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Pulsed injections of boron carbide granules into Wendelstein 7-X stellarator (W7-X) plasmas transiently increase the plasma stored energy and core ion temperatures above the reference W7-X experimental programs by up to 30%. In a series of 4 MW electron cyclotron resonance heating experiments, the PPPL Probe Mounted Powder Injector provided 50 ms bursts of 100 μm granules every 350 ms at estimated quantities ranging from approximately 1 mg/pulse to over 30 mg/pulse. For each injection, the stored energy was observed to initially drop and the radiated power transiently increased, while the radial electron density profile rose at the edge as material was assimilated. Once the injected boron carbide was fully absorbed, the density rise transitioned to the core while the stored energy increased above the previous baseline level by an amount linearly correlated with the injection quantity. During the injection, the ion temperature gradient steepened with peak core ion temperatures observed to increase from a nominal 1.7 keV to over 2.6 keV for the largest injection amounts. Enhanced performance is accompanied by a reversal of the radial electric field at ρ < 0.3, indicating that the core transport has switched to the ion root. These observations are suggestive of a change in transport and provide further evidence that externally induced profile modifications provide a possible path to enhanced W7-X performance metrics. [ABSTRACT FROM AUTHOR]

A scalable system for real-time analysis of electron temperature and density based on signals from the Thomson scattering diagnostic, initially developed for and installed on the NSTX-U experiment, was recently adapted for the Large Helical Device and operated for the first time during plasma discharges. During its initial operation run, it routinely recorded and processed signals for four spatial points at the laser repetition rate of 30 Hz, well within the system's rated capability for 60 Hz. We present examples of data collected from this initial run and describe subsequent adaptations to the analysis code to improve the fidelity of the temperature calculations. [ABSTRACT FROM AUTHOR]

A model-based feedback system is presented enabling the simultaneous control of the stored energy through βn and the toroidal rotation profile of the plasma in National Spherical Torus eXperiment Upgrade device. Actuation is obtained using the momentum from six injected neutral beams and the neoclassical toroidal viscosity generated by applying three-dimensional magnetic fields. Based on a model of the momentum diffusion and torque balance, a feedback controller is designed and tested in closed-loop simulations using TRANSP, a time dependent transport analysis code, in predictive mode. Promising results for the ongoing experimental implementation of controllers are obtained. [ABSTRACT FROM AUTHOR]

Study Question: Does hormonal stimulation with corifollitropin alpha (CFA) only, mimicking a step down protocol, result in lower incidence of progesterone elevation on the day of hCGtrigger as compared to sustained stimulation with recombinant FSH (rFSH)?Summary Answer: The current findings support the concept that sustained FSH stimulus contributes to premature progesterone elevation in stimulated IVF cycles.What Is Known Already: Serum progesterone rise during the follicular phase of ovarian stimulation for IVF treatment seems to be related to a poorer reproductive outcome. However, the mechanism by which the rise in progesterone is caused is not yet fully understood.Study Design, Size, Duration: This study was a post hoc analysis of data from two multi-center, randomized, double-blind, double-dummy, active-controlled, non-inferiority trials, ENGAGE and PURSUE, conducted from June 2006 to January 2008 and from July 2010 to October 2012 respectively.Participants/materials, Setting, Methods: In the ENGAGE-study, 1506 women, aged 18-36 years, were allocated to either a single injection of 150 mg CFA or daily injections of 200 IU rFSH in the first week of stimulation, using a standard GnRH antagonist protocol. In the PURSUE-study, a total of 1390 women, aged 35-42 years, were allocated to either a single injection of 150 mg of CFA or daily 300 IU of rFSH for the first week, again using a standard GnRH antagonist protocol. In both trials, daily rFSH was continued until three follicles reached >17 mm in size. All women had a body weight of between 50 and 90 kg, regular menstrual cycles and an indication for ovarian stimulation before IVF. The incidence of progesterone elevation on day of hCG-trigger in patients with CFA only or rFSH stimulation, and triggered on Day 8 of stimulation, was analyzed.Main Results and the Role Of Chance: Of patients with CFA only stimulation, 5.4% (13/239 patients) showed a progesterone elevation above 1.5 ng/ml on day of hCG-trigger, whereas patients with rFSH stimulation had a significant higher incidence of progesterone elevation (18.3%; 62/339 patients) (P < 0.001).Limitations, Reasons For Caution: Post hoc analysis of data from previously published trials could be considered as a reason for caution.Wider Implications Of the Findings: Future studies should evaluate whether it would be possible to prevent a premature progesterone rise in cycles stimulated with daily FSH by using a step down protocol towards the end of the follicular phase.Study Funding/competing Interests: Financial/Material Support was provided by Merck & Co., Inc., Kenilworth, NJ, USA. Davis Gates is an employee of Merck Sharp & Dohme Corp., a subsidiary of Merck & Co., Inc., Kenilworth, NJ, USA and may own stock and/or hold stock options in the company. Fabiola Beligotti is an employee of MSD, Italy, a subsidiary of Merck & Co., Inc., Kenilworth, NJ, USA and may own stock and/or hold stock options in the company. Barbara Lawrenz, Nils Engelmann and Human M. Fatemi have no conflict of interest.Trial Registration Number: ENGAGE study: ClinicalTrials.gov identifier NTC00696800. PURSUE-study: NCT01144416. [ABSTRACT FROM AUTHOR]

