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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.

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.

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

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

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.

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.

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.

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.

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.

We demonstrate that a wide variety of current-pulse shapes can be generated using a linear-transformer-driver (LTD) module that drives an internal water-insulated transmission line. The shapes are produced by varying the timing and initial charge voltage of each of the module’s cavities. The LTD-driven accelerator architecture outlined in [Phys. Rev. ST Accel. Beams 10, 030401 (2007)PRABFM1098-440210.1103/PhysRevSTAB.10.030401] provides additional pulse-shaping flexibility by allowing the modules that drive the accelerator to be triggered at different times. The module output pulses would be combined and symmetrized by water-insulated radial-transmission-line impedance transformers [Phys. Rev. ST Accel. Beams 11, 030401 (2008)PRABFM1098-440210.1103/PhysRevSTAB.11.030401].

In the presentation we present a survey of the application driven development of new high-current (25+ MA), fast risetime (~ 100 ns) pulsed-power technology at the Sandia National Laboratories (SNL), but this paper only focuses on megagauss magnetic field driver development in order to preserve first-publication rights for the co-authors with the other technologies. Thus in the presentation we note that the Z accelerator has fired nearly 200 shots since the upgrade, and is now being operated at 4 to 5 shots per week. Initial failures after the upgrade have been resolved with improvements in the 6-MV laser-triggered switches, the oil water barriers, and the insulator stack. With the upgrade Z is also able to produce tailored current profiles for isentropic compression of materials using staggered timing of the laser triggered switches. We also describe the Linear Transformer Driver (LTD) technology that is being investigated as a potential follow-on pulsed power technology for the design of the next larger machine that has a high repetition-rate (up to 0.1 Hz) capability. Ten 1-MA LTD modules have been fabricated and are being setup as a 1 MV, 1 MA sub-module prototype. A lower-current, more compact LTD design is also being tested for radiography applications. For both of these designs 200-kV, 25-kA gas switches are being developed and tested to both increase switch lifetime and to minimize cost. A compact, portable, 2-MA, 2 μsec driver, which we describe in this paper, is being developed to drive single-turn magnetic field coils for laser-plasma experiments in vacuum. (4 pages)

Advanced Tokamak research in DIII-D seeks to develop a scientific basis for steady-state high-performance tokamak operation. Fully noninductive (fNI≈ 100%) in-principle steady-state discharges have been maintained for several confinement times. These plasmas have weak negative central shear with qmin≈ 1.5-2, βN≈ 3.5, and large, well-aligned bootstrap current. The loop voltage is near zero across the entire profile. The remaining current is provided by neutral beam current drive (NBCD) and electron cyclotron current drive (ECCD). Similar plasmas are stationary with fNI≈ 90-95% and duration up to 2 s, limited only by hardware. In other experiments, βN≈ 4 is maintained for 2 s with internal transport barriers, exceeding previously achieved performance under similar conditions. This is allowed by broadened profiles and active magnetohydrodynamic instability control. Modifications now underway on DIII-D are expected to allow extension of these results to higher performance and longer duration. A new pumped divertor will allow density control in high triangularity double-null divertor configurations, facilitating access to similar in-principle steady-state regimes with βN> 4. Additional current drive capabilities, both off-axis ECCD and on-axis fast wave current drive (FWCD), will increase the magnitude, duration, and flexibility of externally driven current. © 2006 Elsevier B.V. All rights reserved.

Robust validation of predictive turbulent transport models requires quantitative comparisons to experimental measurements at multiple levels, over a range of physically relevant conditions. Toward this end, a series of carefully designed validation experiments has been performed on the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] to obtain comprehensive multifield, multipoint, multiwavenumber fluctuation measurements and their scalings with key dimensionless parameters. The results of two representative validation studies are presented: an elongation scaling study performed in beam heated L-mode discharges and an electron heating power scan performed in quiescent H-mode (QH-mode) discharges. A 50% increase in the elongation κ is observed to lead to a ∼50% increase in energy confinement time τe and accompanying decrease in fluctuation levels, qualitatively consistent with a priori theoretical predictions and nonlinear GYRO [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] simulation...

This paper presents new measurements of the cross-phase angle, αneTe, between long-wavelength (kθρs

New measurements show that long-wavelength k s 0.5 electron temperature fluctuations can play an important role in determining electron thermal transport in low-confinement mode L-mode tokamak plasmas. In neutral beam-heated L-mode tokamak plasmas, electron thermal transport and the amplitude of long-wavelength electron temperature fluctuations both increase in cases where local electron cyclotron heating ECH is used to modify the plasma profiles. In contrast, the amplitude of simultaneously measured long-wavelength density fluctuations does not significantly increase. Linear stability analysis indicates that the ratio of the trapped electron mode TEM to ion temperature gradient ITG mode growth rates increases in the cases with ECH. The increased importance of the TEM drive relative to the ITG mode drive in the cases with ECH may be associated with the increases in electron thermal transport and electron temperature fluctuations. © 2010 American Institute of Physics. doi:10.1063/1.3318469 The importance of long-wavelength turbulent fluctuations in determining electron thermal transport in the core of tokamak plasmas is of considerable interest to the field of magnetically confined fusion. In the absence of magnetohydrodynamic MHD instabilities, small amplitude 1%, low frequency ci, where ci is the ion cyclotron frequency turbulent fluctuations associated with drift-wavetype instabilities are widely believed to drive radial transport. 1 The turbulence driven transport can lead to heat and particle losses that reduce the performance of magnetically confined fusion devices. 2 One expectation from driftwave theory that can be tested directly is that the ratio of electron temperature and density fluctuation amplitudes, T ˜ e /Te /n /n, should scale with the ratio of linear growth rates of the trapped electron mode TEM and the ion temperature gradient ITG mode, TEM/ITG. 3,4 To test this, multifield fluctuation measurements are needed. Electron temperature fluctuations and density fluctuations have previously been studied in tokamak experiments 5‐7 during electron cyclotron heating ECH, but the two fluctuating fields were not measured simultaneously. We present here the first simultaneous measurements of long-wavelength electron temperature fluctuations and density fluctuations in the core of an L-mode tokamak plasma heated with neutral beams and ECH. The experiment was performed on the DIII-D tokamak major radius R=1.67 m, minor radius a=0.61 m. 8 The discharges have magnetic field BT2.0 T, plasma current Ip=1 MA, with edge safety factor q955.2 in the time period of interest, and are inner wall limited. Afirst discharge is heated with 2.5 MW of neutral beam power injected in the direction of the plasma current beginning early in time t=300 ms. The L-mode plasma of interest 1500t 1800 ms is sawtooth-free and MHD-free. In a second

