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Collective Thomson scattering (CTS) is a diagnostic method that measures the ion velocity distribution of a plasma. CO

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.

A heterodyne detection scheme is combined with a 10.59 μm CO

A full-scale ITER toroidal interferometer and polarimeter (TIP) prototype, including an active feedback alignment system, has been installed and tested on the DIII-D tokamak. In the TIP prototype, a two-color interferometry measurement of line-integrated density is carried out at 10.59 μm and 5.22 μm using a CO2 and quantum cascade laser, respectively, while a separate polarimetry measurement of the plasma-induced Faraday effect is made at 10.59 μm. The TIP prototype is equipped with a piezo tip/tilt stage active feedback alignment system that minimizes noise in the measurement and keeps the diagnostic aligned throughout DIII-D discharges. The measured phase resolution for the polarimeter and interferometer is 0.05° (100 Hz bandwidth) and 1.9° (1 kHz bandwidth), respectively. The corresponding line-integrated density resolution for the vibration-compensated interferometer is δnL = 1.5 × 1018 m-2, and the magnetic field-weighted line-integrated density from the polarimeter is δnBL = 1.5 × 1019 Tm-2. Both interferometer and polarimeter measurements during DIII-D discharges compare well with the expectations based on calculations using Thomson scattering measured density profiles and magnetic equilibrium reconstructions. Additionally, larger bandwidth interferometer measurements show that the diagnostic is a sensitive monitor of core density fluctuations with demonstrated measurements of Alfven eigenmodes and tearing modes.

Translatable in-vessel mirrors have enabled the DIII-D Thomson scattering system to diagnose the divertor plasma in high triangularity shaped plasmas. Previous divertor Thomson scattering measurements in DIII-D were restricted to spatial locations along a Nd:YAG laser beam that was directed through a vertical port. This only allowed measurements to be made in low triangularity shaped plasmas. The new mirrors re-route the laser underneath floor tiles to a position of smaller major radius as necessary for high triangularity plasmas. New in-vessel collection optics transmit scattered light from regions inaccessible to external lenses. Damage to mirrors and high stray light levels are challenges that were overcome to successfully make these measurements. Through the careful use of baffles and light shields, stray light leakage into polychromator detector channels was reduced to negligible levels, allowing temperature measurements below 1 eV. The system is described and the initial results presented.

For future burning plasma experiments, all diagnostics must be re-evaluated in terms of their measurement capabilities and robustness in the harsh, high-temperature environment. This is certainly true for interferometry measurements where conventional approaches may not be ideal and interpretation may require modification due to high plasma temperatures. Optimizing these systems to provide maximum information is crucial to understanding burning plasma dynamics. This paper explores a variety of phase measurement techniques for the main body and divertor regions that can be utilized on fusion plasma experiments like ITER and beyond.

Recent upgrades to many of the diagnostic systems on DIII-D (Luxon J.L. 2002 Nucl. Fusion 42 614) such as the CO2 interferometer, far-infrared scattering, beam-emission spectroscopy (BES), and quadrature reflectometer have significantly extended their capabilities and made possible the experimental study of Alfven eigenmodes (AEs) through observation of the AE induced density perturbation. Measurements have revealed the presence of several different classes of AEs in DIII-D discharges including the toroidal Alfven eigenmode (TAE), reverse shear AE (RSAE or Alfven cascade) and ellipticity induced Alfven eigenmode. Based on a simple model for the RSAE frequency, a sensitive diagnostic for the evolution of the minimum magnetic safety factor (qmin) is presented and results are compared with motional Stark effect (MSE) measurements. Strong localization of high toroidal mode number RSAEs to regions near the minimum of the magnetic safety factor is exhibited on the CO2 interferometer and BES measurements. Based on this observation, a method for providing constraints on the radial location of qmin is demonstrated and a favourable comparison to MSE measurements is made. Detailed measurements of TAEs using a new all-digital large bandwidth two-colour CO2 interferometer system show a strong asymmetry between vertical and radial viewing interferometer chords confirming previously reported results. Additionally, effects related to line-integrated observations are clearly illustrated by comparison to local BES measurements and potential issues related to this are discussed.

Measurements are presented of toroidicity-induced (TAEs) and reverse shear (RSAEs) Alfven eigenmodes made using the standard two-colour CO2 interferometer on DIII-D modified for increased bandwidth. Typical values of the effective line-integrated density perturbation in DIII-D are found to be d(nL)/nL ~ 10−3, and comparisons are made with NOVA calculations as well as magnetic measurements. There is a strong difference in the measured power spectrum between vertical and radial chords through the plasma. On average, vertical views are characterized by a larger line-integrated density perturbation due to TAEs than radial chords. Radial chords, however, can be used much more reliably than vertical chords to identify the presence of RSAEs in the plasma—a result found to be due to the radially localized nature of these modes. In general, the apparent amplitude of the observed modes for both TAE and RSAE is found to be highly dependent on viewing location.