The growth of a tearing mode is described by reduced MHD equations. For a cylindrical equilibrium, tearing mode growth is governed by the modified Rutherford equation, i.e., the nonlinear Δ'(w). For a low beta plasma without external heating, Δ'(w) can be approximately described by two terms, Δ'ql(w), Δ′A(w) [White et al., Phys. Fluids 20, 800 (1977); Phys. Plasmas 22, 022514 (2015)]. In this work, we present a simple method to calculate the quasilinear stability index Δ′ql rigorously, for poloidal mode number m≥2. Δ′ql is derived by solving the outer equation through the Frobenius method. Δ'ql is composed of four terms proportional to: constant Δ'0, w, wlnw, and w 2. Δ′A is proportional to the asymmetry of island that is roughly proportional to w. The sum of Δ′ql and Δ′A is consistent with the more accurate expression calculated perturbatively [Arcis et al., Phys. Plasmas 13, 052305 (2006)]. The reduced MHD equations are also solved numerically through a 3D MHD code M3D-C1 [Jardin et al., Comput. Sci. Discovery 5, 014002 (2012)]. The analytical expression of the perturbed helical flux and the saturated island width agree with the simulation results. It is also confirmed by the simulation that the Δ′A has to be considered in calculating island saturation. [ABSTRACT FROM AUTHOR]

The physical mechanism behind the tokamak density limit scaling is described in terms of a nonlinear theory of tearing mode growth in cylindrical geometry coupled to a model for thermal transport in the island. Important new physics features of the model include: (1) island asymmetry due to finite island width in cylindrical geometry, (2) a model of radiation based on local coronal equilibrium including impurity radiation, (3) current perturbations due the perturbed resistivity, and (4) numerical solution of the cylindrical eigenfunctions and Δ'. The semi-analytic cylindrical model is then solved for a wide range of current profiles, magnetic field values, and plasma currents using reasonable assumptions for impurity densities and the Greenwald limit [M. Greenwald et al., Nucl. Fusion 28, 2199 (1988)] is reproduced. The limit is shown to be only weakly dependent on variations in the assumed parameters. [ABSTRACT FROM AUTHOR]

The behavior of magnetic islands with 3D electron temperature and the corresponding 3D resistivity effects on growth are examined for islands with near-zero net heating in the island interior. We refer to the resulting class of non-linearities as thermo-resistive effects. In particular, the effects of varying impurity mix on the previously proposed local island onset threshold [Gates and Delgado-Aparicio, Phys. Rev. Lett. 108, 165004 (2012)] are examined and shown to be consistent with the well established experimental scalings for tokamaks at the density limit. A surprisingly simple semi-analytic theory is developed which imposes the effects of heating/cooling in the island interior as well as the effects of island geometry. For the class of current profiles considered, it is found that a new term that accounts for the thermal effects of island asymmetry is required in the modified Rutherford equation. The resultant model is shown to exhibit a robust onset of a rapidly growing tearing mode--consistent with the disruption mechanism observed at the density limit in tokamaks. A fully non-linear 3D cylindrical calculation is performed that simulates the effect of net island heating/cooling by raising/suppressing the temperature in the core of the island. In both the analytic theory and the numerical simulation, the sudden threshold for rapid growth is found to be due to an interaction between three distinct thermal non-linearities which affect the island resistivity, thereby modifying the growth dynamics. [ABSTRACT FROM AUTHOR]

