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The L-H transition power threshold ( P LH ) in favourable magnetic geometry (ion ∇ B drift pointing towards X-point) is much lower than in the unfavourable magnetic geometry (ion ∇ B drift pointing away from X-point) on multiple tokamaks. In a systematic experiment on DIII-D, the ion ∇ B drift direction was changed continuously from the unfavourable to favourable configuration during plasma discharges. During such process, the input neutral beam power was kept constant at a value that was above P LH for favourable configuration, but lower than P LH for unfavourable configuration. Toroidal field and plasma current were also kept constant and there was little change in the edge electron density n e and electron temperature T e profiles. The density fluctuation amplitude was reduced approaching the transition, while a large increase of turbulence Reynolds stress and flow shear were simultaneously observed. The turbulence decorrelation rate was found to increase as the ion ∇ B drift direction was moving towards the favourable configuration, but the flow shear also increased and exceeded the turbulence decorrelation rate. These measurements demonstrate an important correlation between turbulence and turbulence-driven flow and a lowering of P LH , provide insights into the underlyingphysics behind the hidden parameters and inform a more complete physics-based model of the L-H transition power threshold. This article is part of a discussion meeting issue 'H-mode transition and pedestal studies in fusion plasmas'.

A new high radial resolution 2D multichannel Charge eXchange Imaging (CXI) diagnostic is under development for deployment at DIII-D. The diagnostic system will measure low-to-intermediate radial wavenumber carbon density fluctuations by observing the n = 8 − 7 ( λ = 529.06 nm) C-VI emission line, resulting from charge exchange collisions between heating neutral beam atoms and the intrinsic carbon ion density. The new CXI diagnostic will provide measurements with Δ R ∼ 0.4 cm to access higher k r instabilities ( k r < 8 cm−1) predicted to arise in the steep-gradient region of the H-mode pedestal. The CXI system will feature 60 fiber bundles in a 12 × 5 arrangement, with each bundle consisting of four 1 mm fibers. A custom optical system has been designed to filter and image incoming signals onto an 8 × 8 avalanche photodiode array. Additionally, a novel electronics suite has been designed and commissioned to amplify and digitize the relatively low-intensity carbon signal at a 2 MHz bandwidth. Forward modeling results of the active C-VI emission suggest sufficient signal to noise ratios to resolve turbulent fluctuations. Prototype measurements demonstrate the ability to perform high frequency pedestal measurements.

We elaborate on the nature of the prompt core confinement improvement observed at the L–H transition in DIII-D, which is a long-standing issue unsolved for more than two decades and can impact future fusion reactor performance. Dynamic transport analysis suggests the essential role of the profile stiffness for understanding the mechanism of the prompt core confinement improvement. Beam emission spectroscopy shows that transport reduction at the core cannot be explained only by the ion scale turbulence density fluctuation suppression. Properties of nonlocal confinement improvement across the L–H transition are experimentally assessed in hydrogen (H) and deuterium (D) plasmas. Prompt core confinement improvement is found to be more rapid in the lighter hydrogen isotope.

A novel spectroscopy diagnostic for measuring internal magnetic fields in high temperature magnetized plasmas has been developed. It involves spectrally resolving the Balmer- α (656 nm) neutral beam radiation split by the motional Stark effect with a spatial heterodyne spectrometer (SHS). The unique combination of high optical throughput (3.7 mm2sr) and spectral resolution ( δλ ∼ 0.1 nm) allows these measurements to be made with time resolution ≪1 ms. The high throughput is effectively utilized by incorporating a novel geometric Doppler broadening compensation technique in the spectrometer. The technique significantly reduces the spectral resolution penalty inherent to using large area, high-throughput optics while still collecting the large photon flux provided by such optics. In this work, fluxes of order 1010 s−1 support the measurement of deviations of [Formula: see text] mT (Δ λ Stark ∼ 10−4 nm) in the local magnetic field with 50 µs time resolution. Example high time resolution measurements of the pedestal magnetic field throughout the ELM cycle of a DIII-D tokamak plasma are presented. Local magnetic field measurements give access to the dynamics of the edge current density, which is essential to understanding stability limits, edge localized mode generation and suppression, and predicting performance of H-mode tokamaks.

Global gyrokinetic simulations of mesoscale reversed shear Alfven eigenmodes (RSAE) excited by energetic particles (EP) in fusion plasmas find that RSAE amplitude and EP transport are much higher than experimental levels at nonlinear saturation, but quickly diminish to very low levels after the saturation if background microturbulence is artificially suppressed. In contrast, in simulations coupling micro-meso scales, the RSAE amplitude and EP transport decrease drastically at the initial saturation but later increases to the experimental levels in the quasisteady state with bursty dynamics due to regulation by thermal ion temperature gradient (ITG) microturbulence. The quasisteady state EP transport is larger for a stronger microturbulence. The RSAE amplitude in the quasisteady state ITG-RSAE turbulence from gyrokinetic simulations, for the first time, agrees very well with experimental measurements.

