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Coherent magnetohydrodynamic modes have been observed previously during neutral beam injection in the PDX tokamak [Phys. Rev. Lett. 50, 891 (1983)] and they have now been seen in the TFTR tokamak [Phys. Fluids 26, 2958 (1983)]. Periodic bursts of oscillations were detected with several plasma diagnostics, and Fokker–Planck calculations show that the populations of trapped particles in both tokamaks are sufficient to account for fishbone destabilization if a resonant interaction, between the mode and the beam ions, is assumed. Estimates of mode parameters are in reasonable agreement with the experiments, and they indicate that the fishbone mode may continue to affect the performance of intensely heated tokamaks.

Measurements of the toroidal rotation speed vφ(r) driven by neutral beam injection in tokamak plasmas and, in particular, simultaneous profile measurements of vφ, Ti, Te, and ne, have provided new insights into the nature of anomalous transport in tokamaks. Low‐recycling plasmas heated with unidirectional neutral beam injection exhibit a strong correlation among the local diffusivities, χφ≊χi>χe. Recent measurements have confirmed similar behavior in broad‐density L‐mode plasmas. These results are consistent with the conjecture that electrostatic turbulence is the dominant transport mechanism in the tokamak fusion test reactor tokamak (TFTR) [Phys. Rev. Lett. 58, 1004 (1987)], and are inconsistent with predictions both from test‐particle models of strong magnetic turbulence and from ripple transport. Toroidal rotation speed measurements in peaked‐density TFTR ‘‘supershots’’ with partially unbalanced beam injection indicate that momentum transport decreases as the density profile becomes more peaked. In hi...

Magnetohydrodynamic (MHD) activity within three zones (core, half‐radius, and edge) of TFTR [Plasma Physics and Controlled Nuclear Fusion Research 1986 (IAEA, Vienna, 1987), Vol. 1, p. 51] tokamak plasmas are discussed. Near the core of the plasma column, sawteeth are often observed. Two types of sawteeth are studied in detail; one with complete, and the other with incomplete, magnetic reconnection. Their characteristics are determined by the shape of the q profile. Near the half‐radius the m/n=3/2 and 2/1 resistive ballooning modes are found to correlate with a beta collapse. The pressure and the pressure gradient at the mode rational surface are found to play an important role in stability. MHD activity is also studied at the plasma edge during limiter H modes. The edge localized modes (ELM’s) are found to have a precursor mode with a frequency between 50–200 kHz and a mode number m/n=1/0. The mode does not show a ballooning structure. While these instabilities have been studied on many other machines, ...

The first-ever successful operation of a tokamak with a large area (40% of the total plasma surface area) liquid lithium wall has been achieved in the Lithium Tokamak eXperiment (LTX). These results were obtained with a new, electron beam-based lithium evaporation system, which can deposit a lithium coating on the limiting wall of LTX in a five-minute period. Preliminary analyses of diamagnetic and other data for discharges operated with a liquid lithium wall indicate that confinement times increased by 10× compared to discharges with helium-dispersed solid lithium coatings. Ohmic energy confinement times with fresh lithium walls, solid and liquid, exceed several relevant empirical scaling expressions. Spectroscopic analysis of the discharges indicates that oxygen levels in the discharges limited on liquid lithium walls were significantly reduced compared to discharges limited on solid lithium walls. Tokamak operations with a full liquid lithium wall (85% of the total plasma surface area) have recently started.

The spherical torus or spherical tokamak (ST) is a member of the tokamak family with its aspect ratio (A = R0/a) reduced to A ∼ 1.5, well below the normal tokamak operating range of A ≥ 2.5. As the aspect ratio is reduced, the ideal tokamak beta β (radio of plasma to magnetic pressure) stability limit increases rapidly, approximately as β ∼ 1/A. The plasma current it can sustain for a given edge safety factor q-95 also increases rapidly. Because of the above, as well as the natural elongation κ, which makes its plasma shape appear spherical, the ST configuration can yield exceptionally high tokamak performance in a compact geometry. Due to its compactness and high performance, the ST configuration has various near term applications, including a compact fusion neutron source with low tritium consumption, in addition to its longer term goal of an attractive fusion energy power source. Since the start of the two mega-ampere class ST facilities in 2000, the National Spherical Torus Experiment in the United St...