New experimental observations are reported on the structure and dynamics of short-lived periodic (1, 1) "fishbone"-like oscillations that appear during radio frequency heating and current-drive experiments in tokamak plasmas. For the first time, measurements can directly relate changes in the high energy electrons to the mode onset, saturation, and damping. In the relatively high collisionality of Alcator C-Mod with lower hybrid current drive, the instability appears to be destabilized by the non-resonant suprathermal electron pressure--rather than by wave-particle resonance, rotates toroidally with the plasma and grows independently of the (1, 1) sawtooth crash driven by the thermal plasma pressure. [ABSTRACT FROM AUTHOR]

We propose the recently defined "Holoraumy Tensor" to play a critical role in defining a metric to establish a correspondence between 4D, -extended 0-brane-based valise supermultiplet representations and, correspondingly via "SUSY Holography," on the space of 1D, N-extended network-based adinkras. Using an analogy with the su(3) algebra, it is argued the 0-brane holoraumy tensors play the role of the "-coefficients" and provide a newly established tool for investigating supersymmetric representation theory. [ABSTRACT FROM AUTHOR]

Magnetic reconnection is ubiquitous in the magnetosphere, the solar corona, and in toroidal fusion research discharges. In a fusion device, a magnetic island saturates at a width which produces a minimum in the magnetic energy of the configuration. At saturation, the modified current density profile, a function of the flux in the island, is essentially flat, the growth rate proportional to the difference in the current at the O-point and the X-point. Further modification of the current density profile in the island interior causes a change in the island stability and additional growth or contraction of the saturated island. Because field lines in an island are isolated from the outside plasma, an island can heat or cool preferentially depending on the balance of Ohmic heating and radiation loss in the interior, changing the resistivity and hence the current in the island. A simple model of island destabilization due to radiation cooling of the island is constructed, and the effect of modification of the current within an island is calculated. An additional destabilization effect is described, and it is shown that a small imbalance of heating can lead to exponential growth of the island. A destabilized magnetic island near the plasma edge can lead to plasma loss, and because the radiation is proportional to plasma density and charge, this effect can cause an impurity dependent density limit. [ABSTRACT FROM AUTHOR]

A comparative study of energy transport for medium- to high-density discharges in the stellarator-heliotrons TJ-II, W7-AS and LHD is carried out. The specific discharge parameters are chosen to apply a recently concluded benchmarking study of neoclassical (NC) transport coefficients (Beidler et al 2011 Nucl. Fusion 51 076001) to perform this validation study. In contrast to previous experiments at low densities for which electron transport was predominant (Yokoyama et al 2007 Nucl. Fusion 47 1213), the current discharges also exhibit significant ion energy transport. As it affects the energy transport in 3D devices, the ambipolar radial electric field is addressed as well. For the discharges described, ion-root conditions, i.e. a small negative radial electric field were found. The energy transport in the peripheral region cannot be explained by NC theory. Within a 'core region' (r/a < 1/2 ~ 2/3), the predicted NC energy fluxes comply with experimental findings for W7-AS. For TJ-II, compliance in the core region is found for the particle transport and the electron energy transport. For the specific LHD discharges, the core energy transport complied with NC theory except for the electron energy transport in the inward-shifted magnetic configuration. The NC radial electric field tends to agree with experimental results for all devices but is measured to be more negative in the core of both LHD and TJ-II. As a general observation, the energy confinement time approaches the gyro-Bohm-type confinement scaling ISS04 (Yamada et al 2005 Nucl. Fusion 45 1684). This work is carried out within the International Stellarator-Heliotron Profile Database (www.ipp.mpg.de/ISS and http://ishpdb.nifs.ac.jp/index.html). [ABSTRACT FROM AUTHOR]