New high speed localized measurements of the pedestal magnetic field during the edge localized mode (ELM) cycle of a DIII-D High confinement mode (H-mode) discharge indicate a temporally and spatial complex redistribution of the edge current density profile, jedge. The measurement technique extracts the magnetic field magnitude, B, via the spectral separation of Stark-split neutral beam radiation in the pedestal. Single spatial channel measurements from a novel spatial heterodyne spectrometer are validated in discharges with core current profile changes. The technique measures Stark-splitting changes that imply B changes as small as 1 mT with high time resolution (50 μs). At normalized poloidal flux [Formula: see text], B appears saturated in the inter-ELM period and then rapidly decreases in edge, B increases at the ELM crash. The behavior is consistent with a rapid collapse of jedge at the ELM crash and subsequent pedestal recovery. In some discharges, at [Formula: see text], changes in B are observed throughout the ELM cycle. In others, B recovers and is relatively stable until a few ms leading up to the next crash. Measurements of B during the H-mode transition show a large increase at [Formula: see text] with little change at [Formula: see text], consistent with the formation of the edge bootstrap current density peak. The [Formula: see text] spectrum is complicated by predicted changes to the Stark component intensities with density at the L–H transition.

DIII-D research is addressing critical challenges in preparation for ITER and the next generation of fusion devices through focusing on plasma physics fundamentals that underpin key fusion goals, understanding the interaction of disparate core and boundary plasma physics, and developing integrated scenarios for achieving high performance fusion regimes. Fundamental investigations into fusion energy science find that anomalous dissipation of runaway electrons (RE) that arise following a disruption is likely due to interactions with RE-driven kinetic instabilities, some of which have been directly observed, opening a new avenue for RE energy dissipation using naturally excited waves. Dimensionless parameter scaling of intrinsic rotation and gyrokinetic simulations give a predicted ITER rotation profile with significant turbulence stabilization. Coherence imaging spectroscopy confirms near sonic flow throughout the divertor towards the target, which may account for the convection-dominated parallel heat flux. Core-boundary integration studies show that the small angle slot divertor achieves detachment at lower density and extends plasma cooling across the divertor target plate, which is essential for controlling heat flux and erosion. The Super H-mode regime has been extended to high plasma current (2.0 MA) and density to achieve very high pedestal pressures (~30 kPa) and stored energy (3.2 MJ) with H 98y2 ≈ 1.6–2.4. In scenario work, the ITER baseline Q = 10 scenario with zero injected torque is found to have a fusion gain metric independent of current between q 95 = 2.8–3.7, and a lower limit of pedestal rotation for RMP ELM suppression has been found. In the wide pedestal QH-mode regime that exhibits improved performance and no ELMs, the start-up counter torque has been eliminated so that the entire discharge uses ≈0 injected torque and the operating space is more ITER-relevant. Finally, the high- (⩽3.8) hybrid scenario has been extended to the high-density levels necessary for radiating divertor operation, achieving ~40% divertor heat flux reduction using either argon or neon with P tot up to 15 MW.

Pulsed power accelerators compress electrical energy in space and time to provide versatile experimental platforms for high energy density and inertial confinement fusion science. The 80-TW “Z” pulsed power facility at Sandia National Laboratories is the largest pulsed power device in the world today. Z discharges up to 22 MJ of energy stored in its capacitor banks into a current pulse that rises in 100 ns and peaks at a current as high as 30 MA in low-inductance cylindrical targets. Considerable progress has been made over the past 15 years in the use of pulsed power as a precision scientific tool. This paper reviews developments at Sandia in inertial confinement fusion, dynamic materials science, x-ray radiation science, and pulsed power engineering, with an emphasis on progress since a previous review of research on Z in Physics of Plasmas in 2005.

The Z machine is a 36-module, multi-megavolt, low impedance magnetic pressure driver for high-energy-density physics experiments. In 2007, a major re-build doubled the stored energy and increased the peak current capability of Z. The upgraded system routinely drives 27 MA through low inductance dynamic loads with 110 nanosecond time to peak current. The Z pulsed power system is expected to be prepared for a full-energy experiment every day, with a small (

We have conceived a new class of prime-power sources for pulsed-power accelerators: impedance-matched Marx generators (IMGs). The fundamental building block of an IMG is a brick, which consists of two capacitors connected electrically in series with a single switch. An IMG comprises a single stage or several stages distributed axially and connected in series. Each stage is powered by a single brick or several bricks distributed azimuthally within the stage and connected in parallel. The stages of a multistage IMG drive an impedance-matched coaxial transmission line with a conical center conductor. When the stages are triggered sequentially to launch a coherent traveling wave along the coaxial line, the IMG achieves electromagnetic-power amplification by triggered emission of radiation. Hence a multistage IMG is a pulsed-power analogue of a laser. To illustrate the IMG approach to prime power, we have developed conceptual designs of two ten-stage IMGs with $LC$ time constants on the order of 100 ns. One design includes 20 bricks per stage, and delivers a peak electrical power of 1.05 TW to a matched-impedance $1.22\text{\ensuremath{-}}\ensuremath{\Omega}$ load. The design generates 113 kV per stage and has a maximum energy efficiency of 89%. The other design includes a single brick per stage, delivers 68 GW to a matched-impedance $19\text{\ensuremath{-}}\ensuremath{\Omega}$ load, generates 113 kV per stage, and has a maximum energy efficiency of 90%. For a given electrical-power-output time history, an IMG is less expensive and slightly more efficient than a linear transformer driver, since an IMG does not use ferromagnetic cores.