In tokamak plasmas, a key element in the understanding of magnetohydrodynamic (MHD) instabilities is the determination of their toroidal and poloidal mode numbers. This is usually accomplished by correlating measurements of radiation, particles, or magnetic oscillations at separate positions toroidally and poloidally around the torus. It is possible, however, to obtain the same information with arrays of soft x-ray detectors viewing the horizontal midplane and the poloidal cross section of the tokamak from a single location. The perturbation due to an MHD mode on a toroidal shell will modulate the soft x-ray emission along sightlines intersecting it. The number of maxima, and the angular separation between the channels detecting them, have a straightforward geometrical relationship to the mode number. This approach thus avoids the need for complex inversion of line-integrated signals. The method has already been used to determine poloidal mode numbers on the Princeton beta experiment-modification tokamak,...

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

Edge localized modes (ELMs), edge turbulence and other fluctuations in the edge region of NSTX plasma are observed by fast cameras. ELMs categorized as Type I, III, and V are found in NSTX and other MARFE-like activity is seen near the center stack in double null plasmas. Edge turbulence filaments are almost always found in NSTX. An interpretation of ELM images, which shows a complex ELMs structure, is presented

Summary form only given. We are developing large pixel count, fast (/spl ges/100 kHz) and continuously sampling soft X-ray (SXR) arrays for the diagnosis of MHD and turbulent fluctuations in MFE experiments. The arrays are based on efficient scintillators, high throughput multi-clad fiber optics, and multi-channel light amplification and integration. Compared to conventional X-ray diode arrays, such systems can provide vastly increased spatial coverage (e.g. 2-D tangential imaging), access to difficult locations and potentially circumvent the neutron damage encountered with typical X-ray diodes. A prototype eight-channel array is being built using columnar CsI:Tl deposited on a fiber optic faceplate as an SXR converter and a multi-anode PMT and APD as photoamplifiers. Columnar CsI:Tl was found to have conversion efficiency superior to that of amorphous depositions. The overall system efficiency is measured using laboratory SXR sources, while the time response and signal-to-noise performance are evaluated by recording MHD activity from the Current Drive Experiment Upgrade (CDX-U) and the National Spherical Torus Experiment (NSTX) at PPPL.

Recent experiments in the Current Drive Experiment-Upgrade provide a first-ever test of large area liquid lithium surfaces as a tokamak first wall, to gain engineering experience with a liquid metal first wall, and to investigate whether very low recycling plasma regimes can be accessed with lithium walls. The CDX-U is a compact (R = 34 cm, a = 22 cm, B{sub toroidal} = 2 kG, I{sub P} = 100 kA, T{sub e}(0) = 100 eV, n{sub e}(0) {approx} 5 x 10{sup 19} m{sup -3}) spherical torus at the Princeton Plasma Physics Laboratory. A toroidal liquid lithium tray limiter with an area of 2000 cm{sup 2} (half the total plasma limiting surface) has been installed in CDX-U. Tokamak discharges which used the liquid lithium limiter required a fourfold lower loop voltage to sustain the plasma current, and a factor of 5-8 increase in gas fueling to achieve a comparable density, indicating that recycling is strongly reduced. Modeling of the discharges demonstrated that the lithium-limited discharges are consistent with Z{sub effective} < 1.2 (compared to 2.4 for the pre-lithium discharges), a broadened current channel, and a 25% increase in the core electron temperature. Spectroscopic measurements indicate that edge oxygen and carbon radiationmore » are strongly reduced.« less

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 )