In previous work (Gates and Delgado-Aparicio 2012 Phys. Rev. Lett. 108 165004), the onset criterion for radiation-driven islands (Rebut etal 1985 Proc. 10th Int. Conf. on Plasma Physics and Controlled Nuclear Fusion Research 1984 (London, UK, 1984) vol 2 (Vienna: IAEA) p 197) in combination with a simple cylindrical model of tokamak current channel behaviour was shown to be consistent with the empirical scaling of the tokamak density limit (Greenwald et al 1988 Nucl. Fusion 28 2199). A number of the unexplained phenomena at the density limit are consistent with this novel physics mechanism. In this work, a more formal theoretical underpinning, consistent with cylindrical tearing mode theory, is developed for the onset criteria of these modes. The appropriate derivation of the radiation-driven addition to the modified Rutherford equation (MRE) is discussed. Additionally, the ordering of the terms in the MRE is examined in a regime near the density limit. It is hoped that, given the apparent success of this simple model in explaining the observed global scalings, it will lead to a more comprehensive analysis of the possibility that radiation-driven islands are the physics mechanism responsible for the density limit. In particular, with modern diagnostic capabilities detailed measurements of current densities, electron densities and impurity concentrations at rational surfaces should be possible, enabling verification of the concepts described above. [ABSTRACT FROM AUTHOR]

First results of ion and electron temperature profile measurements from the x-ray imaging crystal spectrometer (XICS) diagnostic on the Large Helical Device (LHD) are presented. This diagnostic system has been operational since the beginning of the 2011 LHD experimental campaign and is the first application of the XICS diagnostic technique to helical plasma geometry. The XICS diagnostic provides measurements of ion and electron temperature profiles in LHD with a spatial resolution of 2 cm and a maximum time resolution of 5 ms (typically 20 ms). Ion temperature profiles from the XICS diagnostic are possible under conditions where charge exchange recombination spectroscopy (CXRS) is not possible (high density) or is perturbative to the plasma (low density or radio frequency heated plasmas). Measurements are made by using a spherically bent crystal to provide a spectrally resolved 1D image of the plasma from line integrated emission of helium-like Ar16 +. The final hardware design and configuration are detailed along with the calibration procedures. Line-integrated ion and electron temperature measurements are presented, and the measurement accuracy is discussed. Finally central temperature measurements from the XICS system are compared to measurements from the Thomson scattering and CXRS systems, showing excellent agreement. [ABSTRACT FROM AUTHOR]

The spherical tokamak (ST) is a leading candidate for a Fusion Nuclear Science Facility (FNSF) due to its compact size and modular configuration. The National Spherical Torus eXperiment (NSTX) is a MA-class ST facility in the US actively developing the physics basis for an ST-based FNSF. In plasma transport research, ST experiments exhibit a strong (nearly inverse) scaling of normalized confinement with collisionality, and if this trend holds at low collisionality, high fusion neutron fluences could be achievable in very compact ST devices. A major motivation for the NSTX Upgrade (NSTX-U) is to span the next factor of 3-6 reduction in collisionality. To achieve this collisionality reduction with equilibrated profiles, NSTX-U will double the toroidal field, plasma current, and NBI heating power and increase the pulse length from 1-1.5 s to 5-8 s. In the area of stability and advanced scenarios, plasmas with higher aspect ratio and elongation, high βN, and broad current profiles approaching those of an ST-based FNSF have been produced in NSTX using active control of the plasma β and advanced resistive wall mode control. High non-inductive current fractions of 70% have been sustained for many current diffusion times, and the more tangential injection of the 2nd NBI of the Upgrade is projected to increase the NBI current drive by up to a factor of 2 and support 100% non-inductive operation. More tangential NBI injection is also projected to provide non-solenoidal current ramp-up as needed for an ST-based FNSF. In boundary physics, NSTX measures an inverse relationship between the scrape-off layer heat-flux width and plasma current that could unfavourably impact next-step devices. Recently, NSTX has successfully demonstrated substantial heat-flux reduction using a snowflake divertor configuration, and this type of divertor is incorporated in the NSTX-U design. The physics and engineering design supporting NSTX Upgrade is described. [ABSTRACT FROM AUTHOR]

