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The surface eroding thermocouple (SETC) has proved to be a useful fast local temperature and heat flux diagnostic since its application in DIII-D. The SETC system in DIII-D was upgraded, and seven new SETCs were installed in the small angle slot (SAS) divertor to create a localized poloidal array to provide heat flux profiles with sub-10-ms temporal resolution. In addition, a smaller SETC was developed to increase the spatial resolution and simplify installation in the SAS divertor. The SETC-containing tiles were re-engineered with a double-wide format so that the distance between the SETCs and the tile edges is extended to 10 cm on both sides, giving additional tolerance for tile-to-tile misalignments. Mineral insulated thermocouple cable with a metal sheath has also been implemented to further reduce noise and magnetic pickup. Recent measurements obtained from the SETCs indicate that the newly installed SETC array in the SAS divertor is able to provide important heat flux information to various physics experiments in closed divertors.

Negative-ion neutral-beam injection (NNBI) is an important source of heating and current drive for next-step fusion devices where the injected energy can range from hundreds of keV to 1 MeV. Few diagnostics are suitable for phase-space resolved measurements of fast ions with energy in excess of 100 keV. A study to assess the feasibility of fast-ion deuterium-alpha (FIDA) spectroscopy to diagnose high-energy ions produced by NNBI is presented. Case studies with the Large Helical Device (LHD) and JT-60SA illustrate possible solutions for the measurement. The distribution function of fast ions produced by NNBI is calculated for both devices, and the FIDA spectrum is predicted by synthetic diagnostic simulation. Results with 180 keV NNBI in LHD show that, with a judicious choice of viewing geometry, the FIDA intensity is comparable to that obtained with the existing FIDA system. The measurement is more challenging with the 500 keV NNBI in JT-60SA. Simulations predict the FIDA intensity to be about 1% of the background bremsstrahlung, which is small compared to existing FIDA implementations with positive neutral-beam injection where signal levels are an order of magnitude larger. The sampling time required to extract the small FIDA signal is determined using a probabilistic approach. Results indicate that long averaging periods, from ones to tens of seconds, are needed to resolve the FIDA signal in JT-60SA. These long averaging times are suitable in long-pulse (∼100 s), steady-state devices like JT-60SA where an important measurement objective is the spatial profile of the slowing-down distribution of fast ions.

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 Faraday-effect-based radial-interferometer-polarimeter diagnostic has been developed to explore fast magnetic dynamics in high-performance DIII-D plasmas. The instrument measures radial magnetic field perturbations using three chords positioned near the magnetic axis. Newly developed solid-state sources operating at 650 GHz provide phase noise down to 0.01°/kHz and tunable bandwidth up to 10 MHz. Various systematic errors which can contaminate the polarimetric measurement have been investigated in detail. Distortion of circular polarization due to non-ideal optical components is calibrated using a rotating quarter wave plate technique. The impact of perpendicular magnetic field, i.e., the Cotton-Mouton effect, is evaluated. The error due to non-collinearity of probe beams is minimized to less than 0.5° for electron density up to 7 × 1019 m−3 by alignment optimization. Optical feedback, due to multiple reflections induced by the double-pass configuration, is identified and reduced. Coherent and broadband high-frequency magnetic fluctuations for DIII-D H-mode plasmas are observed.A Faraday-effect-based radial-interferometer-polarimeter diagnostic has been developed to explore fast magnetic dynamics in high-performance DIII-D plasmas. The instrument measures radial magnetic field perturbations using three chords positioned near the magnetic axis. Newly developed solid-state sources operating at 650 GHz provide phase noise down to 0.01°/kHz and tunable bandwidth up to 10 MHz. Various systematic errors which can contaminate the polarimetric measurement have been investigated in detail. Distortion of circular polarization due to non-ideal optical components is calibrated using a rotating quarter wave plate technique. The impact of perpendicular magnetic field, i.e., the Cotton-Mouton effect, is evaluated. The error due to non-collinearity of probe beams is minimized to less than 0.5° for electron density up to 7 × 1019 m−3 by alignment optimization. Optical feedback, due to multiple reflections induced by the double-pass configuration, is identified and reduced. Coherent and broadband...