Part of the development of liquid metals as a first wall or divertor for reactor applications must involve the investigation of plasma-liquid metal interactions in a functioning tokamak. Most of the interest in liquid-metal walls has focused on lithium. Experiments with lithium limiters have now been conducted in the Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory. Initial experiments used a liquid-lithium rail limiter (L3) built by the University of California at San Diego. Spectroscopic measurements showed some reduction of impurities in CDX-U plasmas with the L3, compared to discharges with a boron carbide limiter. While no reduction in recycling was observed with the L3, which had a plasma-wet area of approximately 40 cm2, subsequent experiments with a larger area fully toroidal lithium limiter demonstrated significant reductions in both recycling and in impurity levels. Two series of experiments with the toroidal limiter have now be en performed. In each series, the area of exposed, clean lithium was increased, until in the latest experiments the liquid-lithium plasma-facing area was increased to 2000 cm2. Under these conditions, the reduction in recycling required a factor of eight increase in gas fueling in order to maintain the plasma density. The loop voltage required to sustain the plasma current was reduced from 2 V to 0.5 V. This paper summarizes the technical preparations for lithium experiments and the conditioning required to prepare the lithium surface for plasma operations. The mechanical response of the liquid metal to induced currents, especially through contact with the plasma, is discussed. The effect of the lithium-filled toroidal limiter on plasma performance is also briefly described.

Use of a large-area liquid lithium surface as a first wall 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 important issue for future tokamaks is real‐time plasma control for the avoidance of magnetohydrodynamic instabilities and other applications that require detailed plasma profile and fluctuation data. Although measurements from diagnostics providing this information require significantly more processing than magnetic flux data, recent advancements could make them practical for adjusting operational settings for plasma heating and current drive systems as well as field coil currents. On the Princeton Beta Experiment‐Modification (PBX‐M), the lower hybrid current drive phasing can be varied during a plasma shot using digitally programmable ferrite phase shifters, and neural beam functions can be fully computer controlled. PBX‐M diagnostics that may be used for control purposes include motional Stark‐effect polarimetry for magnetic field pitch angle profiles, soft x‐ray arrays for plasma position control and the separation of βp from li, hard x‐ray detectors for energetic electron distributions, a multich...

Electron density profile measurements have been obtained on PBX‐M by amplitude modulation reflectometry. With this technique, the measurement of the time delay is obtained from the phase delay of the modulating envelope (200 MHz in PBX‐M). The system operates with the extraordinary mode, in the range 32–50 GHz. Under these conditions the density profile of PBX‐M can be probed from the scrape‐off layer up to typically r/a=0.7. With a final bandwidth of 40 kHz, the reflectometer is able to obtain the edge profile in 1 ms. The profiles obtained are relatively noise free and in good agreement with Thomson scattering measurements. Perturbations due to the strong‐edge turbulence are kept to a minimum, and no software filtering or other signal processing was necessary to extract the time delay information from the raw data. Profiles have been measured for ohmic and Neutron Beam Injection heated discharges.

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.

A fast X-ray system based on a Micro Pattern Gas Detector has been used, for the first time, to investigate emission from the plasma core of the National Spherical Tokamak eXperiment (NSTX) at the Princeton Plasma Physics Laboratory. The results presented in this work demonstrate the capability of such a device to measure with a time resolution of the order of 1 ms the curvature and the elongation of the X-ray iso-emissivity contours, under various plasma conditions. Also, comparisons with the magnetic surface structure calculated by the EFIT code show good agreement between reconstructed flux surface and the soft X-ray emissions (SXR) for poloidal beta values up to 0.6. For greater values of beta, X-ray iso-emissivity contours become circular, while magnetic flux surface reconstructions yield elongation 1.5 < k < 2.2. The X-ray images have been acquired with a (statistical) signal to noise ratio (SNR) per pixel of about 30. Thanks to the direct and efficient X-ray conversion and its operation in a photon counting mode, this new diagnostic tool allows the routine investigation of the plasma core with a sampling rate of 1 kHz and extremely high SNR under all experimental conditions in NSTX.