High current pulsed-power generators efficiently store and deliver magnetic energy to z-pinch targets. We review applications of magnetically driven implosions (MDIs) to inertial confinement fusion. Previous research on MDIs of wire-array z-pinches for radiation-driven indirect-drive target designs is summarized. Indirect-drive designs are compared with new targets that are imploded by direct application of magnetic pressure produced by the pulsed-power current pulse. We describe target design elements such as larger absorbed energy, magnetized and pre-heated fuel, and cryogenic fuel layers that may relax fusion requirements. These elements are embodied in the magnetized liner inertial fusion (MagLIF) concept [Slutz “Pulsed-power-driven cylindrical liner implosions of laser pre-heated fuel magnetized with an axial field,” Phys. Plasmas, 17, 056303 (2010), and Stephen A. Slutz and Roger A. Vesey, “High-Gain Magnetized Inertial Fusion,” Phys. Rev. Lett., 108, 025003 (2012)]. MagLIF is in the class of magneto-inertial fusion targets. In MagLIF, the large drive currents produce an azimuthal magnetic field that compresses cylindrical liners containing pre-heated and axially pre-magnetized fusion fuel. Scientific breakeven may be achievable on the Z facility with this concept. Simulations of MagLIF with deuterium-tritium fuel indicate that the fusion energy yield can exceed the energy invested in heating the fuel at a peak drive current of about 27 MA. Scientific breakeven does not require alpha particle self-heating and is therefore not equivalent to ignition. Capabilities to perform these experiments will be developed on Z starting in 2013. These simulations and predictions must be validated against a series of experiments over the next five years. Near-term experiments are planned at drive currents of 16 MA with D2 fuel. MagLIF increases the efficiency of coupling energy (=target absorbed energy/driver stored energy) to targets by 10-150X relative to indirect-drive targets. MagLIF also increases the absolute energy absorbed by the target by 10-50X relative to indirect-drive targets. These increases could lead to higher fusion gains and yields. Single-shot high yields are of great utility to national security missions. Higher efficiency and higher gains may also translate into more compelling (lower cost and complexity) fusion reactor designs. We will discuss the broad goals of the emerging research on the MagLIF concept and identify some of the challenges. We will also summarize advances in pulsed-power technology and pulsed-power driver architectures that double the efficiency of the driver.

We have developed a conceptual design of a next-generation pulsed-power accelerator that is optimized for megajoule-class dynamic-material-physics experiments. Sufficient electrical energy is delivered by the accelerator to a physics load to achieve---within centimeter-scale samples---material pressures as high as 1 TPa. The accelerator design is based on an architecture that is founded on three concepts: single-stage electrical-pulse compression, impedance matching, and transit-time-isolated drive circuits. The prime power source of the accelerator consists of 600 independent impedance-matched Marx generators. Each Marx comprises eight 5.8-GW bricks connected electrically in series, and generates a 100-ns 46-GW electrical-power pulse. A 450-ns-long water-insulated coaxial-transmission-line impedance transformer transports the power generated by each Marx to a system of twelve 2.5-m-radius water-insulated conical transmission lines. The conical lines are connected electrically in parallel at a 66-cm radius by a water-insulated 45-post sextuple-post-hole convolute. The convolute sums the electrical currents at the outputs of the conical lines, and delivers the combined current to a single solid-dielectric-insulated radial transmission line. The radial line in turn transmits the combined current to the load. Since much of the accelerator is water insulated, we refer to it as Neptune. Neptune is 40 m in diameter, stores 4.8 MJ of electrical energy in its Marx capacitors, and generates 28 TW of peak electrical power. Since the Marxes are transit-time isolated from each other for 900 ns, they can be triggered at different times to construct--over an interval as long as $1\text{ }\text{ }\ensuremath{\mu}\mathrm{s}$--the specific load-current time history required for a given experiment. Neptune delivers 1 MJ and 20 MA in a 380-ns current pulse to an $18\text{\ensuremath{-}}\mathrm{m}\mathrm{\ensuremath{\Omega}}$ load; hence Neptune is a megajoule-class 20-MA arbitrary waveform generator. Neptune will allow the international scientific community to conduct dynamic equation-of-state, phase-transition, mechanical-property, and other material-physics experiments with a wide variety of drive-pressure time histories.