Final design studies in preparation for manufacturing have been performed for functional components of the vacuum portion of the ITER Low-Field Side Reflectometer (LFSR). These components consist of an antenna array, electron cyclotron heating (ECH) protection mirrors, phase calibration mirrors, and vacuum windows. Evaluation of these components was conducted at the LFSR test facility and DIII-D. The antenna array consists of six corrugated-waveguide antennas for simultaneous profile, fluctuation, and Doppler measurements. A diffraction grating, incorporated into the plasma-facing miter bend, provides protection of sensitive components from stray ECH at 170 GHz. For in situ phase calibration of the LFSR profile reflectometer, an embossed mirror is incorporated into the adjacent miter bend. Measurements of the radiated beam profile indicate that these components have a small, acceptable effect on mode conversion and beam quality. Baseline transmission characteristics of the dual-disk vacuum window are obtained and are used to guide ongoing developments. Preliminary simulations indicate that a surface-relief structure on the window surfaces can greatly improve transmission. The workability of real-time phase measurements was demonstrated on the DIII-D profile reflectometer. The new automated real-time analysis agrees well with the standard post-processing routine.

Employing an array of surface eroding thermocouples (SETCs), the heat flux measured at the outer strike point of the DIII-D Small Angle Slot (SAS) divertor was found to be reduced by ~50% when the plasma reached the detachment regime compared to the attached state. Reduction in heat flux of similar magnitude has been found in both toroidal field directions (obtained by ramping up the plasma density) but with different characteristics: With the ion B × ∇B drift direction toward the SAS divertor, the heat flux measured by the SETCs first increased as plasma density increased, which may due to reduction in energy losses in the upstream and hence increase in the power flow into the divertor, and then began to roll over concurrent with a reduction in ion saturation current density on nearby Langmuir probes. In contrast, with the ion B × ∇B drift directed away from the SAS, the heat flux began at a higher value but decreases monotonically with increasing plasma density. This leads to lower heat flux in this BT direction at up to 20% lower line-averaged density. These significant differences in the onset of detachment in opposing BT directions are believed to be largely determined by the impacts of E × B drifts. Because ITER’s divertor must be operated with some degree of plasma detachment to radiate most of the power arriving in the scrape-off-layer, it is crucial to further understand the impact of magnetic drifts on plasma detachment behavior in a tight divertor geometry.

A heterodyne detection scheme is combined with a 10.59 μm CO

Motivated by the need to measure fast equilibrium temporal dynamics, non-axisymmetric structures, and core magnetic fluctuations (coherent and broadband), a three-chord Faraday-effect polarimeter-interferometer system with fast time response and high phase resolution has recently been installed on the DIII-D tokamak. A novel detection scheme utilizing two probe beams and two detectors for each chord results in reduced phase noise and increased time response [δb ∼ 1G with up to 3 MHz bandwidth]. First measurement results were obtained during the recent DIII-D experimental campaign. Simultaneous Faraday and density measurements have been successfully demonstrated and high-frequency, up to 100 kHz, Faraday-effect perturbations have been observed. Preliminary comparisons with EFIT are used to validate diagnostic performance. Principle of the diagnostic and first experimental results is presented.