The Rogowski loop is one of the most basic diagnostics for tokamak operations. On the National Spherical Torus Experiment (NSTX), the plasma current Rogowski Loop had constraints of very limited space available on the center stack, 5000 volt isolation, flexibility requirements as it remained a part of the Center Stack assembly after the first phase of operation, and with a /spl plusmn/120/spl deg/C temperature requirement. For the second phase of operation, four halo current Rogowski loops under the Center Stack tiles will be installed having +600/spl deg/C and limited space requirements. Also as part of the second operational phase, up to ten Rogowski Loops will installed to measure eddy currents in the passive plate support structures with +350 /spl deg/C, restricted space, and flexibility requirements. This presentation will provide the details of the material selection, fabrication techniques, testing, and installation results of the Rogowski loops that were fabricated for the high temperature operation and bakeout requirements, high voltage isolation requirements, and the space and flexibility requirements imposed upon the Rogowski loops. In the future operational phases of NSTX, additional Rogowski loops could be anticipated that will measure toroidal plasma currents in the vacuum vessel and in the passive plate assemblies.

A high-speed, 80-kV, self-triggering crowbar has been constructed to protect the Princeton Beta Experiment-Modified (PBX-M) diagnostic neutral probe beam (NPB) ion source from arc damage and assist in rapid high-voltage conditioning. The crowbar, operating in conjunction with two magnetic-core snubbers, quickly limits and diverts ion-source arc current. The charge flowing in the ion-source arc is reduced an order of magnitude below the level expected with only snubber protection. The crowbar firing time is less than 100 ns at an 80-kV operating level. >

Third harmonic hydrogen cyclotron fast wave heating studies are planned in the near term on CDX-U to investigate the potential for bulk ion heating. In preparation for these studies, the available radio-frequency power in CDX-U has been increased to 0.5 MW. The operating frequency of the CDX-U radio-frequency transmitter was lowered to operate in the range of 8-10 MHz, providing access to the ion harmonic range 2* {approx} 4* in hydrogen. A similar regime is accessible for the 30 MHz radio-frequency system on the National Spherical Torus Experiment (NSTX), at 0.6 Tesla in hydrogen. Preliminary computational studies over the plasma regimes of interest for NSTX and CDX-U indicate the possibility of strong localized absorption on bulk ion species.

The PBX-M passive plate stabilization system provides significant stabilization of long-wavelength external kink modes, the slowing of vertical instability growth rates, and the amelioration of disruption characteristics. The passive plate stabilization system has allowed the use of LHCD and IBW to induce current density and pressure profile modifications, and m=1 divertor biasing for modifying edge plasma transport. Improvements in the passive plate system insulators and support structures have provided reliable operation. Impurity influxes with the close-fitting passive plates are low. Solid target boronization is applied routinely to reduce conditioning time and maintain clean conditions.

The NSTX project will provide a national facility for the study of plasma confinement, heating, and current drive in a low aspect ratio, spherical torus (ST) configuration, the ST configuration is an alternate confinement concept which is characterized by high /spl beta/, high elongation, high bootstrap fraction, and low B/sub T/ compared to conventional tokamaks, NSTX is the next step ST experiment following smaller experiments such as the PPPL CDX-U (Current Drive Experiment, Upgrade), the START (Small Tight Aspect Ratio Tokamak) at Culham, and the HIT (Helicity Injected Tokamak) at University of Washington, and is similar in scale to the MAST (Meg-Amp Spherical Tokamak) machine now under construction at Culham. This paper provides a description of the mission and gives an overview of the engineering features of the design of the machine and facility and discusses some of the key design solutions.

PBX-M upgrades have been in progress for enhanced capability and reliability. These include upgrades to the vessel O-ring system for impurity control, the Ion Bernstein Wave and Lower Hybrid Current Drive Systems for plasma profile control, a double viewing upper and lower CHERS based Poloidal Rotation Diagnostic, and a new digital plasma control system.