Experimental results from NSTX indicate that the snowflake divertor (D. Ryutov, Phys. Plasmas 14, 064502 (2007)) may be a viable solution for outstanding tokamak plasma-material interface issues. Steady-state handling of divertor heat flux and divertor plate erosion remains to be critical issues for ITER and future concept devices based on conventional and spherical tokamak geometry with high power density divertors. Experiments conducted in 4-6 MW NBI-heated H-mode plasmas in NSTX demonstrated that the snowflake divertor is compatible with high-confinement core plasma operation, while being very effective in steady-state divertor heat flux mitigation and impurity reduction. A steady-state snowflake divertor was obtained in recent NSTX experiments for up to 600 ms using three divertor magnetic coils. The high magnetic flux expansion region of the scrape-off layer (SOL) spanning up to 50% of the SOL width λq was partially detached in the snowflake divertor. In the detached zone, the heat flux profile flattened and decreased to 0.5-1 MW/m2 (from 4-7 MW/m2 in the standard divertor) indicative of radiative heating. An up to 50% increase in divertor, Prad in the snowflake divertor was accompanied by broadening of the intrinsic C III and C IV radiation zones, and a nearly order of magnitude increase in divertor high-n Balmer line emission indicative of volumetric recombination onset. Magnetic reconstructions showed that the x-point connection length, divertor plasma-wetted area and divertor volume, all critical parameters for geometric reduction of deposited heat flux, and increased volumetric divertor losses were significantly increased in the snowflake divertor, as expected from theory. [ABSTRACT FROM AUTHOR]

A high-resolution x-ray imaging crystal spectrometer, whose concept was tested on NSTX and Alcator C-Mod, is being designed for the large helical device (LHD). This instrument will record spatially resolved spectra of helium-like Ar16+ and will provide ion temperature profiles with spatial and temporal resolutions of <2 cm and ≥10 ms, respectively. The spectrometer layout and instrumental features are largely determined by the magnetic field structure of LHD. The stellarator equilibrium reconstruction codes, STELLOPT and PIES, will be used for the tomographic inversion of the spectral data. [ABSTRACT FROM AUTHOR]

The far infrared tangential interferometer/polarimeter (FIReTIP) of the National Spherical Torus Experiment (NSTX) has been set up to provide reliable electron density signals for a real-time density feedback control system. This work consists of two main parts: suppression of the fringe jumps that have been prohibiting the plasma density from use in the direct feedback to actuators and the conceptual design of a density feedback control system including the FIReTIP, control hardware, and software that takes advantage of the NSTX plasma control system (PCS). By investigating numerous shot data after July 2009 when the new electronics were installed, fringe jumps in the FIReTIP are well characterized, and consequently the suppressing algorithms are working properly as shown in comparisons with the Thomson scattering diagnostic. This approach is also applicable to signals taken at a 5 kHz sampling rate, which is a fundamental constraint imposed by the digitizers providing inputs to the PCS. The fringe jump correction algorithm, as well as safety and feedback modules, will be included as submodules either in the gas injection system category or a new category of density in the PCS. [ABSTRACT FROM AUTHOR]

Steady-state handling of divertor heat flux is a critical issue for both ITER and spherical torus-based devices with compact high power density divertors. Significant reduction of heat flux to the divertor plate has been achieved simultaneously with favorable core and pedestal confinement and stability properties in a highly shaped lower single null configuration in the National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 2000] using high magnetic flux expansion at the divertor strike point and the radiative divertor technique. A partial detachment of the outer strike point was achieved with divertor deuterium injection leading to peak flux reduction from 4–6 MW m-2 to 0.5–2 MW m-2 in small-ELM 0.8–1.0 MA, 4–6 MW neutral beam injection-heated H-mode discharges. A self-consistent picture of the outer strike point partial detachment was evident from divertor heat flux profiles and recombination, particle flux and neutral pressure measurements. Analytic scrape-off layer parallel transport models were used for interpretation of NSTX detachment experiments. The modeling showed that the observed peak heat flux reduction and detachment are possible with high radiated power and momentum loss fractions, achievable with divertor gas injection, and nearly impossible to achieve with main electron density, divertor neutral density or recombination increases alone. [ABSTRACT FROM AUTHOR]

