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Objective. Even though the techniques used for breast cancer identification have advanced over the years, current mammography based on x-rays absorption, the ‘gold standard’ screening test at present, still has some shortcomings as concerns sensitivity and specificity to early-stage cancers, due to poor differentiation between tumor and normal tissues, especially in the case of the dense breasts. We investigate a possible additional technique for breast cancer detection with higher sensitivity and low dose, x-ray phase-contrast or refraction-based imaging with ultrahigh angular sensitivity grating interferometers, having several meters length. Approach. Towards this goal, we built and tested on a mammography phantom, a table-top laboratory setup based on a 5.7 m long Talbot-Lau interferometer with angular sensitivity better than 1 μrad. We used a high-power x-ray tungsten anode tube with a 400 μm focal spot, operated at 40 kVp and 15 mA with a 2 mm aluminum filter. Main results. The results reported in our paper confirm the ultrahigh sensitivity and dose economy possible with our setup. The visibility of objects simulating cancerous formations is strongly increased in the refraction images over the attenuation ones, even at a low dose of 0.32 mGy. Notably, the smallest fiber of 400 μm diameter and calcifications specs of 160 μm in diameter are detected, even though the spatial resolution at the object of our magnification M ∼ 2 setup with a 400 μm source spot is only ∼250 μm. Significance. Our experiments on a mammography phantom illustrate the capabilities of the proposed technique and can open the way toward low-dose interferometric mammography.

Ultrahigh sensitivity X-ray interferometry using micron-period gratings in long Talbot-Lau geometries is developed at ELI-NP, towards laser-based X-ray phase contrast imaging for medical applications. Results on mammography phantoms confirm the capabilities of the proposed method.

Talbot–Lau x-ray interferometry is a refraction-based diagnostic that can map electron density gradients through phase-contrast methods. The Talbot–Lau x-ray deflectometry (TXD) diagnostics have been deployed in several high energy density experiments. To improve diagnostic performance, a monochromatic TXD was implemented on the Multi-Tera Watt (MTW) laser using 8 keV multilayer mirrors (Δθ/θ = 4.5%-5.6%). Copper foil and wire targets were irradiated at 1014–1015 W/cm2. Laser pulse length (∼10 to 80 ps) and backlighter target configurations were explored in the context of Moiré fringe contrast and spatial resolution. Foil and wire targets delivered increased contrast

Interferometry is one of the most sensitive and successful diagnostic methods for plasmas. However, owing to the design of most common interferometric systems, the wavelengths of operation and, therefore, the range of densities and temperatures that can be probed are severely limited. Talbot–Lau interferometry offers the possibility of extending interferometry measurements to x-ray wavelengths by means of the Talbot effect. While there have been several proof-of-concept experiments showing the efficacy of this method, it is only recently that experiments to probe High Energy Density (HED) plasmas using Talbot–Lau interferometry are starting to take place. To improve these experimental designs, we present here the Talbot-Interferometry Analyzer (TIA) tool, a forward model for generating and postprocessing synthetic x-ray interferometry images from a Talbot–Lau interferometer. Although TIA can work with any two-dimensional hydrodynamic code to study plasma conditions as close to reality as possible, this software has been designed to work by default with output files from the hydrodynamic code FLASH, making the tool user-friendly and accessible to the general plasma physics community. The model has been built into a standalone app, which can be installed by anyone with access to the MATLAB runtime installer and is available upon request to the authors.

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.

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.

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.

Standard diagnostics for fundamental plasma parameters and for plasma physics are routinely utilized for daily operation and physics studies in the large helical device (LHD) with high reliability. Diagnostics for steady-state plasma are under intensive development, especially for Te, ne (yttrium–aluminium garnet (YAG) laser Thomson, CO2 laser polarimeter), data acquisition in steady-state and heat-resistant probes. To clarify the plasma properties of the helical structure, two- or three-dimensional diagnostics are being aggressively developed: tangential cameras (fast SX TV, photon counting CCD, Hα TV); tomography (tangential SX CCD, bolometer); imaging (bolometer, ECE, reflectometer). Divertor and edge physics are important key issues for steady-state operation. Diagnostics for neutral flux (Hα array, Zeeman spectroscopy) and ne (fast scanning probe, Li beam probe, pulsed radar reflectometer) are also in advanced stages of development. In addition to these, advanced diagnostics are being intensively developed in LHD through domestic and international collaborations.

