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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

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.

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.

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

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

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

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.

Impurity seeding in both the Joint European Torus (JET) and DIII-D tokamaks has produced L-mode discharges with confinement enhancements comparable to H-mode and a near doubling of the core ion temperature when compared,to similar unseeded discharges. Although Z(eff) increases with the neon injection, the total neutron rate is as high, or higher, than reference discharges and the calculated thermal neutron rate increases dramatically in both devices. Modelling with the gyrokinetic simulation code shows a reduction in low k turbulence growth rates with neon injection decreasing to less than the E x B shearing rate, consistent with stabilization of ion temperature gradient modes in both JET and DIII-D. Reductions in ion thermal diffusivity are also observed with impurity seeding. Neoclassical m/n = 3/2 tearing modes limit the duration of best performance in DIII-D with neon injection, while n = 1 and n = 2 magnetohydrodynamic modes limit the performance in JET.

Imploding wire arrays on the 20 MA Z generator have recently provided some of the most powerful and energetic laboratory sources of multi-keV photons, including ∼375 kJ of Al K-shell emission (hν ∼ 1–2 keV), ∼80 kJ of stainless steel K-shell emission (hν ∼ 5–9 keV) and a kJ-level of Mo K-shell emission (hν ∼ 17 keV). While the global implosion dynamics of these different wire arrays are very similar, the physical process that dominates the emission from these x-ray sources fall into three broad categories. Al wire arrays produce a column of plasma with densities up to ∼3 × 1021 ions/cm3, where opacity inhibits the escape of K-shell photons. Significant structure from instabilities can reduce the density and increase the surface area, therefore increase the K-shell emission. In contrast, stainless steel wire arrays operate in a regime where achieving a high pinch temperature (achieved by thermalizing a high implosion kinetic energy) is critical and, while opacity is present, it has less impact on the pinch...

The Tokamak Fusion Test Reactor (TFTR) has performed initial high-power experiments with the plasma fueled by deuterium and tritium to nominally equal densities. Compared to pure deuterium plasmas, the energy stored in the electron and ions increased by ~20%. These increases indicate improvements in confinement associated with the use of tritium and possibly heating of electrons by α-particles.

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...

Reversed shear Alfv́n eigenmodes (RSAEs) are typically thought of as being localized near the minima in the magnetic safety factor profile, however, their spatial coupling to global toroidal Alfv́n eigenmodes (TAEs) has been observed in DIII-D discharges. For a decreasing minimum magnetic safety factor, the RSAE frequency chirps up through that of stable and unstable TAEs. Coupling creates a small gap at the frequency degeneracy point forming two distinct global modes. The core-localized RSAE mode structure changes and becomes temporarily global. Similarly, near the mode frequency crossing point, the global TAE extends deeper into the plasma core. The frequency splitting and spatial structure of the two modes throughout the various coupling stages, as measured by an array of internal fluctuation diagnostics, are in close agreement with linear ideal MHD calculations using the NOVA code. The implications of this coupling for eigenmode stability is also investigated and marked changes are noted throughout the coupling process. © 2007 American Institute of Physics.

Reversed shear Alfven eigenmodes (RSAE) that were observed in the Joint European Torus (JET) [P. H. Rebut and B. E. Keen, Fusion Technol.11, 13 (1987)] and DIII-D [J. L. Luxon, Nucl. Fusion42, 614 (2002)] are studied with the ideal magnetohydrodynamic code NOVA-K [C. Z. Cheng, Phys. Rep.211, 1 (1992)]. It was found that the frequency behavior of the RSAEs can be described accurately by the NOVA-K code when plasma compressibility effects and toroidal plasma rotation are taken into account. For the mode activity on JET, the calculated drive exceeds the mode damping rate, consistent with experimental observations, while on DIII-D the growth rate from neutral beam ions for modes with high toroidal mode numbers is insufficient to account for the excitation of the modes and a major part of the drive comes from the background plasma.

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...

This work was funded by the U.S. Department of Energy under Grant Nos. DE-FC02-04ER54698, DE-FG02- 04ER54758, DE-FG03-01ER54615, W-7405-ENG-48, DEFG03-96ER54373, DE-FG02-89ER53297, DE-AC05- 00OR22725, and DE-AC04-94AL85000.

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