Various spatial heterodyne polarization interferometers for spectrally-resolved optical imaging of edge and core parameters in high temperature magnetized plasmas are described. Applications for such “coherence imaging” (CI) systems include imaging motional Stark effect and Zeeman effect polarimetry for determination of the magnetic field pitch angle, and passive and active (charge exchange recombination spectroscopy - CXRS) Doppler imaging of plasma temperature and flow. In this paper we describe spatial heterodyne coherence imaging systems and present first results of Doppler flow imaging in the DIII-D divertor. Instruments have been installed for imaging flows in the divertor and scrape-off-layer in the DIII-D tokamak and also for Doppler imaging on the H-1 heliac [1]. In the former case, single snapshot interferometric images of the plasma in CII 514nm, and CIII 465nm emission have been demodulated to obtain flow and ion temperature projections in both the scrape-off-layer and divertor. Flow field amplitudes in the divertor are found to be broad agreement with UEDGE modeling [2], and point the way towards experiments that address important divertor transport issues in future (© 2011 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

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.

A novel type of surface eroding thermocouple (SETC) has been tested and demonstrated in the small angle slot (SAS) divertor of DIII-D for fast local heat flux measurements. The thermojunction of the SETC is formed between two thin (10 μm) ribbons, which are filed over to create microfiber junctions. These thermocouples are able to be exposed directly to the plasma at surface temperatures exceeding 2000 °C and are capable of sub-10 ms time resolution. Before installation in SAS, the SETCs were exposed in the lower DIII-D divertor during L-mode and H-mode discharges, from which results are presented. In preliminary tests, SETCs proved to be a qualified diagnostic to accurately measure both the intra-edge localized mode (ELM) and inter-ELM heat flux during H-mode shots with high frequency ELMs (hundreds of Hz) and to resolve heat flux profiles during strike point sweeps. The heat fluxes measured by using SETCs are consistent with the heat fluxes measured by using IR cameras and Langmuir probes. These new diagnostic capabilities will complement the existing IR camera measurements and will be of particularly significant value to measure surface heat flux in the SAS divertor or other regions where the IR camera lacks line of sight.

Mirrors will be used in ITER in all optical diagnostic systems observing the plasma radiation in the ultraviolet, visible and infrared ranges. Diagnostic mirrors in ITER will suffer from electromagnetic radiation, energetic particles and neutron irradiation. Erosion due to impact of fast neutrals from plasma and deposition of plasma impurities may significantly degrade optical and polarization characteristics of mirrors influencing the overall performance of the respective diagnostics. Therefore, maintaining the best possible performance of mirrors is of the crucial importance for the ITER optical diagnostics. Mirrors in ITER divertor are expected to suffer from deposition of impurities. The dedicated experiment in a tokamak divertor was needed to address this issue. Investigations with molybdenum diagnostic mirrors were made in DIII-D divertor. Mirror samples were exposed at different temperatures in the private flux region to a series of ELMy H-mode discharges with partially detached divertor plasmas. An increase of temperature of mirrors during the exposure generally led to the mitigation of carbon deposition, primarily due to temperature-enhanced chemical erosion of carbon layers by D atoms. Finally, for the mirrors exposed at the temperature of ∼160 °C neither carbon deposition nor degradation of optical properties was detected.

The DIII-D National Fusion Facility has long been a center of innovation and development of diagnostics for magnetic fusion devices. The DIII-D device, a moderate size tokamak, with a high flexibility shaping coil set, neutral beam injection (NBI), electron cyclotron heating (ECH) and ion cyclotron heating (ICH), supports a very broad research program in fusion science, including critical aspects related to burning plasmas expected to be encountered in ITER. This scientific program is supported by a large set of diagnostics (approximately 50), which is the product of a highly collaborative program between universities, national laboratories and industry. Although many diagnostic systems are now routinely employed to measure a wide range of plasma parameters, such as temperature, rotation, density and current profiles, there are many areas that are inherently difficult or prohibitively expensive to diagnose. Such areas include the measurements associated with energetic ion populations or with the characterization of plasma flows in the divertor/edge area. In addition, the study of burning plasmas will require the development of new and updated techniques, which need to be developed and tested in existing devices in relevant plasma conditions.