Here, we have developed conceptual designs of two petawatt-class pulsed-power accelerators: Z 300 and Z 800. The designs are based on an accelerator architecture that is founded on two concepts: single-stage electrical-pulse compression and impedance matching [Phys. Rev. ST Accel. Beams 10, 030401 (2007)]. The prime power source of each machine consists of 90 linear-transformer-driver (LTD) modules. Each module comprises LTD cavities connected electrically in series, each of which is powered by 5-GW LTD bricks connected electrically in parallel. (A brick comprises a single switch and two capacitors in series.) Six water-insulated radial-transmission-line impedance transformers transport the power generated by the modules to a six-level vacuum-insulator stack. The stack serves as the accelerator’s water-vacuum interface. The stack is connected to six conical outer magnetically insulated vacuum transmission lines (MITLs), which are joined in parallel at a 10-cm radius by a triple-post-hole vacuum convolute. The convolute sums the electrical currents at the outputs of the six outer MITLs, and delivers the combined current to a single short inner MITL. The inner MITL transmits the combined current to the accelerator’s physics-package load. Z 300 is 35 m in diameter and stores 48 MJ of electrical energy in its LTD capacitors. The acceleratormore » generates 320 TW of electrical power at the output of the LTD system, and delivers 48 MA in 154 ns to a magnetized-liner inertial-fusion (MagLIF) target [Phys. Plasmas 17, 056303 (2010)]. The peak electrical power at the MagLIF target is 870 TW, which is the highest power throughout the accelerator. Power amplification is accomplished by the centrally located vacuum section, which serves as an intermediate inductive-energy-storage device. The principal goal of Z 300 is to achieve thermonuclear ignition; i.e., a fusion yield that exceeds the energy transmitted by the accelerator to the liner. 2D magnetohydrodynamic (MHD) simulations suggest Z 300 will deliver 4.3 MJ to the liner, and achieve a yield on the order of 18 MJ. Z 800 is 52 m in diameter and stores 130 MJ. This accelerator generates 890 TW at the output of its LTD system, and delivers 65 MA in 113 ns to a MagLIF target. The peak electrical power at the MagLIF liner is 2500 TW. The principal goal of Z 800 is to achieve high-yield thermonuclear fusion; i.e., a yield that exceeds the energy initially stored by the accelerator’s capacitors. 2D MHD simulations suggest Z 800 will deliver 8.0 MJ to the liner, and achieve a yield on the order of 440 MJ. Z 300 and Z 800, or variations of these accelerators, will allow the international high-energy-density-physics community to conduct advanced inertial-confinement-fusion, radiation-physics, material-physics, and laboratory-astrophysics experiments over heretofore-inaccessible parameter regimes.« less

Advanced tokamak (AT) research seeks to develop steady-state operating scenarios for ITER and other future devices from a demonstrated scientific basis. Normalized target parameters for steady-state operation on ITER are 100% non-inductive current operation with a bootstrap current fraction f BS ≥ 60%, q 95 ∼ 4-5 and G ≡ β N H scaling /q 2 95 ≥ 0.3. Progress in realizing such plasmas is considered in terms of the development of plasma control capabilities and scientific understanding, leading to improved AT performance. NSTX has demonstrated active resistive wall mode stabilization with low, ITER-relevant, rotation rates below the critical value required for passive stabilization. On DIII-D, experimental observations and GYRO simulations indicate that ion internal transport barrier (ITB) formation at rational-q surfaces is due to equilibrium zonal flows generating high local E x B shear levels. In addition, stability modelling for DIII-D indicates a path to operation at β N ≥ 4 with q min ≥ 2, using broad, hollow current profiles to increase the ideal wall stability limit. Both NSTX and DIII-D have optimized plasma performance and expanded AT operational limits. NSTX now has long-pulse, high performance discharges meeting the normalized targets for an spherical torus-based component test facility. DIII-D has developed sustained discharges combining high beta and ITBs, with performance approaching levels required for AT reactor concepts, e.g. β N = 4, H 89 = 2.5, with f BS > 60%. Most importantly, DIII-D has developed ITER steady-state demonstration discharges, simultaneously meeting the targets for steady-state Q ≥ 5 operation on ITER set out above, substantially increasing confidence in ITER meeting its steady-state performance objective.

Zonal flows, thought crucial to the saturation and self-regulation of turbulence and turbulent transport in magnetically confined plasmas, have been observed and characterized in the edge region of DIII-D plasmas. These flows exhibit temperature scaling characteristics and spatial features predicted for geodesic acoustic modes (GAMs), a class of higher-frequency zonal flows seen in nonlinear simulations of plasma turbulence. The zonal flows (GAMs) have been observed in the turbulence flow-field in the radial region 0.85 ≤ r/a ≤ 1.0 via application of time-delay-estimation techniques to two-dimensional measurements of density fluctuations, obtained with beam emission spectroscopy. Spatial and temporal analysis of the resulting flow-field demonstrates the existence of a coherent oscillation (approximately 15 kHz) in the poloidal flow of density fluctuations that has a long poloidal wavelength, possibly m = 0, narrow radial extent (krρi < 0.2), and a frequency that varies monotonically with the local temperature. The approximate effective shearing rate, dvθ/d r, of the flow is of the same order of magnitude as the measured nonlinear decorrelation rate of the turbulence. These characteristics are consistent with predicted features of zonal flows, specifically identified as GAMs, observed in three-dimensional Braginskii simulations of core/edge turbulence.

A271 EXPERIMENTAL CHARACTERIZATION OF COHERENT, RADIALLY-SHEARED ZONAL FLOWS IN THE DIII-D TOKAMAK. Application of time-delay-estimation techniques to two-dimensional measurements of density fluctuations, obtained with beam emission spectroscopy in DIII-D plasmas, has provided temporally and spatially resolved measurements of the turbulence flow-field. Features that are characteristic of self-generated zonal flows are observed in the radial region near 0.85 {

Two-dimensional measurements of density fluctuations are obtained in the radial and poloidal plane of the DIII-D tokamak with the Beam Emission Spectroscopy (BES) diagnostic system. The goals are to visualize the spatial structure and time evolution of turbulent eddies, as well as to obtain the 2D statistical properties of turbulence. The measurements are obtained with an array of localized BES spatial channels configured to image a midplane region of the plasma. 32 channels have been deployed, each with a spatial resolution of about 1 cm in the radial and poloidal directions, thus providing measurements of turbulence in the wave number range 0