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) at Princeton 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 5 MW of neutral beam injection, NBI, at an injection energy of 90 keV and up to 6 MW of high harmonic fast wave, HHFW, at 30 MHz. NSTX has achieved beta T of 32%. A variety of MHD phenomena have been observed to limit eta. NSTX has now begun addressing E scaling, eta limits and current drive issues. During the NBI heating experiments, a broad Ti profile with Ti up to 2 keV, Ti > Te and a large toroidal rotation. Transport analysis suggests that the impurity ions have diffusivities approaching neoclassical. For L-Mode plasmas, E is up to two times the ITER-89P L-Mode scaling and exceeds the ITER-98pby2 H-Mode scaling in some cases. Transitions to H-Mode have been observed which result in an approximate doubling of tE. after the transition in some conditions. During HH FW heating, Te > Ti and Te up to 3.5 keV were observed. Current drive has been studied using coaxial helicity injection (CHI), which has produced 390 kA of toroidal current and HHFW, which has produced H-modes with significant bootstrap current fraction at low Ip, high q and high{sub etap}.

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.

Liquid metal walls have the potential solve to first-wall problems for fusion reactors, such as heat load and erosion of dry walls, neutron damage and activation, and tritium inventory and breeding. In the near term, such walls can serve as the basis for schemes to stabilize magnetohydrodynamic (MHD) modes. Furthermore, the low recycling characteristics of lithium walls can be used for particle control. Liquid lithium experiments have already begun in the Current Drive eXperiment-Upgrade (CDX-U). Plasmas limited with a toroidally localized limiter have been investigated, and experiments with a fully toroidal lithium limiter are in progress. A liquid surface module (LSM) has been proposed for the National Spherical Torus Experiment (NSTX). In this larger ST, plasma currents are in excess of 1 MA and a typical discharge radius is about 68 cm. The primary motivation for the LSM is particle control, and options for mounting it on the horizontal midplane or in the divertor region are under consideration. A key consideration is the magnitude of the eddy currents at the location of a liquid lithium surface. During plasma start up and disruptions, the force due to such currents and the magnetic field can force a conducting liquid off of the surface behind it. The Tokamak Simulation Code (TSC) has been used to estimate the magnitude of this effect. This program is a two dimensional, time dependent, free boundary simulation code that solves the MHD equations for an axisymmetric toroidal plasma. From calculations that match actual ST equilibria, the eddy current densities can be determined at the locations of the liquid lithium. Initial results have shown that the effects could be significant, and ways of explicitly treating toroidally local structures are under investigation.

In the CDX-U spherical torus, approximately 100% conversion of thermal EBWs to X-mode has been observed by controlling the electron density scale length (Ln) in the conversion region with a local limiter outside the last closed flux surface. The radiation temperature profile agrees with Thomson scattering electron temperature data. Results are consistent with theoretical calculations of conversion efficiency using measured Ln. By reciprocity of the conversion process, prospects for efficient coupling in EBW heating and current drive scenarios are strongly supported.

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.

Electron transport has been measured in the Current Drive Experiment-Upgrade (CDX-U) using two separate perturbative techniques. Gas modulation at the plasma edge was used to introduce cold-pulses which propagate towards the plasma center, providing time-of-flight information leading to a determination of chi(subscript e) as a function of radius. Sawteeth at the q=1 radius (r/a {approx} 0.15) induced heat-pulses which propagated outward towards the plasma edge, providing a complementary time-of-flight based chi(subscript e) profile measurement. This work represents the first localized measurement of chi(subscript e) in a spherical torus. It is found that chi(subscript e) = 1-2 meters squared per second in the plasma core (r/a < 1/3), increasing by an order of magnitude or more outside of this region. Furthermore, the chi(subscript e) profile exhibits a sharp transition near r/a = 1/3. Spectral and profile analyses of the soft X-rays, scanning interferometer, and edge probe data show no evidence of a significant magnetic island causing the high chi(subscript e) region.

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.

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 that any CHI discharges previously produced in a Spheromak or a Spherical Torus (ST).

The concept of a flowing lithium first wall for a fusion reactor may lead to a significant advance in reactor design, since it could virtually eliminate the concerns with power density and erosion, tritium retention, and cooling associated with solid walls. As part of investigations to determine the feasibility of this approach, plasma interaction questions in a toroidal plasma geometry are being addressed in the Current Drive eXperiment-Upgrade (CDX-U) spherical torus (ST). The first experiments involved a toroidally local lithium limiter (L3). Measurements of pumpout rates indicated that deuterium pumping was greater for the L3 compared to conventional boron carbide limiters. The difference in the pumpout rates between the two limiter types decreased with plasma exposure, but argon glow discharge cleaning was able to restore the pumping effectiveness of the L3. At no point, however, was the extremely low recycling regime reported in previous lithium experiments achieved. This may be due to the much larger lithium surfaces that were exposed to the plasma in the earlier work. The possibility will be studied in the next set of CDX-U experiments, which are to be conducted with a large area, fully toroidal lithium limiter.