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.

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.

The authors present initial progress toward a clinically compatible x-ray phase contrast CT system, using glancing-angle x-ray grating interferometry to provide high contrast soft tissue images at estimated by computer simulation dose levels comparable to conventional absorption based CT.DPC-CT scans of a joint phantom and of soft tissues were performed in order to answer several important questions from a clinical setup point of view. A comparison between high and low fringe visibility systems is presented. The standard phase stepping method was compared with sliding window interlaced scanning. Using estimated dose values obtained with a Monte-Carlo code the authors studied the dependence of the phase image contrast on exposure time and dose.Using a glancing angle interferometer at high x-ray energy (∼ 45 keV mean value) in combination with a conventional x-ray tube the authors achieved fringe visibility values of nearly 50%, never reported before. High fringe visibility is shown to be an indispensable parameter for a potential clinical scanner. Sliding window interlaced scanning proved to have higher SNRs and CNRs in a region of interest and to also be a crucial part of a low dose CT system. DPC-CT images of a soft tissue phantom at exposures in the range typical for absorption based CT of musculoskeletal extremities were obtained. Assuming a human knee as the CT target, good soft tissue phase contrast could be obtained at an estimated absorbed dose level around 8 mGy, similar to conventional CT.DPC-CT with glancing-angle interferometers provides improved soft tissue contrast over absorption CT even at clinically compatible dose levels (estimated by a Monte-Carlo computer simulation). Further steps in image processing, data reconstruction, and spectral matching could make the technique fully clinically compatible. Nevertheless, due to its increased scan time and complexity the technique should be thought of not as replacing, but as complimentary to conventional CT, to be used in specific applications.

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

A simplified radiative collapse model has been applied to a high-Z (Ce) low-inductance vacuum spark plasma, in order to assess the possibility of using this simple discharge to generate N-shell atomic quasi-continua for optical pumping of soft x-ray lasers. The radiative losses of the complex Ce ions are computed using statistical methods developed for unresolved transition arrays, and atomic data from the average ion model. The calculations predict that at sufficiently high current, a radiative collapse followed by micropinching takes place at the opening of the M-shell, in good agreement with experimental observations. The model predicts that the radiative losses associated with the quasi-continuum emission allow the plasma to be compressed to relatively high density prior to the micropinch. Nevertheless, despite this `mild' radiative collapse, the average quasi-continuum brightness is below that required for optical pumping. Scaling the results to discharge currents shows this limitation to be intrinsic to Z-pinch type discharges.

Low aspect-ratio tokamaks (LARTs) have intrinsic features which require specific diagnostics for particle transport and confinement studies. Such spectroscopic diagnostics are developed in collaboration between the CDX-U group at Princeton Plasma Physics Laboratory and the Plasma Spectroscopy group at Johns Hopkins University. This paper presents results obtained from the low aspect-ratio CDX-U tokamak, as well as future diagnostic developments for CDX-U and other LARTs. These results are based on spectroscopy in the VUV (500 A - 2000 A) using a multiple grating normal incidence spectrometer equipped with an optical multichannel analyzer, and in the XUV range on emission recorded using multilayer mirror devices.

We attempt to interpret the relative line intensities in the soft X-ray spectrum of Al III ions emitted from a Penning discharge, on the assumption that the 2p53s2 and 2p5(3s3p 3P) autoionizing levels are excited predominantly by inner-shell ionization from Al II ground and metastable levels, respectively. The plasma has been spectroscopically diagnosed and the Al III soft X-ray spectrum predicted using a collisional radiative model incorporating the experimental plasma parameters and theoretical predictions for the properties of the autoionizing levels. The model includes collisional inner-shell ionization from Al II levels by fast electrons, inelastic collisions with cold Maxwellian electrons and radiative and autoionization decays. The comparison between the experimental and the predicted relative line intensities seems to confirm that most of the 2p5(3s3p 3P) Al III levels have large radiative branching ratios. This conclusion is supported also by a comparison between Al III and Mg II spectra obtained in similar excitation conditions. Several 2p63p 2P-2p5(3s3p 3P)2P, 2D Al III transitions having radiative yields close to one and large gain cross sections, may be of interest for soft X-ray lasing.