Measured turbulence characteristics (correlation lengths, spectra, etc.) in low-confinement (L-mode) and high-performance plasmas in the DIII-D tokamak [Luxon et al., Proceedings Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] show many similarities with the characteristics determined from turbulence simulations. Radial correlation lengths Δr of density fluctuations from L-mode discharges are found to be numerically similar to the ion poloidal gyroradius ρθ,s, or 5–10 times the ion gyroradius ρs over the radial region 0.2

One mechanism for transport of energy and particles in a plasma is by discrete, intermittent, uncorrelated events, often called avalanches. This paper reports observations and quantitative characterization of avalanche events in a magnetically confined plasma. The observations are primarily of electron temperature fluctuations. Avalanches are identified by their large spatial scale, up to the system size, by self-similar behavior in the frequency spectrum and the autocorrelation function and by propagation. The two-point cross-correlation function allows determination of a characteristic velocity, which typically varies from several hundred meters per second in the outer part of the plasma to zero or even inward near the axis. This can be interpreted as resulting from the prevalence of negative avalanches (i.e., holes) near the axis. The presence of a long-tailed probability distribution is indicated by a Hurst parameter (H) in the range 0.80 to 0.95, which becomes smaller in the outer quarter of the plasma radius. Density fluctuation spectra from the plasma core also show self-similar behavior. Power transport estimates show that about half of the heat flux is carried by the avalanche events under conditions with no magnetohydrodynamic activity. These observations are qualitatively similar to results of modeling calculations based on drift wave turbulence. It is reasonable to infer that avalanches are the macroscopic manifestation of turbulence which inherently has a small spatial scale and, thus, allow a local, gyro-Bohm scaling process to show global Bohm-type behavior.

The quiescent double barrier (QDB) regime is a high performance regime recently identified in DIII-D and characterized by a double transport barrier structure (core and edge) that can be maintained for several seconds, often limited only by the pulse length capabilities of the DIII-D hardware. The QDB regime has been sustained for up to 25 τE with fusion performance of up to βNH89≈7. The edge barrier is ELM-free, but modulated by low frequency MHD activity that allows density control via an external cryopump. The core barrier is similar to those seen in previous internal transport barrier experiments, but is maintained without complete stabilization of turbulence. Instead, the turbulence correlation lengths become very short so as to minimize the transport length scales. The two barriers are separated by a region of high transport that is a consequence of a zero-crossing in the E×B shearing rate. These discharges typically possess highly peaked density profiles. This has several implications: narrow bootstrap current profile, reduced beta limit and increased impurity retention. We will report on studies of each of these issues.

Intermittent plasma objects (IPOs) featuring higher pressure than the surrounding plasma, and responsible for ∼50% of the E×BT radial transport, are observed in the scrape off layer (SOL) and edge of the DIII-D tokamak [J. Watkins et al., Rev. Sci. Instrum. 63, 4728 (1992)]. Conditional averaging reveals that the IPOs, produced at a rate of ∼3×103 s−1, are positively charged and also polarized, featuring poloidal electric fields of up to 4000 V/m. The IPOs move poloidally at speeds of up to 5000 m/s and radially with E×BT/B2 velocities of ∼2600 m/s near the last closed flux surface (LCFS), and ∼330 m/s near the wall. The IPOs slow down as they shrink in radial size from 4 cm at the LCFS to 0.5 cm near the wall. The IPOs appear in the SOL of both L and H mode discharges and are responsible for nearly 50% of the SOL radial E×B transport at all radii; however, they are highly reduced in absolute amplitude in H-mode conditions.

High-confinement (H-mode) operation is the choice for next-step tokamak devices based either on conventional or advanced tokamak physics. This choice, however, comes at a significant cost for both the conventional and advanced tokamaks because of the effects of edge localized modes (ELMs). ELMs can produce significant erosion in the divertor and can affect the beta limit and reduced core transport regions needed for advanced tokamak operation. Experimental results from DIII-D [J. L. Luxon et al., Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] this year have demonstrated a new operating regime, the quiescent H-mode regime, which solves these problems. We have achieved quiescent H-mode operation that is ELM-free and yet has good density and impurity control. In addition, we have demonstrated that an internal transport barrier can be produced and maintained inside the H-mode edge barrier for long periods of time (>3.5 s or >25 en...

Tokamak discharges with electron densities as high as 1.4 times the Greenwald density, and good energy confinement, H ITER89P =1.9, were obtained with D 2 gas puffing on DIII-D. The divertor configuration of these discharges, low triangularity with pumping of the private flux region, was important in avoiding a transition to the L-mode or Type III ELM regimes in which energy confinement was reduced. Although these discharges show a decrease in H-mode pedestal energy at high density through reduction in the edge pressure gradient, peaking of the density profile compensated for the effect of this decrease on the overall stored energy. Spontaneous density profile peaking occurred under conditions which enhance the neoclassical Ware pinch. The high density phase was terminated by an internal MHD event that has the characteristics of a neoclassical tearing mode.

A wavelet-transform-based spectral analysis is examined for application to beam emission spectroscopy (BES) data to extract poloidal rotation velocity fluctuations from the density turbulence data. Frequency transfer functions for a wavelet cross-phase extraction method are calculated. Numerical noise is reduced by shifting the data to give an average zero time delay, and the applicable frequency range is extended by numerical oversampling of the measured density fluctuations. This approach offers potential for direct measurements of turbulent transport and detection of zonal flows in tokamak plasma turbulence.