Recent technological developments make a laser Thomson scattering diagnostic, operating in the incoherent electron scattering regime, a good candidate to accomplish measurements of fast electron dynamics (i.e., fast equilibrium changes, turbulence, and electron fluctuations) in high temperature plasmas. Pulse repetition rates for current generation Thomson scattering diagnostics have been limited to about 100 Hz, with the possible exception of “burst” modes in which multiple laser systems provide a limited sequence of closely spaced pulses. To overcome this limitation in laser capability, we propose that recent advances in compact, high power, diode-pumped solid state lasers be applied to a fast Thomson scattering diagnostic for fusion research. To illustrate the possibilities, we present an overview of a diagnostic system designed for the Madison Symmetry Torus reversed field pinch. The operational goal for this single-laser system is to measure Te, ne, and pe with a measurement rate of at least 10 kHz...

Details of the low-high (L-H) confinement transition are studied by a new technique which allows for an arbitrarily slow transition between the L- and H-modes on DIII-D. During the transition, the plasma is in an intermediate state (IM-mode) of temperature, density, confinement and edge shear flow. The IM-mode is characterized by periodic bursts of an edge instability, governed by relaxation oscillations, which evolve into type III edge localized modes (ELMs) as the neutral beam heating power is raised. An ELM-free H-mode is achieved when the edge pressure gradient is large enough to support shear flow sufficient to quell the IM-mode edge instability.

Recent measurements from the DIII-D tokamak indicate that shear in the group velocity of the edge density fluctuations is at least partly responsible for the factor 2 change in the H-mode power threshold that is observed when the direction of the ion ∇B drift relative to the X-point location is reversed. Spatially resolved edge density fluctuation measurements show a change in the poloidal group velocity of the fluctuations when the ∇B drift direction was changed, even though the edge profiles of density and temperature remained nearly the same. High (low) shear in the poloidal velocity is associated with a low (high) power threshold.

The resistive X point mode is shown to be the dominant mode in boundary plasmas in X point divertor geometry. The poloidal fluctuation phase velocity from the simulation results of the resistive X point turbulence shows a structure across the separatrix that is experimentally measured in many fusion devices. The fluctuation phase velocity is larger than the E × B velocity in both L and H mode phases. It is also demonstrated that there is a strong poloidal asymmetry of particle flux in the proximity of the separatrix. Turbulence suppression in the L-H transition results when sources of energy and particles drive sufficient gradients, as in the experiments.

A full-scale 120 m path length ITER toroidal interferometer and polarimeter (TIP) prototype, including an active feedback alignment system, has been constructed and undergone initial testing at General Atomics. In the TIP prototype, two-color interferometry is carried out at 10.59 μm and 5.22 μm using a CO2 and quantum cascade laser (QCL) respectively while a separate polarimetry measurement of the plasma induced Faraday effect is made at 10.59 μm. The polarimeter system uses co-linear right and left-hand circularly polarized beams upshifted by 40 and 44 MHz acousto-optic cells respectively, to generate the necessary beat signal for heterodyne phase detection, while interferometry measurements are carried out at both 40 MHz and 44 MHz for the CO2 laser and 40 MHz for the QCL. The high-resolution phase information is obtained using an all-digital FPGA based phase demodulation scheme and precision clock source. The TIP prototype is equipped with a piezo tip/tilt stage active feedback alignment system responsible for minimizing noise in the measurement and keeping the TIP diagnostic aligned indefinitely on its 120 m beam path including as the ITER vessel is brought from ambient to operating temperatures. The prototype beam path incorporates translation stages to simulate ITER motion through a bake cycle as well as other sources of motion or misalignment. Even in the presence of significant motion, the TIP prototype is able to meet ITER's density measurement requirements over 1000 s shot durations with demonstrated phase resolution of 0.06° and 1.5° for the polarimeter and vibration compensated interferometer respectively. TIP vibration compensated interferometer measurements of a plasma have also been made in a pulsed radio frequency device and show a line-integrated density resolution of m−2.