High power density, phased antenna operation can often be limited by antenna voltage handling and/or impurity and density production. Using a pair of two-strap antennas for comparison, the performance of a four-strap, fast wave antenna is assessed for a variety of configurations and antenna phases in Alcator C-Mod. To obtain robust voltage handling, the antenna was reconfigured to eliminate regions where the RF E-field is parallel to B or to reduce the RF E-field to

The optimization of a tokamak towards steady state and high performance has been the focus of advanced tokamak (AT) research for the past decade. A central theme of AT research line is plasma control: control of the plasma shape; of the profiles of current, pressure, and rotation; of transport; and of MHD stability. To optimize the performance, measurements of crucial parameters such as the current density and the plasma pressure are required with appropriate spatial coverage and resolution. In addition, measurements of other parameters will be necessary to develop a fundamental understanding of the complex nonlinear interactions amongst the current density profile, the pressure profile and transport (e.g. turbulence) in high β AT plasmas. Present day experiments are providing physics insight into what a burning plasma experiment (BPX) will require as measurements. Recent research has focused on MHD stability aspects such as the neoclassical tearing mode and resistive wall mode stabilization and control of the current profile. However, in burning plasmas, new factors such as alpha particles, with their heating contribution and their relationship to transport barriers, will be increasingly important. The close relationship between measurements and active control, and the resultant impact on the requirements, will be discussed.

Research on the Alcator C-Mod tokamak [1] is focused on high particle- and power-density plasma regimes to understand particle and energy transport in the core, the dynamics of the H-mode pedestal, and scrape-off layer and divertor physics. The auxiliary heating is provided exclusively by RF waves, and both the physics and technology of RF heating and current drive are studied. The momentum which is manifested in strong toroidal rotation, in the absence of direct momentum input, has been shown to be transported in from the edge of the plasma following the L–H transition, with timescale comparable to that for energy transport. In discharges which develop internal transport barriers, the rotation slows first inside the barrier region, and then subsequently outside of the barrier foot. Heat pulse propagation studies using sawteeth indicate a very narrow region of strongly reduced energy transport, located near r/a = 0.5. Addition of on-axis ICRF heating arrests the buildup of density and impurities, leading to quasi-steady conditions. The quasi-coherent mode associated with enhanced D-Alpha (EDA) H-mode appears to be due to a resistive ballooning instability. As the pedestal pressure gradient and temperature are increased in EDA H-mode, small ELMs appear; detailed modelling indicates that these are due to intermediate n peeling–ballooning modes. Phase contrast imaging has been used to directly detect density fluctuations driven by ICRF waves in the core of the plasma, and mode conversion to an intermediate wavelength ion cyclotron wave has been observed for the first time. The bursty turbulent density fluctuations, observed to drive rapid cross-field particle transport in the edge plasma, appear to play a key role in the dynamics of the density limit. Preparations for quasi-steady-state advanced tokamak studies with lower hybrid current drive are well underway, and time dependent modelling indicates that regimes with high bootstrap fraction can be produced.

A neutral particle analyser (NPA) has been used to make measurements of the ion distribution in the Alcator C-Mod tokamak. We have used the analyser to measure the energy distribution of majority deuterons and minority hydrogen (up to 20 keV) at 0.4≤r/a≤0.6, during minority ion cyclotron range of frequencies (ICRF) heating. These energy spectra were also simulated by using the predicted minority ion distribution from a combined Fokker Planck ICRF code in a model for the neutral particle flux. The model predictions for neutral flux were found to be in qualitative agreement with the measured energy spectra. The energy spectra for two different ICRF antenna configurations were also measured and found to be very similar.

