnostics planned for the first phase of NSTX, and highlights the design features needed to satisfy the difficult constraints the device imposes on magnetic diagnostics.

If the ion Bernstein wave (IBW) heating power in an H mode discharge of the PBX-M experiment exceeds a threshold power of about 200 kW, a core transport barrier is created in the central region of the plasma. At lower neutral beam injection (NBI) powers, the core barrier is accompanied by an edge L mode. The high edge localized mode (ELM) repetition frequency (1 kHz) prevents the creation of a strong barrier, so the edge first has to make an H-to-L transition before a strong core transport barrier can be created. At higher NBI powers, the ELM repetition frequency is lowered to less than 200 Hz, which allows the immediate creation of a strong core barrier. Edge localized mode loss, which propagates radially first on a fast (non-diffusive) and then on a slow (diffusive) time-scale all the way to the plasma core, is strongly reduced in the core barrier region. Correlated with the reduced ELM loss, the fluctuations in the core barrier region are also strongly reduced, both during the ELM and during the quiet periods between the ELMs. There is strong evidence that the IBW induced poloidal flow shear is responsible for the stabilization of core turbulence and the creation of the core transport barrier. The large perpendicular E ? B flow shear component of the measured toroidal velocity in co-injection neutral beam heated discharges seems to be largely cancelled by the ion diamagnetic drift shear produced by large ion pressure gradients in the core barrier region. The value of IBW induced poloidal flow has not been experimentally determined, but its numerical value is found to be a factor of 4 larger than either the toroidal velocity or the ion diamagnetic drift shear components, leaving only IBW induced flow shear as the most probable cause for the turbulence stabilization. The core turbulence suppression and the creation of the core transport barrier is also consistent with expectations from a comparison between the E ? B flow shear rate and a rough estimate of the linear ion temperature gradient (ITG) growth rate. The presence of the core barrier region also strongly modifies the other MHD events: crashes on the q = 1.5, 2 surfaces and the disruption.

Two new types of high frequency (HF) MHD mode activity are observed when a core transport barrier is formed during ion Bernstein wave (IBW) heating in PBX-M. These modes, localized in the core transport barrier, seem to be mainly responsible for the soft saturation in the strength of the core transport barrier. The two types of HF mode activity are determined by the presence or absence of a strong 1/1 mode. The n = m modes are observed with the 1/1 mode, while the n = m - 1 modes are seen when the 1/1 mode is absent. The m numbers of both of these modes vary between 4 and 13. They all seem to be coupled and rotate as a rigid body. The n = m - 1 modes seemed to be narrowly localized on their corresponding rational surface, i.e. q = m/(m - 1). The modes are ballooning-like, and their amplitudes increase with increasing β. Calculations with the CASTOR code show that the n = m - 1 modes are most probably pressure driven tearing modes, but the possibility of kinetic ballooning modes (KBMs) cannot be fully excluded. The n = m modes appear on or around the q = 1 surface, and are localized at the X point of the 1/1 mode island. The frequency offset of the n = m modes, which is too low for a TAE, is shown to be consistent with β-induced toroidal Alfven eigenmodes (BAEs).

The Princeton Beta Experiment-Modification (PBX-M) was run both with elliptical and with bean-shaped plasmas during the 1992 and 1993 operating periods. Two deuterium-fed neutral beams were used for auxiliary heating, and during 1992 the average power was . This will be referred to as the lower neutral-beam power (LNBP) period. As many as four deuterium-fed neutral beams were used during 1993, and the average power was . This will be referred to as the medium neutral-beam power (MNBP) period. The neutron source strength, Sn, showed a scaling with injected power , for both the LMBP and MNBP periods. A much wider range of shaping parameters was studied during the MNBP as compared with the LNBP period. A weak positive dependence on bean shaping was observed for the LNBP, and a stronger positive dependence on shaping was observed for MNBP, viz . High values of Sn were obtained in bean-shaped plasmas for the highest values of at for the LNBP. For the MNBP the highest values of Sn and stored energy were obtained at , and the highest values of were obtained at . The achievement of high Sn is aided by high neutral-beam power, high toroidal field, strong shaping, high electron temperature, and broad profiles. The achievement of high is aided by low toroidal field, high density, less shaping, broad profiles, and access to the H-mode, viz . The achievement of high is aided by strong shaping, high density, broad profiles, and access to the H-mode, viz . Some comparisons with the previous higher neutral-beam (HNBP) period in 1989 are also made.