Edge conditions in DIII-D are being quantified in order to provide insight into the physics of the H?mode regime. Several studies show that electron temperature is not the key parameter that controls the L-H transition. Gradients of edge temperature and pressure are much more promising candidates for elements of such parameters. They systematically increase during the L phases of discharges which make a transition to H?mode, and these increases are typically larger than the increases in the underlying quantities. The quality of H?mode confinement is strongly correlated with the height of the H?mode pedestal for the pressure. The gradient of the pressure is limited by MHD modes, in particular by ideal kink ballooning modes with finite mode number n. For a wide variety of discharges, the width of the barrier for electron pressure is well described by a relationship that is proportional to (?pedp)1/2. A new regime of confinement, called the quiescent H?mode, which provides steady state operation with no ELMs, low radiated power and normal H?mode confinement, has been discovered. A coherent edge MHD mode provides adequate particle transport to control the plasma density while permitting the pressure pedestal to remain almost identical to that observed in ELMing discharges.

Detailed measurements in two dimensions by probes and Thomson scattering reveal unexpected local electric potential and electron pressure (pe) maxima near the divertor X point in L-mode plasmas in the DIII-D tokamak [J. L. Luxon and L. G. Davis, Fusion Technol. 8, 441 (1985)]. The potential drives E×B circulation about the X point, thereby exchanging plasma between closed and open magnetic surfaces at rates that can be comparable to the total cross-separatrix transport. The potential is consistent with the classical parallel Ohm’s law. A simple model is proposed to explain the pressure and potential hills in low power, nearly detached plasmas. Recent two-dimensional edge transport modeling with plasma drifts also shows X-point pressure and potential hills but by a different mechanism. These experimental and theoretical results demonstrate that low power tokamak plasmas can be far from poloidal uniformity in a boundary layer just inside the separatrix. Additional data, although preliminary and incomplete, ...

Electric potential, electron temperature, and electron density were measured in two dimensions (R,Z) throughout the divertor and X-point in DIII-D tokamak plasmas. An electric potential hill (∼100 eV) and an associated electron pressure hill were discovered at the divertor X-point in L-mode plasmas. The potential hill extends previously reported divertor E x B circulation, convecting particles, energy and toroidal momentum into and out of closed magnetic surfaces, and contributes significantly to transport across the boundary. The potential is explained by classical parallel (to B) plasma physics, when the X-point T i is clamped lower than upstream T i . The low X-point T i state might be self-sustaining at low heating power due to the same E x B circulation. We speculate that if the circulation transport is incompatible with H-mode, then the spontaneous L-H transition might require as a precondition that the X-point T i become equalized on the near-separatrix magnetic surfaces.

Analysis of DIII-D discharge data with fluid plasma and Monte Carlo neutrals transport codes reveals that core particle fueling stays relatively constant between the L-mode and the ELM-free H-mode phase immediately following the L–H transition. This indicates that in the ELM-free phase nearly all of the increase in plasma electron density comes from a decrease in the cross-field transport rate and an increase in the impurity influx. This result differs from conclusions of previous work in that the effects of the thinner H-mode scrape-off-layer do not appear to be as important in a plasma that is fueled primarily from divertor recycling as would be expected if the fueling from limiter recycling were dominant. In both L-mode and H-mode the calculated core particle confinement times are less than, but within 50% of, the corresponding energy confinement times.

A heterodyne detection scheme is combined with a 10.59 μm CO2 laser dispersion interferometer for the first time to allow large bandwidth measurements in the 10-100 MHz range. The approach employed utilizes a 40 MHz acousto-optic cell operating on the frequency doubled CO2 beam which is obtained using a high 2nd harmonic conversion efficiency orientation patterned gallium arsenide crystal. The measured standard deviation of the line integrated electron density equivalent phase resolution obtained with digital phase demodulation technique, is 4 × 1017 m−2. Air flow was found to significantly affect the baseline of the phase signal, which an optical table cover was able to reduce considerably. The heterodyne dispersion interferometer (DI) approach is found to be robustly insensitive to motion, with measured phase shifts below baseline drifts even in the presence of several centimeters of retroreflector induced path length variations. Plasma induced dispersion was simulated with a wedged ZnSe plate and the m...

The specific size and structure of the edge current profile has important effects on the magnetohydrodynamic stability and ultimate performance of many advanced tokamak (AT) operating modes. This is true for both bootstrap and externally driven currents that may be used to tailor the edge shear. Absent a direct local measurement of j(r), the best alternative is a determination of the poloidal field. Measurements of the precision (0.1°–0.01° in magnetic pitch angle and 1–10 ms) necessary to address issues of stability and control and provide constraints for EFIT are difficult to do in the region of interest (ρ=0.9–1.1). Using Zeeman polarization spectroscopy of the 2S–2P lithium resonance line emission from the DIII-D LIBEAM [D. M. Thomas, Rev. Sci. Instrum. 66, 806 (1995); D. M. Thomas, A. W. Hyatt, and M. P. Thomas, Rev. Sci. Instrum. 61, 340 (1990)] measurements of the various field components may be made to the necessary precision in exactly the region of interest to these studies. Because of the negli...