Mode-converted electron Bernstein waves (EBWs) potentially allow the measurement of local electron temperature (Te) and the implementation of local heating and current drive in spherical torus (ST) devices, which are not directly accessible to low harmonic electron cyclotron waves. This paper reports on the measurement of X-mode radiation mode-converted from EBWs observed normal to the magnetic field on the midplane of the Current Drive Experiment-Upgrade (CDX-U) and the National Spherical Torus Experiment (NSTX) spherical torus plasmas. The radiation temperature of the EBW emission was compared to Te measured by Thomson scattering and Langmuir probes. EBW mode-conversion efficiencies of over 20% were measured on both CDX-U and NSTX. Sudden increases of mode-conversion efficiency, of over a factor of three, were observed at high-confinement-mode transitions on NSTX, when the measured edge density profile steepened. The EBW mode-conversion efficiency was found to depend on the density gradient at the mode-conversion layer in the plasma scrape-off, consistent with theoretical predictions. The EBW emission source was determined by a perturbation technique to be localized at the electron cyclotron resonance layer and was successfully used for radial transport studies. Recently, a new in-vessel antenna and Langmuir probe array were installed on CDX-U to better characterize and enhance the EBW mode-conversion process. The probe incorporates a local adjustable limiter to control and maximize the mode-conversion efficiency in front of the antenna by modifying the density profile in the plasma scrape-off where fundamental EBW mode conversion occurs. Initial results show that the mode-conversion efficiency can be increased to {approximately}100% when the local limiter is inserted near the mode-conversion layer. Plans for future EBW research, including EBW heating and current-drive studies, are discussed.

We discuss the fluctuation characteristics expected from MHD modes including ballooning modes and show how they can be studied using measurements of ECE. A novel technique, ECE correlation radiometry, which enables the location and spatial structure of long wavelength (L≳0.1×a) MHD modes to be determined, is described. Measurements on JET with a 44‐channel ECE heterodyne radiometer are presented and show the existence of high‐frequency (high n number) MHD modes under high poloidal beta, pellet‐enhanced performance (PEP) plasma conditions. Similar measurements are planned for PBX‐M and the measurement system is described.

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. where the plasma frequency is much greater than the electron cyclotron frequency, as in a spherical torus) or they have insufficient density and temperature to reach the blackbody condition. 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 perpendicular wavenumber. This paper reports on measurements of EBW emission on the CDX-U spherical torus, where B{sub o} {approximately} 2 kG, {approximately}10{sup 13} cm{sup {minus}3} and T{sub e} {approx} to 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 measured by a multi-point Thomson scattering diagnostic. Depending on the plasma conditions, the mode converted EBW radiation temperature was found to be less than or equal to T{sub e} and the emission source was determined to be radially localized at the electron cyclotron resonance layer. A Langmuir triple probe and a 140 GHz interferometer were employed to measure changes in edge density profile in the vicinity of the upper hybrid resonance, where the mode conversion of the EBWs is expected to occur. Initial results suggest EBW emission and EBW heating are viable concepts for overdense plasmas.

The Rogowski Loop is one of the most basic diagnostics for tokamak operations. On the National Spherical Torus Experiment (NSTX), the plasma current Rogowski Loop had the constraints of the very limited space available on the center stack, 5,000 volt isolation, flexibility requirements as it remained a part of the Center Stack assembly after the first phase of operation, and a +120 C temperature requirement. For the second phase of operation, four Halo Current Rogowski Loops under the Center Stack tiles will be installed having +600 C and limited space requirements. Also as part of the second operational phase, up to ten Rogowski Loops will installed to measure eddy currents in the Passive Plate support structures with +350 C, restricted space, and flexibility requirements. This presentation will provide the details of the material selection, fabrication techniques, testing, and installation results of the Rogowski Loops that were fabricated for the high temperature operational and bakeout requirements, high voltage isolation requirements, and the space and flexibility requirements imposed upon the Rogowski Loops. In the future operational phases of NSTX, additional Rogowski Loops could be anticipated that will measure toroidal plasma currents in the vacuum vessel and in the Passive Plate assemblies.