Transport phenomena are studied in Advanced Tokamak (AT) regimes in the DIII-D tokamak [Plasma Physics and Controlled Nuclear Fusion Research, 1986 (International Atomics Energy Agency, Vienna, 1987), Vol. I, p. 159], with the goal of developing understanding and control during each of three phases: Formation of the internal transport barrier (ITB) with counter neutral beam injection taking place when the heating power exceeds a threshold value of about 9 MW, contrasting to co-NBI injection, where Pthreshold

The Z Refurbishment Project was completed in September 2007. Prior to the shutdown of the Z facility in July 2006 to install the new hardware, it provided currents of {le} 20 MA to produce energetic, intense X-ray sources ({approx} 1.6 MJ, > 200 TW) for performing high energy density science experiments and to produce high magnetic fields and pressures for performing dynamic material property experiments. The refurbishment project doubled the stored energy within the existing tank structure and replaced older components with modern, conventional technology and systems that were designed to drive both short-pulse Z-pinch implosions and long-pulse dynamic material property experiments. The project goals were to increase the delivered current for additional performance capability, improve overall precision and pulse shape flexibility for better reproducibility and data quality, and provide the capacity to perform more shots. Experiments over the past year have been devoted to bringing the facility up to full operating capabilities and implementing a refurbished suite of diagnostics. In addition, we have enhanced our X-ray backlighting diagnostics through the addition of a two-frame capability to the Z-Beamlet system and the addition of a high power laser (Z-Petawatt). In this paper, we will summarize the changes made to themore » Z facility, highlight the new capabilities, and discuss the results of some of the early experiments.« less

Beam emission spectroscopy (BES) measures localized density fluctuations in fusion plasma turbulence. A wavelet transform method of spectral analysis is applied to BES data to extract fluctuations without the long time-averaging characteristic of Fourier transform spectral analysis methods. This technique is used to observe an abrupt (∼100 μs) L–H transition on DIII–D. Wavelet-based cross correlation analysis can be used to obtain fluctuations of poloidal rotation velocity, and suggests opportunity for measurement of electrostatic potential fluctuation measurements.

In a magnetically diverted tokamak, the scrape-off layer (SOL) and divertor plasma separates the first wall from the core plasma, intercepting impurities generated at the wall before they reach the core plasma. The divertor plasma can also serve to spread the heat and particle flux over a large area of divertor structure wall using impurity radiation and neutral charge exchange, thus reducing peak heat and particle fluxes at the divertor strike plate. Such a reduction will be required in the next generation of tokamaks, for without it the divertor engineering requirements are very demanding. To successfully demonstrate a radiative divertor, a highly radiative condition with significant volume recombination must be achieved in the divertor, while maintaining a low impurity content in the core plasma. Divertor plasma properties are determined by a complex interaction of classical parallel transport, anomalous perpendicular transport, impurity transport and radiation, and plasma - wall interaction. In this paper we will describe a set of experiments on DIII-D designed to provide detailed two-dimensional documentation of the divertor and SOL plasma. Measurements have been made in operating modes where the plasma is attached to the divertor strike plate and in highly radiating cases where the plasma is detached from the divertor strike plate. We will also discuss the results of experiments designed to influence the distribution of impurities in the plasma using enhanced SOL plasma flow. Extensive modelling efforts will be described which are successfully reproducing attached plasma conditions and are helping to elucidate the important plasma and atomic physics involved in the detachment process.

Fusion produced non-thermal alpha particle radial profile measurements are obtained with the α-CHERS diagnostic in deuterium-tritium (DT) supershot plasmas on the Tokamak Fusion Test Reactor (TFTR). Alpha particles in the energy range 0.15 ≤ Eα ≤ 0.6 MeV are observed spectroscopically over a five point radial profile. The extracted non-thermal alpha signal is ≤ 1% of the background bremsstrahlung intensity for typical total fast alpha densities in the range (0.5-1.0) × 1017 m-3. The profiles obtained in two sets of discharges vary slightly, and are best described by a slowing down alpha distribution subject to neoclassical diffusion plus a small anomalous cross-field diffusion. The data are consistent with an effective anomalous diffusion coefficient in the range 0.00 ≤ Dα,a ≤ 0.10 m2/s, where Dα ,a is constant with alpha energy and with radius

Experience with the α charge exchange recombination spectroscopy and pellet charge exchange confined, nonthermal alpha particle diagnostics over the first two years of Tokamak test fusion reactor (TFTR) D–T operation is summarized. A brief summary of the concept, instrumentation, and analysis techniques for each diagnostic is given, followed by examples of alpha physics results. Issues important to further development of these diagnostic techniques for TFTR and ITER are discussed.

Comprehensive 2D turbulence and eddy flow velocity measurements on DIII-D demonstrate a rapidly increasing turbulence-driven shear flow that develops ∼100 μs prior to the low-confinement (L mode) to high-confinement (H mode) transition and appears to trigger it. These changes are localized to a narrow layer 1-2 cm inside the magnetic boundary. Increasing heating power increases the Reynolds stress, the energy transfer from turbulence to the poloidal flow, and the edge flow shearing rate that then exceeds the decorrelation rate, suppressing turbulence and triggering the transition.