Theories predict that neutrals play a role in the low to high (L-H) confinement mode transition in tokamak plasmas via charge exchange damping and other effects. Previous estimates of neutral damping have been based on calculations of the edge neutral density. This work introduces a new method of measuring the neutral density near the X point, where simulations predict it to be a maximum. The technique employed uses Dα light from a TV camera reconstructed onto a poloidal plane, along with Thomson scattering measurements of the electron temperature and density. Measured neutral densities span the range 109-1013cm-3. Good agreement, considering the neutral density error bars, is found between the measurements and the 2-D simulations. This work represents the first step in verifying previous 2-D simulations and in corroborating previous conclusions that the neutral damping is large enough to play a role in the L-H transition process.

In order to study possible effects of neutrals on the power threshold for transition from L-mode to H-mode confinement in DIII-D, the problem of reconstructing neutral particle distributions inside the separatrix from available edge, scrape-off layer, and divertor plasma diagnostic data is addressed. Neutral particle profiles in the shear layer are reconstructed primarily with the DEGAS code using plasma parameters obtained by fitting discharge data with the B2.5 code. A series of discharges in which the ion grad B drift direction is toward the X-point is analyzed. For reversed Bt, the L to H transition power threshold is 2–4 times higher than normal Bt discharges at the same density: one discharge of this type is analyzed. The variation of the L–H transition power threshold divided by the density is seen to be correlated with the neutral density scale length in the shear layer for both directions of the magnetic field. Radial and poloidal neutral particle profile results for three low density discharges in which the L–H transition power threshold varies from 1.3 to 5.3 MW are presented.

To study the effect of edge neutrals on the low-to-high confinement transition threshold, a broad range of plasma discharges has been analyzed. These discharges vary by gas puffing and pumping rates, position of the X point, and line-averaged density. It is shown that the determination of the neutral density (or neutral pressure) in the scrape-off layer (SOL) can give a misleading indication of the neutral population inside the separatrix. An increase of neutral density in the SOL creates an increase of plasma density that, in turn, increases the opacity to the neutrals and results in reduced neutral penetration. At a constant magnetic field, the transition power divided by the density appears to be a function of a single parameter measuring the neutrals effect. From this analysis, this parameter cannot be uniquely identified. For instance, it may be the radial decay length of the neutral profile or the charge-exchange damping rate at about r/a≈0.95. A similar correlation exists between these neutral para...

The scalings of heat transport with safety factor (q), normalized collisionality (v), plasma beta ({beta}), and relative gyroradius ({rho}*) have been measured on the DIII-D tokamak. The measured {rho}* {beta} and v scalings of heat transport indicate that E x B transport from drive wave turbulence is a plausible basis for anomalous transport. For high confinement (H) mode plasmas where the safety factor was varied at fixed magnetic shear, the effective (or one-fluid) thermal diffusivity was found to scale like {chi}{sub eff} {proportional_to} q{sup 2.3{+-}0.64}, with the ion and electron fluids having the same q scaling to within the experimental errors except near the plasma edge. The scaling of the thermal confinement time with safety factor was in good agreement with this local transport dependence, {tau}{sub th} {proportional_to} q{sup {minus}2.42{+-}0.31}; however, when the magnetic shear was allowed to vary to keep q{sub 0} fixed during the (edge) safety factor scan, a weaker global dependence was observed, {tau}{sub th} {proportional_to} q{sub 95}{sup {minus}1.43{+-}0.23}. This weaker dependence was mainly due to the change in the local value of q between the two types of scans. The combined {rho}*, {beta}, v and q scalings of heat transport for H-mode plasmas on DIII-D reproduce the empirical confinement scaling using physical (dimensionless) parameters with the exception of weaker power degradation.

The H-mode power threshold has a weak but positive dependence when the ion drift is away from the X-point, in contrast to the nearly linear dependence when the ion drift is toward the X-point. This indicates that geometry plays an important role in the H-mode power threshold scaling. A simple model of the drift effect failed to predict this behaviour, but successfully predicted the sign change of gas puffing and low X-point height on the power threshold. The difference between the threshold power required for sawtooth and non-sawtooth-triggered transitions can be substantial. This effect may contribute to the observed B scaling of the H-mode power threshold.

To assess the effect of edge neutrals on the low-to-high confinement transition threshold, a broad range of plasma discharges has been analysed. From this analysis, the transition power divided by the density, at constant magnetic field, appears to be a function of a single parameter measuring the neutrals' effect. This results suggest that there is a missing parameter linked to the neutrals in the power-threshold scaling laws.