Diagnostics are described which are currently installed or under active development for the newly commissioned NSTX device. The low aspect ratio (R/a less than or equal to 1.3) and low toroidal field (0.1-0.3T) used in this device dictate adaptations in many standard diagnostic techniques. Technical summaries of each diagnostic are given, and adaptations, where significant, are highlighted.

The design of the more than 300 in-vessel sensor systems for the National Spherical Torus Experiment (NSTX) has encountered several challenging fusion reactor diagnostic issues involving high temperatures and space constraints. This has resulted in unique miniature, high temperature in-vessel sensor systems mounted in small spaces behind plasma facing armor tiles, and they are prototypical of possible high power reactor first-wall applications. In the Center Stack, Divertor, Passive Plate, and vessel wall regions, the small magnetic sensors, large magnetic sensors, flux loops, Rogowski Coils, thermocouples, and Langmuir Probes are qualified for 600 degrees C operation. This rating will accommodate both peak rear-face graphite tile temperatures during operations and the 350 degrees C bake-out conditions. Similar sensor systems including flux loops, on other vacuum vessel regions are qualified for 350 degrees C operation. Cabling from the sensors embedded in the graphite tiles follows narrow routes to exit the vessel. The detailed sensor design and installation methods of these diagnostic systems developed for high-powered ST operation are discussed.

The concept of a flowing lithium first wall for a fusion reactor may lead to a significant advance in reactor design, since it could virtually eliminate the concerns with power density and erosion, tritium retention, and cooling associated with solid walls. Sputtering and erosion tests are currently underway in the PISCES device at the University of California at San Diego (UCSD). To complement this effort, plasma interaction questions in a toroidal plasma geometry will be addressed by a proposed new groundbreaking experiment in the Current Drive eXperiment-Upgrade (CDX-U) spherical torus (ST). The CDX-U plasma is intensely heated and well diagnosed, and an extensive liquid lithium plasma-facing surface will be used for the first time with a toroidal plasma. Since CDX-U is a small ST, only approximately1 liter or less of lithium is required to produce a toroidal liquid lithium limiter target, leading to a quick and cost-effective experiment.

Time dependent calculations of lower hybrid (LH) current profile control in Alcator C-Mod have been done using the TRANSP [1], FPPRF [2], and LSC [3] codes. Up to 3 MW of LH current drive power was applied in plasmas with high power ICRF minority heating (PICH=1.8−3 MW) and fast current ramp up. Using the experimentally measured temperature profiles, off-axis current generation resulted in nonmonotonic q-profiles with qmin≃1.6. Self-consistent effects of off-axis electron heating by the LH power were also included in the analysis and significant broadening of the electron temperature profile was found with qmin≳2 and a larger shear reversal radius.

Most magnetically confined plasma devices cannot take advantage of standard Electron Cyclotron Emission (ECE) diagnostics to measure temperature. They either operate at high density relative to their magnetic field or they do not have sufficient density and temperature to reach the blackbody condition. The standard ECE technique measures the electromagnetic waves emanating from the plasma. Here we propose to measure electron Bernstein waves (EBW) to ascertain the local electron temperature in these plasmas. The optical thickness of EBW is extremely high because it is an electrostatic wave with a large k(subscript i). One can reach the blackbody condition with a plasma density approximately equal to 10(superscript 11) cm(superscript -3) and electron temperature approximately equal to 1 eV. This makes it attractive to most plasma devices. One serious issue with using EBW is the wave accessibility. EBW may be accessible by either direct coupling or mode conversion through an extremely narrow layer (approximately 1-2 mm) in low field devices.