The tokamak scrape-off (the region between the vacuum vessel wall and the magnetically confined fusion plasma edge), represents a source/sink for the hot fusion plasma. The electron densities and temperatures are in the ranges 10 to the 11th - 10 to the 13th/cu cm and 1-40 eV, respectively (depending on the size, magnetic field intensity and configuration, plasma current, etc). In the work reported, the electron temperature and density have been estimated in a Penning ionization discharge by comparing its spectroscopic emission in the VUV with that predicted by a collisional radiative model. An attempt to directly compare this emission with that of the tokamak edge is briefly described.

Observations of magneto-hydro-dynamic (MHD) induced redistribution or loss of energetic ions measured using the vertically scanning capability of the Neutral Particle Analyzer diagnostic on the National Spherical Torus Experiment (NSTX) are presented along with TRANSP and ORBIT code analysis of the results. Although redistribution or loss of energetic ions due to bursting fishbone-like and low-frequency (f ~ 10 kHz) kinktype MHD activity has been reported previously, the primary goal of this work is to study redistribution or loss due to continuous Alfvenic (f ~ 20 – 150 kHz) modes, a topic that heretofore has not been investigated in detail for NSTX plasmas. Initial indications are that the former drive energetic ion loss whereas the continuous Alfvenic modes only cause redistribution and the energetic ions remain confined.

A fast soft x-ray (SXR) pinhole camera has been implemented on the National Spherical Torus Experiment (NSTX). This paper presents observations and describes the Be foil Filter Knee Imaging (FKI) technique for reconstructions of a m/n=1/1 mode on NSTX. The SXR camera has a wide-angle (28{sup o}) field of view of the plasma. The camera images nearly the entire diameter of the plasma and a comparable region in the vertical direction. SXR photons pass through a beryllium foil and are imaged by a pinhole onto a P47 scintillator deposited on a fiber optic faceplate. An electrostatic image intensifier demagnifies the visible image by 6:1 to match it to the size of the charge-coupled device (CCD) chip. A pair of lenses couples the image to the CCD chip.

Saturated internal kink modes have been observed in many of the highest toroidal {beta} discharges of the National Spherical Torus Experiment (NSTX). These modes often cause rotation flattening in the plasma core, can degrade energy 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-{beta} may contribute to mode nonlinear stabilization. These saturation mechanisms are investigated for NSTX parameters and compared to experimental data.

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 NSTX spherical torus (ST) provides a unique platform to investigate magnetic confinement in auxiliary-heated plasmas at low aspect ratio. Auxiliary power is routinely coupled to ohmically heated plasmas by deuterium neutral-beam injection (NBI) and by high-harmonic fast waves (HHFW) launch. While theory predicts both techniques to preferentially heat electrons, experiment reveals the electron temperature is greater than the ion temperature during HHFW, but the electron temperature is less than the ion temperature during NBI. In the following we present the experimental data and the results of transport analyses.

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) [Ono, M., Kaye, S.M., Neumeyer, S., et al., Proceedings, 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque, 1999, (IEEE, Piscataway, NJ (1999), p. 53.)] is such a device. An radio-frequency (rf) heating system has been installed on 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 explore the energy transport properties of NSTX plasmas, to create H-mode (high-confinement mode) discharges with a large fraction of bootstrap current and to control the plasma current profile during the early stages of the discharge.

Ideal magnetohydrodynamic stability limits of shaped tokamak plasmas with high bootstrap fraction are systematically determined as a function of plasma aspect ratio. For plasmas with and without wall stabilization of external kink modes, the computed limits are well described by distinct and nearly invariant values of a normalized beta parameter utilizing the total magnetic field energy density inside the plasma. Stability limit data from the low aspect ratio National Spherical Torus Experiment is compared to these theoretical limits and indicates that ideal nonrotating plasma no-wall beta limits have been exceeded in regimes with sufficiently high cylindrical safety factor. These results could impact the choice of aspect ratio in future fusion power plants.