High spatially resolved measurements of the radial electric field, Er, have been made across the transition from L mode to H mode plasmas for many different plasma parameters and conditions. The evolution of the well-like structure of the Er profile formed at the L-H transition has been investigated. No distinct variation in the shape or width of the Er well at the L-H transition is observed as a function of the plasma parameters investigated, such as the plasma current, the toroidal magnetic field and the plasma density. The value of Er is negative just inside the last closed flux surface (LCFS) for all the plasmas studied. There is a variation in the depth of the Er well for different conditions. The experimental results have been compared with theoretical predictions for suppression of plasma turbulence by sheared E × B plasma flow.

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.

Recent measurements of the two-dimensional (2-D) spatial profiles of divertor plasma density, temperature, and emissivity in the DIII-D tokamak [J. Luxon et al., in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1987), p. 159] under highly radiating conditions are presented. Data are obtained using a divertor Thomson scattering system and other diagnostics optimized for measuring the high electron densities and low temperatures in these detached divertor plasmas (ne⩽1021 m−3, 0.5 eV⩽Te). D2 gas injection in the divertor increases the plasma radiation and lowers Te to less than 2 eV in most of the divertor volume. Modeling shows that this temperature is low enough to allow ion–neutral collisions, charge exchange, and volume recombination to play significant roles in reducing the plasma pressure along the magnetic separatrix by a factor of 3–5, consistent with the measurements. Absolutely calibrated vacuum ultravi...

In this paper we describe the newly installed divertor Thomson scattering system for the DIII-D tokamak and present initial results from plasma discharges. Measured plasma densities have ranged from 5 × 1018 to 5 × 1020 m−3 and divertor plasma temperatures from 1 to 500 eV. These data are compared with earlier Langmuir probe data and qualitatively compared with UEDGE computer simulations. The divertor Thomson system uses one of the eight existing core Thomson scattering lasers (1 J, 20 Hz) which has been re-directed to probe the divertor region of the DIII-D vessel. Scattered light from this multipulse Nd:Yag laser is viewed with an f/6.8 collection optics system which provides eight spatial channels from 1–21 cm above the vessel floor (divertor target), each with 1.5 cm vertical resolution. Translating the plasma across the vessel floor using position controls provides a full scan of the divertor plasma.

Electron density, temperature and parallel pressure measurements at several locations along field lines connecting the midplane scrapeoff layer (SOL) with the outer divertor are presented for both attached and partially-detached divertor cases: Ip = 1.4 MA, q95 = 4.2 and Pinput ∼ 6.7 MW under ELMing H-mode conditions. At the onset of the Partially Detached Divertor (PDD), a high density, low temperature plasma forms in the divertor SOL (divertor MARFE). The electron pressure drops by a factor of 2 between the midplane separatrix and the X-point and then an additional ∼ 3–5 times between the X-point and the outboard separatrix strike point. These results are in contrast to the attached (non-PDD) case, where electron pressure in the SOL is reduced by, at most, a factor of two between the midplane and the divertor target. Divertor MARFEs generally have only marginal adverse impact on important H-mode characteristics, such as confinement time. In fact, PDD discharges at low input power (i.e., approximately twice the L-H-mode threshold power) maintain good H-mode characteristics until a high density, low temperature plasma abruptly forms inside the separatrix near the X-point (X-point MARFE). Concurrent with the appearance of this X-point MARFE is a degradation in both energy confinement and the plasma fueling rate and an increase in the carbon impurity concentration inside the core plasma. The formation of the X-point MARFE is consistent with a thermal instability resulting from the temperature dependence of the carbon radiative cooling rate in the range ∼ 7–30 eV.

Thomson scattering measurements of ne and Te in the divertor region of a Tokamak are reported. These data are used as input to boundary physics codes such as UEDGE and DEGAS and to benchmark the predictive capabilities of these codes. These measurements have also contributed to the characterization of tokamak disruptions. A Nd:YAG laser (20 Hz, 1 J, 15 ns, 1064 nm) is directed vertically through the lower divertor region of the DIII–D Tokamak. A custom, aspherical collection lens (f/6.8) images the laser beam from 1 to 21 cm above the target plates into eight spatial channels with 1.5 cm vertical and 0.3 cm radial resolution. Two-dimensional mapping of the divertor region is achieved by sweeping the divertor X-point location radially through the fixed laser beam location. Fiber optics carry the light to polychromators whose interference filters have been optimized for low Te measurements. Silicon avalanche photodiodes measure both the scattered and plasma background light. Temperatures and densities are typically in the range of 5–200 eV and 1–10×1019 m−3, respectively. Low temperatures, Te 8×1020 m−3 have been observed in detached plasmas. Background light levels have not been a significant problem. Reduction of the laser stray light permits Rayleigh calibration. Because of access difficulties, no in-vessel vacuum alignment target could be used. Instead, an in situ laser alignment monitor provides alignment information for each laser pulse. Results are compared with Langmuir probe measurements where good agreement is found except for regions of high ne and low Te as measured by Thomson scattering.