A system of 36 laser-triggered gas switches (LTGS) serves as the last set of command-triggered switches in the 80-TW refurbished Z accelerator at Sandia National Laboratories (SNL). The system is instrumental in the overall performance of Z and allows for flexibility in pulse shaping for various experimental campaigns. It is desirable to push the operating envelope of the switches to higher voltages and currents to allow for a higher peak power to be delivered to the load while at the same time reducing jitter and pre-fire rate for increased precision and reliability. We have accomplished this with a new LTGS (which we refer to as the C1.1 switch) while keeping the overall switch size consistent with the physical space available. Like previous gas switches [1,2], the C1.1 LTGS consists of laser-triggered and cascade electrode sections. However, the C1.1's cascade section is cantilevered and is not supported mechanically near the trigger section. An insulating rod located within the cascade electrodes (which supports the cascade section mechanically) is scalloped to reduce the likelihood of electrical tracking. Material choice for the center support rod was important due to both the mechanical and electrical requirements placed on this component. Mechanical shock testing of the new switch was performed on a shaker table available at SNL prior to installation on Z. All electrical testing of the improved LTGS was performed on the Z machine. To date, we have accumulated over 120 shots on C1.1 switches without a pre-fire. Runtime statistics are determined after each shot and show that the C1.1 switches are very tolerant to voltage and pressure variations, exhibiting median runtimes of ~37-41 ns with a jitter of

The transmission line circuit model of the Z machine is used extensively to aid in the design and analysis of experiments conducted on Z. The circuit model consists of both 1-D and 2-D networks of transmission lines modeling Z's 36 pulselines, vacuum insulator stack, MITLs, vacuum convolute, and load [1].

As the last command-triggered switch in the Refurbished Z accelerator at Sandia National Laboratories (SNL), the laser-triggered gas switch (LTGS) system is instrumental in the overall performance of Z and allows for flexibility in pulse shaping for various experimental campaigns. It is desirable to push the operating envelope of the switch to higher voltages and currents to allow for a higher peak power to be delivered to the load while at the same time reducing jitter and pre-fire rate for increased precision and reliability. We have accomplished this in a version of the LTGS that we call the C1.1 with the constraint of keeping the overall switch size consistent with physical space available. The C1.1 LTGS consists of laser-triggered and cascade portions which has been reported on previously.[1] However, the C1.1 eliminates the trigger plate and supports the cascade section in a cantilevered fashion. Improvements to this iteration of the LTGS were mainly mechanical in nature. Other minor electrical improvements were made to reduce regions of electric field enhancement and to reduce the likelihood of tracking by adding scalloping to the center support rod. Materials choice for the center support rod was important due to both the mechanical and electrical requirements placed on this component. Mechanical shock testing of the improved switch was performed on a shaker table available at SNL prior to installation on Z and showed that the improvements resulted in less displacement of the cantilevered end and less rotation of components in the cascade section. All electrical testing of the improved LTGS was performed on the Z machine. To date, we have accumulated 377 shots on C1.1 switches without a pre-fire. Runtime statistics are determined after each shot and show that the C1.1 switches are very tolerant to voltage and pressure variations exhibiting median runtimes of 43.9 ns with a jitter (1-σ) of 5.4 ns for all switch closures. The modifications made to and the performance results of the improved LTGS system are detailed in this manuscript.

Radial profiles of the density of confined alpha particles with energies in the 0.15 to 0.6 MeV range are spectroscopically observed before and after a sawtooth crash in a TFTR deuterium-tritium plasma. A large drop in the core alpha density is seen, indicating expulsion of alphas from the core to the plasma periphery. The measured changes in the alpha density profiles are consistent with predictions based on the Kolesnichenko sawtooth model in this case

Fusion-produced alpha particles with energy {le}0.7 MeV have been spectroscopically observed in the core of a deuterium-tritium plasma in the TFTR tokamak at alpha densities of 3{times}10{sup 16}m{sup {minus}3}. During a sawtooth-free discharge, the measured energy spectra at {ital r}/{ital a}=0.3 are in good agreement with those predicted on the basis of collisional transport. Time-resolved measurements during the alpha thermalization after alpha source turn-off show decay of the distribution function to lower energies consistent with the classical slowing-down time of 0.5 s.

Peak fusion power production of 6.2 ± 0.4 MW has been achieved in TFTR plasmas heated by deuterium and tritium neutral beams at a total power of 29.5 MW. These plasmas have an inferred central fusion alpha particle density of 1.2 x 1017 m ₋3 without the appearance of either disruptive magnetohydrodynamics events or detectable changes in Alfven wave activity. The measured loss rate of energetic alpha particles agreed with the approximately 5% losses expected from alpha particles which are born on unconfined orbits.

Helium-3 ions with energies from thermal values up to 300 keV produced by ICRF heating have been spectroscopically observed by the α-CHERS diagnostic on the Tokamak Fusion Test Reactor (TFTR). The shape of the spectrum and the temporal and spatial behaviour of the energetic 3He ion emission are consistent with expectations for this ICRF heating case. Because the 3He ion density was similar to the alpha particle density predicted for D-T supershots on TFTR, these observations give confidence that confined alphas produced by D-T fusion reactions can be observed by α-CHERS during D-T operation of TFTR