We present results from the first experiments with a large area liquid lithium limiter in a magnetic fusion device, and its effect on improving plasma performance by reducing particle recycling. Using large area liquid metal surfaces in any major fusion device is unlikely before a test on a smaller scale. This has motivated its demonstration in the CDX-U spherical torus with a unique, fully toroidal lithium limiter. The highest current discharges were obtained with a liquid lithium limiter. There was a reduction in recycling, as indicated by a significant decrease in the deuterium-alpha emission and oxygen radiation. How these results might extrapolate to reactors is suggested in recycling/retention experiments with liquid lithium surfaces under high-flux deuterium and helium plasma bombardment in PISCES-B. Data on deuterium atoms retained in liquid lithium indicate retention of all incident ions until full volumetric conversion to lithium deuteride. The PISCES-B results also show a material loss mechanism that lowers the maximum operating temperature compared to that for the liquid surface equilibrium vapor pressure. This may restrict the lithium temperature in reactors.

The National Spherical Torus Experiment (NSTX) is a low aspect ratio (R/a approximately 1.3) device with auxiliary heating from neutral-beam injection (NBI) and high-harmonic fast-wave heating (HHFW). Typical NSTX parameters are R(subscript ''0'') = 85 cm, a = 67 cm, I(subscript ''p'') = 0.7-1.4 MA, B(subscript ''phi'') = 0.25-0.45 T. Three co-directed deuterium neutral-beam sources have injected P(subscript ''NB'') less than or equal to 4.7 MW. HHFW plasmas typically have delivered P(subscript ''RF'') less than or equal to 3 MW. Important to the understanding of NSTX confinement are the new kinetic profile diagnostics: a multi-pulse Thomson scattering system (MPTS) and a charge-exchange recombination spectroscopy (CHERS) system. The MPTS diagnostic currently measures electron density and temperature profiles at 30 Hz at ten spatial locations. The CHERS system has recently become available to measure carbon ion temperature and toroidal flow at 17 radial positions spanning the outer half of the minor radius with 20 msec time resolution during NBI. Experiments conducted during the last year have produced a wide range of kinetic profiles in NSTX. Some interesting examples are presented below.

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.

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.

Experiments have been performed in the National Spherical Torus Experiment (NSTX) to inject high harmonic fast wave (HHFW) power early during the plasma current ramp-up in an attempt to reduce the current penetration rate to raise the central safety factor during the flattop phase of the discharge. To date, up to 2 MW of HHFW power has been coupled to deuterium plasmas as early as t = 50 ms using the slowest interstrap phasing of k|| approximately equals 14 m(superscript)-1 (nf = 24). Antenna-plasma gap scans have been performed and find that for small gaps (5-8 cm), electron heating is observed with relatively small density rises and modest reductions in current penetration rate. For somewhat larger gaps (10-12 cm), weak electron heating is observed but with a spontaneous density rise at the plasma edge similar to that observed in NSTX H-modes. In the larger gap configuration, EFIT code reconstructions (without MSE [motional Stark effect]) find that resistive flux consumption is reduced as much as 30%, the internal inductance is maintained below 0.6 at 1 MA into the flattop, q(0) is increased significantly, and the MHD stability character of the discharges is strongly modified.

The spherical tokamak (ST), because of its slender central column, has very limited volt-second capability relative to a standard aspect ratio tokamak of similar plasma cross-section. Recent experiments on the National Spherical Torus Experiment (NSTX) have begun to quantify and optimize the ohmic current drive efficiency in a MA-class ST device. Sustainable ramp-rates in excess of 5MA/sec during the current rise phase have been achieved on NSTX, while faster ramps generate significant MHD activity. Discharges with Ip exceeding 1MA have been achieved in NSTX with nominal parameters: aspect ratio A=1.3--1.4, elongation k=2--2.2, triangularity d=0.4, internal inductance li=0.6, and Ejima coefficient CE=0.35. Flux consumption efficiency results, performance improvements associated with first boronization, and comparisons to neoclassical resistivity are described.