We have obtained the first measurements of n{sub e} and T{sub e} in the DIII-D divertor region with a multi-pulse (20 Hz) Divertor Thomson Scattering (DTS) system. Eight measurement locations are distributed vertically up to 21 cm above the divertor plate. Two-dimensional distributions have been obtained by sweeping the divertor plasma across the DTS measurement location. Several operating modes have been studied, including ohmic, L-mode, Elming H-mode, and Radiative Divertor operation with puffing of D{sub 2} and impurities. Mapping of the data to either the (L{sub pol}, {phi}) or (R, Z) planes with the EFIT equilibrium is used to analyze the 2D profiles. We find that in ELMing H-mode: n{sub e}, T{sub e}, and P{sub e} are relatively constant along field lines from the X-point to the divertor plate, especially near the separatrix field line. With D{sub 2} puffing, the DTS profiles indicate that T{sub e} in a large part of divertor region below the X-point is dramatically reduced from {approximately}30-40 eV in ELMing H-mode to 1-2 eV. This results in a fairly uniform low-T{sub e} divertor, with an increased electron density in the range of 2 to 4 x 10{sup 20} m{sup -3}. Detailed comparisons of the spatial profiles of n{sub e}, T{sub e}, and electron pressure P{sub e}, are presented for several operating modes. In addition, these data are compared with initial calculations from the UEDGE fluid code.

This paper describes divertor density and temperature measurements using both a new reciprocating Langmuir probe (XPT-RCP) which plunges vertically above the divertor floor up to the X-point height and swept, single, Langmuir probes fixed horizontally across the divertor floor. These types of measurements are important for testing models of the SOL and divertor which then are used to determine engineering design criteria for plasma facing components in reactor size tokamaks. The 6 mm diameter fixed single probes (19 domed and 2 flat at radially equivalent locations) are incorporated into the lower divertor floor at 19 radial locations and swept at 250 Hz. These probes are critical for determining plasma detachment from the floor during operation with high density, highly radiating divertors. By sweeping the divertor strike point across the fixed probes, different regions of the target plate incident flux profile can be sampled and a high resolution spatial profile can be obtained from each probe tip as the strike point moves past. The X-point reciprocating probe (XPT-RCP) provides ne and Te profiles with high spatial (2 mm) and temporal (0.5 ms) resolution from the target plate to the X-point along a single vertical chord at the same radial location as a fixed probe tip at a different azimuthal location. The probe ne and Te are compared to the divertor Thomson scattering (DTS) ne and Te (eight vertical points at 20 Hz, RThomson = RX-point-rcp). Recent observations have also shown divertor densities from 3 × 1019 to 4 × 1020 m−3 near the target plate with the highest densities observed with D2 gas puffing. Electron temperature is typically of the order of 15–25 eV at the target rising to about 70 eV near the X-point. Lower temperature, higher density plasmas are observed along the inner leg. Generally good agreement among the XPT-RCP, the fixed floor probes, and the DTS is observed. Differences between these diagnostic measurements will also be discussed with respect to different operating regimes and conditions.

The next generation of tokamaks present unique challenges to plasma diagnostic design due to the physical size of the devices and the radiation environment. The need for a density measurement for density feedback control for a prototype reactor such as ITER is well established and several proposals for line average measurements have been put forward. In this article, a design for a line average density diagnostic for ITER using collinear interferometry and Faraday rotation measurements will be presented. Plasma effects on both types of measurements and density resolution will be discussed along with the possibility of combining the information from the two collinear measurements to improve the reliability and quality of the density profile. Survivability of the plasma facing mirrors, in particular the surface flatness and surface roughness, are critical issues and preliminary analysis suggests these may limit the wavelength of probing beams. Thermal and stress analysis of the plasma facing mirrors will be...

The DIII–D Thomson scattering system has been expanded to measure divertor plasma temperatures from 1 to 500 eV and densities from 0.05 to 8×1020 m−3. To complete this system, a difficult stray light problem was overcome to allow for an accurate Rayleigh scattering density calibration. The initial stray light levels were over 500 times higher than the expected Rayleigh scattered signal. Using a charge-coupled device (CCD) camera, various portions of the vessel interior were examined while the laser was fired through the vessel in air at atmospheric pressure. Image relaying, exit window tilting, entrance and exit baffle modifications, and a beam polarizer were then used to reduce the stray light to acceptable levels. The CCD camera gave prompt feedback on the effectiveness of each modification, without the need to reestablish vacuum conditions required when using the normal avalanche photodiode detectors (APD). Once the stray light was sufficiently reduced, the APD detectors provided the signal time history to more accurately identify the source location. We have also found that certain types of high reflectance dielectric coatings produce 10–15 times more scatter than other types of more conventional coatings. By using low-scatter mirror coatings and these new stray light reduction techniques, we now have more flexibility in the design of complex Thomson scattering configurations required to probe the central core and the new radiative divertor regions of the DIII–D vessel.