Repetitive bursting instabilities with strong frequency chirping occur in highbeta, beam-heated plasmas with safety factor q > 1 in the DIII-D tokamak. Although the mode structures differ, in many ways, the off-axis fishbones are similar to the q = 1 fishbones first observed on the Poloidal Divertor Experiment (PDX). The modes are driven by energetic trapped ions at the fastion precession frequency. During a burst, the frequency changes most rapidly as the mode reaches its maximum amplitude. Larger amplitude bursts have larger growth rates and frequency chirps. Unlike PDX fishbones, the decay phase is highly variable and is usually shorter than the growth phase. Also, the waveform is highly distorted by higher harmonics during the latter portion of a burst. The radial mode structure alters its shape during the burst. Like PDX fishbones, the modes expel trapped ions in a ‘beacon’ with a definite phase relationship relative to the mode. Seven types of loss detectors measure the beacon. The losses scale linearly with mode amplitude. The neutron rate changes most rapidly at maximum mode amplitude but, depending on the loss diagnostic, the losses often peak a few cycles later. The non-ambipolar fast-ion losses cause a sudden change in toroidal rotation frequency across the entire plasma. In addition to an overall drop, the neutron signal oscillates in response to the wave. Unlike the beacon of lost particles, which maintains a fixed phase relative to the mode, the phase of the neutron oscillations steadily increases throughout the burst, with the greatest phase slippage occurring in the highly nonlinear phase near maximum mode amplitude US Department of Energy SC-G903402, DE-FC02-04ER54698, DE-FG02-07ER54917

We have developed a new type of convolute called the Clam Shell MITL (CSMITL) to couple multi-level accelerators to a common load. The CSMITL has magnetic nulls only at large radius where the cathode electric field is kept below the threshold for emission, has only a simply connected magnetic topology to avoid plasma motion along magnetic field lines into highly stressed gaps, and has electron injectors that ensure efficient electron flow even in the limiting case of self-limited MITLs. We report the first experimental results on a CSMITL, which convolutes two disk feeds on the Saturn accelerator into a single disk feed. Experiments with a high impedance electron beam load operating at twice the self-limited impedance of the CSMITL confirm key design features and demonstrate robust operation.

Utilizing an array of new diagnostics and simulation/modelling techniques, recent DIII-D experiments have elucidated a variety of energetic ion transport behaviour in the presence of instabilities ranging from large-scale sawteeth to fine spatial scale microturbulence. Important new insights include sawteeth, such as those of the ITER baseline scenario, causing major redistribution of the energetic ion population; high levels of transport induced by low-amplitude Alfven eigenmodes can be caused by the integrated effect of a large number of simultaneous modes; ´ and microturbulence can contribute to the removal of alpha ash while having little effect on fusion alphas. This paper provides an overview of recent and upcoming results from the DIII-D Energetic Particles research programme. US Department of Energy SC-G903402, DE-FC02-04ER54698, DE-FG02-89ER53296, DE-FG02-08ER54999, DE-AC05-00OR22725, DE-AC02-09CH11466, DE-FG03-08ER54984, DE-FG02-07ER54917

A 6.1-MV, 0.79-MA laser-triggered gas switch (LTGS) is used to synchronize the 36 modules of the $Z$ machine at Sandia National Laboratories. Each module includes one switch, which serves as the last command-fired switch of the module, and hence is used to determine the time at which each module electrically closes relative to the other modules. The switch is $\ensuremath{\sim}81\mathrm{\text{\ensuremath{-}}}\mathrm{cm}$ in length, $\ensuremath{\sim}45\mathrm{\text{\ensuremath{-}}}\mathrm{cm}$ in diameter, and is immersed in mineral oil. The outer switch envelope consists of six corrugated monomer-cast acrylic insulators and five contoured stainless-steel rings. The trigger electrodes are fabricated from copper-infused tungsten. The switch is pressurized with several atmospheres of sulfur hexafluoride (${\mathrm{SF}}_{6}$), which is turbulently purged within 2 seconds after every shot. Each switch is powered from a 6-MV, 0.78-MJ Marx generator which pulse charges a 24-nF intermediate-store water capacitor in $1.4\mathrm{\text{\ensuremath{-}}}\ensuremath{\mu}\mathrm{s}$. Closure of the switch allows power to flow into pulse-forming transmission lines. The power pulse is subsequently compressed by water switches, which results in a total accelerator output power in excess of 70-TW. A previous version of the LTGS performed exceptionally at a 5.4-MV, 0.7-MA level on an engineering test module used for switch development. It exhibited a $1\mathrm{\text{\ensuremath{-}}}\ensuremath{\sigma}$ jitter of $\ensuremath{\sim}5\text{ }\text{ }\mathrm{ns}$, a prefire and flashover rate less than 0.1%, and a lifetime in excess of 150 shots. When installed on the $Z$ accelerator, however, the switch exhibited a prefire probability of $\ensuremath{\sim}3%$, a flashover probability of $\ensuremath{\sim}7%$, and a 15-ns jitter. The difference in performance is attributed to several factors such as higher total charge transfer, exposure to more debris, and more stressful dynamic mechanical loading upon machine discharge. Under these conditions, the replacement lifetime was less than ten shots. Since refurbishment of $Z$ in October 2007, there have been three LTGS design iterations to improve the performance at 6.1-MV. The most recent design exhibits a prefire rate of less than 0.1%, a flashover rate of $\ensuremath{\sim}0.2%$, a single switch jitter of $\ensuremath{\sim}6\mathrm{\text{\ensuremath{-}}}\mathrm{ns}$, and a lifetime of greater than 75 shots. Modifications to achieve the performance improvement are detailed in this article.