National Spherical Torus Experiment [which M. Ono et al., Nucl. Fusion 40, 557 (2000)] high-power divertor plasma experiments have shown, for the first time, that benefits from lithium coatings applied to plasma facing components found previously in limited plasmas can occur also in high-power diverted configurations. Lithium coatings were applied with pellets injected into helium discharges, and also with an oven that directed a collimated stream of lithium vapor toward the graphite tiles of the lower center stack and divertor. Lithium oven depositions from a few milligrams to 1 g have been applied between discharges. Benefits from the lithium coatings were sometimes, but not always, seen. These benefits sometimes included decreases in plasma density, inductive flux consumption, and edge-localized mode occurrence, and increases in electron temperature, ion temperature, energy confinement, and periods of edge and magnetohydrodynamic quiescence. In addition, reductions in lower divertor D, C, and O luminosity were measured. [ABSTRACT FROM AUTHOR]

The purpose of this study was to examine incivility experienced by direct health care staff in their workplaces. The sample (N = 184) was 91% female and 77% White, with 71% of the participants having earned an associate degree or above and 81% being registered nurses. The Work Limitations Questionnaire and the Incivility in Healthcare Survey were distributed to all direct care staff at a major metropolitan hospital (22% response rate). Correlations were found between workplace incivility from direct supervisors and productivity (r = 0.284, p = .000) and workplace incivility from patients and productivity (r = 0.204, p = .006). Incivility from physicians, incivility from other direct care staff, and general environmental incivility were not shown to be significantly related to productivity. Demographics were not related to levels of workplace incivility. Workplace incivility from patients and management appears to have a greater impact on employees' productivity than workplace incivility from other sources. [ABSTRACT FROM AUTHOR]

In the National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 (2000)], plasmas with strongly reversed magnetic shear, s≡(r/q)(dq/dr)<0, in the plasma core exhibit a marked improvement in electron confinement compared to otherwise similar plasmas with positive or only weakly reversed magnetic shear. The q profile itself is determined by the early evolution of the plasma current, the plasma cross section, and the neutral-beam heating power. In the region of shear reversal, the electron thermal diffusivity can be significantly reduced. Detailed experimental investigation of this phenomenon has been made possible by the successful development of a motional Stark effect (MSE) polarimetry diagnostic suitable for the low magnetic field in NSTX, typically 0.35–0.55 T. Measurements of the electron and ion temperature, density, and plasma toroidal rotation profiles are also available with high spatial and temporal resolution for analysis of the plasma transport properties. [ABSTRACT FROM AUTHOR]

It is estimated that employers spend more than 75 billion dollars annually on obesity-attributable health care. Interventions to reduce or prevent the risk of obesity are increasingly common at worksites and include health fairs, weight loss and nutrition classes, and fitness programs. However, many companies lack the resources to plan and implement these types of programs. Environmental approaches offer companies a low-cost option. A community-based participatory research model was used to bring academic researchers, human resources personnel, and health department educators together to plan and implement an environmental program aimed at increasing healthy eating and physical activity at four small manufacturing companies. The Diffusion of Innovations Theory guided the development of focus group questions. A focus group study was then conducted to gather information from employees and managers at these four companies. The questions identified workplace strategies that would aid in reducing barriers and developing appropriate communication channels to enhance employee participation in the program. The researchers identified themes from manager and employee focus groups regarding the following five environmental components: signs, walking paths, food changes, educational strategies, and advisory groups. [ABSTRACT FROM AUTHOR]