A dispersion interferometer based on the second-harmonic generation of a carbon dioxide laser in orientation-patterned gallium arsenide has been developed for measuring electron density in plasmas. The interferometer includes two nonlinear optical crystals placed on opposite sides of the plasma. This instrument has been used to measure electron line densities in a pulsed radio-frequency generated argon plasma. A simple phase-extraction technique based on combining measurements from two successive pulses of the plasma has been used. The noise-equivalent line density was measured to be 1.7 × 10(17) m(-2) in a detection bandwidth of 950 kHz. One of the orientation-patterned crystals produced 13 mW of peak power at the second-harmonic wavelength from a carbon dioxide laser with 13 W of peak power. Two crystals arranged sequentially produced 58 mW of peak power at the second-harmonic wavelength from a carbon dioxide laser with 37 W of peak power.

One of the systems planned for the measurement of electron density in ITER is a multi-channel tangentially viewing combined interferometer-polarimeter (TIP). This work discusses the current status of the design, including a preliminary optical table layout, calibration options, error sources, and performance projections based on a CO2/CO laser system. In the current design, two-color interferometry is carried out at 10.59 μm and 5.42 μm and a separate polarimetry measurement of the plasma induced Faraday effect, utilizing the rotating wave technique, is made at 10.59 μm. The inclusion of polarimetry provides an independent measure of the electron density and can also be used to correct the conventional two-color interferometer for fringe skips at all densities, up to and beyond the Greenwald limit. The system features five chords with independent first mirrors to reduce risks associated with deposition, erosion, etc., and a common first wall hole to minimize penetration sizes. Simulations of performance for a projected ITER baseline discharge show the diagnostic will function as well as, or better than, comparable existing systems for feedback density control. Calculations also show that finite temperature effects will be significant in ITER even for moderate temperature plasmas and can lead to a significant underestimate of electron density. A secondary role TIP will fulfill is that of a density fluctuation diagnostic; using a toroidal Alfvén eigenmode as an example, simulations show TIP will be extremely robust in this capacity and potentially able to resolve coherent mode fluctuations with perturbed densities as low as δn∕n ≈ 10(-5).

A dispersion interferometer based on second-harmonic generation in orientation-patterned GaAs has been used to measure the line density of electrons in a radio-frequency plasma with a detection limit of 7 × 1016 m−2.

Previous scaling results indicate that the H-mode power threshold increases nearly linearly with the line averaged density, {bar n}{sub e}, and the toroidal field, B{sub t}. The power threshold was measured in similar, ITER-like, discharges in JET and DIII-D, at the same {bar n}{sub e} and B{sub t} in order to determine the size scaling of the power threshold. The results indicate a size scaling proportional to the surface area, S{sup 0.5}, which is weaker than the linear surface area dependence previously assumed.

Measurements of local electron density n{sub e}, electron temperature T{sub e}, and ion temperature T{sub i} have been made at the very edge of the plasma just prior to the transition into H-mode for four different single parameter scans in the DIII-D tokamak. The means and standard derivations of n{sub e}, T{sub e}, and T{sub i} under these conditions for a value of the normalized toroidal flux of 0.98 are respectively, 1.5 {+-} 0.7 x 10{sup 19} m{sup -3}, 0.051 {+-} 0.016 keV, and 0.14 {+-} 0.03 keV. The threshold condition for the transition is more sensitive to temperature than to density. The data indicate that the dependence is not as simple as a requirement for a fixed value of the ion collisionality.

Results of experiments performed on DIII-D since the last H-mode workshop impose a number of constraints on theories. First, measurements demonstrate that both the main ion pressure gradient, , and poloidal rotation, , are important in determining in the H-mode, with the term being the trigger for the transition. Accordingly, theories must include both and effects. Since the term appears to be the trigger, the physics of the power threshold must involve the physics of the perpendicular rotation. Second, the time sequence of changes in and the density and potential fluctuations are consistent with shear causing the change in the fluctuations. Third, the sharp onset of the fluctuation suppression during this same time sequence strongly suggests a critical shear for fluctuation suppression. Fourth, Langmuir probe measurements show the need for a turbulence stabilization theory including multiple fluctuating fields and their relative phases. Finally, theories based on predicted by standard neoclassical theory disagree with experiment.