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.

Resistive MHD modes and associated effects on spherical torus performance are investigated in the CDX-U device for Ip {le} 100 kA. Presently, the growth of resistive MHD modes (n=1/m=3 or n=1/m=2) as the edge q[q(a)] is lowered toward 3.5 appears to limit the maximum current achievable in CDX-U. For low q(a) discharges, a prominent rotating hot spot can be seen with the soft x-ray array, indicative of a magnetic island associated with a n=1/m=1 mode. The edge mode, which is n=1/m=3 or n=1/m=2, can be seen by the soft x-ray and edge magnetic pick up coil array. The growth of those modes in space and amplitude eventually leads to an Internal Reconnection Event (IRE). Prior to the IRE, strong mode-mixing takes place suggesting magnetic island overlap. The IRE causes a rapid heat loss from the core causing a strong plasma elongation and current spike due to the plasma inductance drop. With an appropriate discharge control, a MHD quiescent high confinement regime with over twice the central electron temperature relative to the MHD active regime has been found. To assess the halo-induced effects during the MHD events, a pair of segmented Rogowski coils were installed on the center stack. The observed halo-induced current fraction is generally small (less than 5% of the total plasma current) even for the case of forced disruption.

The soft X‐ray emission of intermediate and high Z elements from a low inductance vacuum spark, has been analyzed by space‐resolved spectroscopy and modeling of the emission. In the case of intermediate Z elements, the observed spatial extent and the estimated plasma parameters of the regions emitting L‐shell ions transitions are consistent with the predictions of the radiative collapse model of the spark. For high Z elements, the estimated characteristics of the plasma emitting radiation of the M‐shell ions seem to indicate that in this case too a radiative collapse may take place, triggered by the intense line emission of Ni‐like ions.

Electron transport is rapid in most National Spherical Torus Experiment, M. Ono et al., Nucl. Fusion 40, 557 (2000) beam heated plasmas. A regime of improved electron confinement is nevertheless observed in low density L-mode (“low-confinement”) discharges heated by early beam injection. Experiments were performed in this regime to study the role of the current profile on thermal transport. Variations in the magnetic shear profile were produced by changing the current ramp rate and onset of neutral beam heating. An increased electron temperature gradient and local minimum in the electron thermal diffusivity were observed at early times in plasmas with the fastest current ramp and earliest beam injection. In addition, an increased ion temperature gradient associated with a region of reduced ion transport is observed at slightly larger radii. Ultrasoft x-ray measurements of double-tearing magnetohydrodynamic activity, together with current diffusion calculations, point to the existence of negative magnetic ...

A pinhole soft x-ray imaging camera is being developed for use on the Caltech solar prominence simulation experiment and also the Caltech spheromak experiment. The camera is based upon a commercial gated intensifier which produces an image on a phosphor screen. Moderate signal level, excellent time resolution, and reasonable imaging have been obtained, but there has not been any determination of the x-ray energy spectrum. An estimation of the spectrum is now underway using filtered AXUV diodes and it is expected that knowledge of the x-ray energy will enable further optimization of the camera.

A multipass/multipulse Thomson scattering system has been implemented on CDX-U in collaboration with the Ioffe Institute, St. Petersburg. The system consists of a low energy (1.5– 2.5 J), passively Q-switched ruby laser, and a multipass optical cavity enclosing the plasma. Multiple reflections of the beam within the cavity increase by about an order of magnitude the number of scattered photons, allowing temperature density to be measured with good accuracy even at very low plasma density. By feeding the returned beam back into the laser, the system can deliver several pulses over a 1 ms period. However, the experiments on CDX-U show that a mechanical shock wave reaching the multipass system affects the feedback and laser output per pulse drops significantly. Therefore we operate the system by fine tuning the laser cavity, so that output is practically independent of feedback from multipass cavity. Also, by optimizing the transmission of the passive Q-switch and the pumping power, we obtain that laser outp...