There has been a substantial international research effort in the fusion community to identify tokamak operating regimes with either small or no periodic bursts of particles and power from the edge plasma, known as edge-localized modes (ELMs). While several candidate regimes have been presented in the literature, very little has been published on the characteristics of the small ELMs themselves. One such small ELM regime, also known as the Type V ELM regime, was recently identified in the National Spherical Torus Experiment [M. Ono, S. M. Kaye, Y.-K. M. Peng et al., Nucl. Fusion 40, 557 (2000)]. In this paper, the spatial and temporal structure of the Type V ELMs is presented, as measured by several different diagnostics. The composite picture of the Type V ELM is of an instability with one or two filaments that rotate toroidally at ∼5–10 km/s, in the direction opposite to the plasma current and neutral beam injection. The toroidal extent of Type V ELMs is typically ∼5 m, whereas the cross-field (radial) extent is typically ∼10 cm (3 cm), yielding a portrait of an electromagnetic, ribbon-like perturbation aligned with the total magnetic field. The filaments comprising the Type V ELM appear to be destabilized near the top of the H-mode pedestal and drift radially outward as they rotate toroidally. After the filaments come in contact with the open field lines, the divertor plasma perturbations are qualitatively similar to other ELM types, albeit with only one or two filaments in the Type V ELM versus more filaments for Type I or Type III ELMs. Preliminary stability calculations eliminate pressure driven modes as the underlying instability for Type V ELMs, but more work is required to determine if current driven modes are responsible for destabilization. [ABSTRACT FROM AUTHOR]

The National Spherical Torus Experiment [Ono et al., Nucl. Fusion, 44, 452 (2004)] is targeting long pulse high performance, noninductive sustained operations at low aspect ratio, and the demonstration of nonsolenoidal startup and current rampup. The modeling of these plasmas provides a framework for experimental planning and identifies the tools to access these regimes. Simulations based on neutral beam injection (NBI)-heated plasmas are made to understand the impact of various modifications and identify the requirements for (1) high elongation and triangularity, (2) density control to optimize the current drive, (3) plasma rotation and/or feedback stabilization to operate above the no-wall β limit, and (4) electron Bernstein waves (EBW) for off-axis heating/current drive (H/CD). Integrated scenarios are constructed to provide the transport evolution and H/CD source modeling, supported by rf and stability analyses. Important factors include the energy confinement, Zeff, early heating/H mode, broadening of the NBI-driven current profile, and maintaining q(0) and qmin>1.0. Simulations show that noninductive sustained plasmas can be reached at IP=800 kA, BT=0.5 T, κ≈2.5, βN<=5, β<=15%, fNI=92%, and q(0)>1.0 with NBI H/CD, density control, and similar global energy confinement to experiments. The noninductive sustained high β plasmas can be reached at IP=1.0 MA, BT=0.35 T, κ≈2.5, βN<=9, β<=43%, fNI=100%, and q(0)>1.5 with NBI H/CD and 3.0 MW of EBW H/CD, density control, and 25% higher global energy confinement than experiments. A scenario for nonsolenoidal plasma current rampup is developed using high harmonic fast wave H/CD in the early low IP and low Te phase, followed by NBI H/CD to continue the current ramp, reaching a maximum of 480 kA after 3.4 s. [ABSTRACT FROM AUTHOR]

The National Spherical Torus Experiment (NSTX) has explored the effects of shaping on plasma performance as determined by many diverse topics including the stability of global magnetohydrodynamic (MHD) modes (e.g., ideal external kinks and resistive wall modes), edge localized modes (ELMs), bootstrap current drive, divertor flux expansion, and heat transport. Improved shaping capability has been crucial to achieving βt∼40%. Precise plasma shape control has been achieved on NSTX using real-time equilibrium reconstruction. NSTX has simultaneously achieved elongation κ∼2.8 and triangularity δ∼0.8. Ideal MHD theory predicts increased stability at high values of shaping factor S≡q95Ip/(aBt), which has been observed at large values of the S∼37[MA/(m·T)] on NSTX. The behavior of ELMs is observed to depend on plasma shape. A description of the ELM regimes attained as shape is varied will be presented. Increased shaping is predicted to increase the bootstrap fraction at fixed Ip. The achievement of strong shaping has enabled operation with 1 s pulses with Ip=1 MA, and for 1.6 s for Ip=700 kA. Analysis of the noninductive current fraction as well as empirical analysis of the achievable plasma pulse length as elongation is varied will be presented. Data are presented showing a reduction in peak divertor heat load due to increasing in flux expansion. [ABSTRACT FROM AUTHOR]