The power required to induce a bifurcation from a low-confinement mode to a high-confinement mode in DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] plasmas is found to depend sensitively on the injected neutral beam torque and consequent toroidal rotation. Plasmas exhibit a factor of 2–4 reduction in this power threshold, dependent on ion ∇B drift direction. Correlated with this change, turbulence velocity measurements near 0.9

The deployment of multiple high-resolution, spatially localized fluctuation diagnostics on the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] opens the door to a new level of core turbulence model validation. Toward this end, the implementation of synthetic diagnostics that model physical beam emission spectroscopy and correlation electron cyclotron emission diagnostics is presented. Initial results from their applications to local gyrokinetic simulations of two locations in a DIII-D L-mode discharge performed with the GYRO code [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] are also discussed. At normalized toroidal flux ρ=0.5, we find very good agreement between experiment and simulation in both the energy flows and fluctuation levels measured by both diagnostics. However, at ρ=0.75, GYRO underpredicts the observed energy flows by roughly a factor of 7, with rms fluctuation levels underpredicted by a factor of 3. Interestingly, at both locations we find good agreement in the sha...

Recent studies on the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] have elucidated key aspects of the dependence of stability, confinement, and density control on the plasma magnetic configuration, leading to the demonstration of nearly noninductive operation for >1 s with pressure 30% above the ideal no-wall stability limit. Achieving fully noninductive tokamak operation requires high pressure, good confinement, and density control through divertor pumping. Plasma geometry affects all of these. Ideal magnetohydrodynamics modeling of external kink stability suggests that it may be optimized by adjusting the shape parameter known as squareness (ζ). Optimizing kink stability leads to an increase in the maximum stable pressure. Experiments confirm that stability varies strongly with ζ, in agreement with the modeling. Optimization of kink stability via ζ is concurrent with an increase in the H-mode edge pressure pedestal stability. Global energy confinement is optimized at the lowest ζ tested, wi...

The spatial transfer function for the beam emission spectroscopy (BES) diagnostic is critical to quantitatively interpret local density fluctuation measurements. A three-dimensional geometrical calculation of the spatial transfer function is presented for the upgraded BES diagnostic on DIII-D to determine its spatial resolution and wave-number sensitivity. The spatial transfer function calculation for the BES system on DIII-D incorporates the high speed (f∕2) collection optics, an optical fiber bundle, neutral beam-sight line geometry, the neutral beam cross-section intensity profile, magnetic field pitch angle, as well as atomic physics of the finite atomic transition time of the collisionally excited beam atoms. The resulting imaged volumes for each BES channel typically have ∼1–2cm radial and poloidal resolutions. In addition, the viewing volume is nominally aligned along a magnetic field line to minimize spatial smearing of the field-aligned turbulent eddies. This calculation is crucial for performing...

High temporal and spatial resolution measurements in the boundary of the DIII-D tokamak show that edge-localized modes (ELMs) are produced in the low field side, are poloidally localized and are composed of fast bursts (∼20 to 40μs long) of hot, dense plasma on a background of less dense, colder plasma (∼5×1018m−3, 50 eV) possibly created by the bursts themselves. The ELMs travel radially in the scrape-off layer (SOL), starting at the separatrix at ∼450m∕s, and slow down to ∼150m∕s near the wall, convecting particles and energy to the SOL and walls. The temperature and density in the ELM plasma initially correspond to those at the top of the density pedestal but quickly decay with radius in the SOL. The temperature decay length (∼1.2 to 1.5 cm) is much shorter than the density decay length (∼3 to 8 cm), and the latter decreases with increasing pedestal (and SOL) density. The local particle and energy flux (assuming Ti=Te) at the midplane wall during the bursts are 10% to 50% (∼1 to 2×1021m−2s−1) and 1% to...

Long wavelength turbulence as well as heat and momentum transport are significantly reduced in the DIII-D tokamak [Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] as a result of neon seeding of a low confinement mode negative central shear discharge. Correspondingly, the energy confinement time increases by up to 80%. Fully saturated turbulence measurements near ρ=0.7 (ρ=r/a) in the wave number range 0.1⩽k⊥ρs⩽0.6, obtained with beam emission spectroscopy, exhibit a significant reduction of fluctuation power after neon injection. Fluctuation measurements obtained with far infrared scattering also show a reduction of turbulence in the core, while the Langmuir probe array measures reduced particle flux in the edge and scrape-off layer. Gyrokinetic linear stability simulations of these plasmas are qualitatively consistent, showing a reduction in the growth rate of ion temperature gradient driven modes for 0