This paper presents an overview of recent experimental progress towards understanding H-mode transition physics and scaling. Terminology and techniques for studying H-mode are reviewed and discussed. The model of shear E x B flow stabilization of edge fluctuations at the L-H transition is gaining wide acceptance and is further supported by observations of edge rotation on a number of new devices. Observations of poloidal asymmetries of edge fluctuations and dephasing of density and potential fluctuations after the transition pose interesting challenges for understanding H-mode physics. Dedicated scans to determine the scaling of the power threshold have now been performed on many machines. A dear B{sub t} dependence is universally observed but dependence on the line averaged density is complicated. Other dependencies are also reported. Studies of the effect of neutrals and error fields on the power threshold are under investigation. The ITER threshold database has matured and offers guidance to the power threshold scaling issues relevant to next-step devices.

A series of scaling studies attempting to correlate the H(high)-mode power threshold (P{sub TH}) with global parameters have been conducted. Data from these discharges is also being used to look for dependence of P{sub TH} on local edge parameters and to test theories of the transition. Boronization and better operational techniques have resulted in lower power thresholds and weaker density scaling. Neon impurity injection experiments show that radiation also plays a role in determining P{sub TH}. A low density threshold for the L(low)-H(high) transition has been linked with the locked mode low density limit, and can be reduced with the use of an error field correcting coil. Highly developed edge diagnostics, with spatial resolution as low as 5 mm, are used to evaluate how the power threshold depends on local edge conditions. Preliminary analysis of local edge conditions for parameter scans of n{sub e}, B{sub T}, and I{sub p} in single-null discharges, and the X-point imbalance in double-null discharges-show that, just before the transition to H-mode, the edge temperatures near the separatrix are approximately constant at 100 < T{sub i} < 220 eV and 35 < T{sub e} < 130 eV, even though the threshold power varied from 1.5 to 14 MW. During a density scan, the edge ion collisionality, v{sub *i}, varied from 2 to 17, demonstrating that a transition condition as simple as v{sub *i} = constant is inconsistent with the data. The local edge parameters of n{sub e}, T{sub e}, and T{sub i} do not always follow the same global scaling as P{sub TH}. Therefore, theories of the L-H transition need not be constrained by these scalings.

In order to compensate for vibrations while simultaneously measuring plasma line-averaged density, the existing real-time interferometer on DIII-D uses both a HeNe and CO2 laser operating at 10.59 and 0.632 μm, respectively. The line density resolution is 1018 m−2 which is predominantly determined by the measurement electronics. In particular, errors resulting from detection of the CO2 phase shift and finite bit size in the processing electronics are the major limiting factors. This article describes a method for postshot error correction generally applicable to two-color interferometers that utilize analog quadrature phase detection. The error correction scheme exploits the facts that error introduced by the phase comparator is periodic and that measurements can be made while the density is zero and only the errors are recorded. Application of this method will reduce unphysical density oscillations which are purely a result of vibration induced phase shifts. Application to DIII-D data shows a 25% reducti...

DIII-D currently operates with a single- or double-null open divertor and graphite walls. Active particle control with a divertor cryopump has demonstrated density control, efficient helium exhaust, and reduction of the inventory of particles in the wall. Gas puffing of D{sub 2} and impurities has demonstrated reduction of the peak divertor beat flux by factors of 3--5 by radiation. A combination of active cryopumping and feedback-controlled D{sub 2} gas puffing has produced similar divertor heat flux reduction with density control. Experiments with neon puffing have shown that the radiation is equally-divided between a localized zone near the X-point and a mantle around the plasma core. The density in these experiments has also been controlled with cryopumping. These experimental results combined with modeling were used to develop the new Radiative Divertor for DIII-D. This is a double-null slot divertor with four cryopumps to provide particle control and neutral shielding for high-triangularity advanced tokamak discharges. UEDGE and DEGAS simulations, benchmarked to experimental data, have been used to optimize the design.

Thomson scattering produces ne profiles from measurement of scattered laser beam intensity. Rayleigh scattering provides a first calibration of the relation ne ∝ ITS, which depends on many factors (e.g., laser alignment and power, optics, and measurement systems). On DIII-D, the ne calibration is adjusted against an absolute ne from the density-driven cutoff of the 48 channel 2nd harmonic X-mode electron cyclotron emission system. This method has been used to calibrate Thomson ne from the edge to near the core (r/a > 0.15). Application of core electron cyclotron heating improves the quality of cutoff and depth of its penetration into the core, and also changes underlying MHD activity, minimizing crashes which confound calibration. Less fueling is needed as “ECH pump-out” generates a plasma ready to take up gas. On removal of gyrotron power, cutoff penetrates into the core as channels fall successively and smoothly into cutoff.