The resistive wall mode (RWM) poses a limit to the maximum β that can be sustained in magnetic fusion experiments. RWM stabilization physics at low aspect ratio is studied in high-β National Spherical Torus Experiment (NSTX) [M. Ono, S. M. Kaye, Y.-K. M. Peng et al., Nucl. Fusion 40, 557 (2000)] plasmas (βt up to 39%; βN up to 6.8) to understand and alleviate this constraint. Plasmas with increased q in NSTX have been maintained with β above the computed ideal no-wall β limit for more than 20 wall times with no signs of RWM growth in cases where toroidal rotation ω[lowercase_phi_synonym]>ωA/4q2 across the entire plasma cross section. Plasmas that violate this stability criterion can suffer a RWM induced collapse within a few wall times. This critical rotation profile for stabilization is in agreement with drift-kinetic theory applied to low frequency magnetohydrodynamics modes [A. Bondeson and M. S. Chu, Phys. Plasmas 3, 3013 (1996)]. A toroidally symmetric array of internal sensors has been used to observe n=1–3 RWMs in NSTX. This array consists of Bp and Br sensors both above and below the midplane at 12 toroidal locations instrumented to detect toroidal mode numbers of n=1–3. RWM perturbations exceeding 30 G have been measured with mode growth rates on the order of 5 ms. Small modes (δB<10 G) which cause minor drops in β, with growth rates ∼1500 s-1 have been observed when βN exceeds 6. Resonant field amplification of an externally applied error field by the stable RWM has been observed. [ABSTRACT FROM AUTHOR]

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

Recent experiments in the low aspect ratio National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 (2000)] have achieved normalized beta values twice the conventional tokamak limit at low internal inductance and with significant bootstrap current. These experimental results have motivated a computational re-examination of the plasma aspect ratio dependence of ideal no-wall magnetohydrodynamic stability limits. These calculations find that the profileoptimized no-wall stability limit in high bootstrap fraction regimes is well described by a nearly aspect ratio invariant normalized beta parameter utilizing the total magnetic field energy density inside the plasma. However, the scaling of normalized beta with internal inductance is found to be strongly aspect ratio dependent at sufficiently low aspect ratio. These calculations and detailed stability analyses of experimental equilibria indicate that the nonrotating plasma no-wall stability limit has been exceeded by as much as 30% in NSTX in a high bootstrap fraction regime. © 2004 American Institute of Physics. [ABSTRACT FROM AUTHOR]

Long pulse, high β scenarios have been established on the National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 (2000)]. β[SUBt](≡ 2μ[SUB0]〈P〉/B[SUP2,SUBt0] &sim 35% has been achieved during transient discharges. The machine improvements that lead to these results, including error field reduction and high temperature bakeout of plasma facing components are described. The highest β[SUBt] plasmas have high triangularity (δ = 0.8) and elongation (κ = 2.0), at low aspect ratio A ≡ R/a = 1.4. The strong shaping permits large values of normalized current, I[SUBN] (≡ I[SUBp]/(aB[SUBt0])≊6 while maintaining moderate values of q[SUB95] = 4. Long pulse discharges up to 1 s in duration have been achieved with substantial bootstrap current. The total noninductive current drive can be as high as 60%, comprised of 50% bootstrap current and ∼10% neutral beam current drive. The confinement enhancement factor H[SUB89P] is in excess of 2.7. β[SUBN]*H[SUB89P] &gsim;15 has been maintained for 8*τ[SUBE] ∼ 1.6*τ[SUBCR] , where τ[SUBCR] is the relaxation time of the first radial moment of the toroidal current density. The ion temperature for these plasmas is significantly higher than that predicted by neoclassical theory. [ABSTRACT FROM AUTHOR]

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

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 Editors, Nucl. Fusion 39, 2175 (1999)] scaling. Correlations of transition and ELM phenomena with turbulent fluctuations revealed by gas puff imaging and reflectometry are observed. In both single-null and double-null divertor discharges, the density peaks off-axis, sometimes developing prominent "ears" which can be sustained for many energy confinement times, τ[SUBE], in the absence of ELMs. A wide variety of ELM behavior is observed, and ELM characteristics depend on configuration and fueling. [ABSTRACT FROM AUTHOR]