A series of experiments are described which are aimed at quantifying the relative contribution of divertor leakage and radial ion transport on neutral pressures surrounding the core plasma. Evidence is presented implying that cross-field transport competes with, or dominates, parallel transport in such a way that plasma exists far out in the scrape-off layer (SOL) in the shadow of limiters and recycles on main chamber surfaces. The transition from L- to H-mode core confinement does not affect the far SOL characteristics nor the scaling of pressures around the plasma. Variations in magnetic equilibrium are used to vary the divertor pressures independent of the core plasma. Based on the analysis of such experiments using a simple neutral flow model we estimate that neutrals escaping from leaks in the lower (closed) divertor during lower x-point operation contribute a smaller fraction (~10-30%) of the midplane pressure than main chamber recycling. The inferred leakage is much larger from the upper (open) divertor during upper x-point operation. Most neutrals escaping from either divertor do not directly travel to the midplane. Instead, they are redirected, most likely by some combination of ionization and/or collisions (elastic, charge exchange).

Double transport barrier modes (simultaneous core and edge transport barrier) have been observed with off-axis ion cyclotron range of frequencies (ICRF) heating in the Alcator C-Mod tokamak [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)]. An internal transport barrier (ITB) is routinely produced in enhanced DαH-mode (EDA) discharges where the minority ion cyclotron resonance layer is at r/a∼|0.5| during the current flat top phase of the discharge. The density profile becomes peaked without the presence of a particle source in the plasma core and continues to peak until the increased core impurity radiation arrests the improved energy confinement, ultimately leading to a barrier collapse. With the addition of moderate (0.6 MW) central ICRF heating, the double barrier mode was maintained for as long as the ICRF power was applied. Modeling shows that the internal thermal barrier was maintained throughout the discharge. The presence of sawteeth throughout most of the ITB discharge allows sawtooth-indu...

Double transport barrier plasmas comprised of an edge enhanced D? (EDA) H-mode pedestal and an internal transport barrier (ITB) have been observed in Alcator C-Mod. The ITB can be routinely produced in ICRF heated plasmas by locating the wave resonance off-axis near |r/a|~0.5, provided the target plasma average density is above ~1.4?1020?m-3, and can develop spontaneously in some Ohmic H-mode discharges. The formation of the barrier appears in conjunction with a decrease or reversal in the central (impurity) toroidal rotation velocity. The ITB foot is located near r/a~0.5, regardless of how the barrier was produced. The ITBs can persist for ~15 energy confinement times (?E), but exhibit a continuous increase of the central electron density, up to values near 1?1021?m-3 (in the absence of an internal particle source), followed by collapse of the barrier. Application of additional on-axis ICRF heating arrests the density and impurity peaking, which occurs along with an increase (co-current) in the core rotation velocity. Quasi-steady state double barrier plasmas have been maintained for 10?E or longer, with a bootstrap fraction of 0.13 near the ITB foot.

A population of suprathermal ions has been observed in the scrape off layer during injection of power in the ion cyclotron frequency range on the Alcator C-Mod tokamak [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)]. These ions generated energetic neutrals which were measured with a charge exchange neutral particle analyzer. In the hydrogen minority heating regime, fast particles were generated only when the injected power was above 500 kW. Fast particles were observed at a low level of injected power when the rf was a multiple of the ion cyclotron frequency in the scrape off layer. The power deposited in the edge was always less than five percent of the injected power and did not increase impurity levels. The theoretical threshold of parametric decay instabilities for decay of the pump wave in the antenna near field into an ion Bernstein wave and an ion cyclotron quasimode matches the observed threshold for fast ion production under a variety of conditions.

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.

The formation of internal transport barriers (ITBs) has been observed in the core region of Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] under a variety of conditions. The improvement in core confinement following pellet injection (pellet enhanced performance or PEP mode) has been well documented on Alcator C-Mod in the past. Recently three new ITB phenomena have been observed which require no externally applied particle or momentum input. Short lived ITBs form spontaneously following the high confinement to low confinement mode transition and are characterized by a large increase in the global neutron production (enhanced neutron or EN modes). Experiments with ion cyclotron range of frequencies power injection to the plasma off-axis on the high field side results in the central density rising abruptly and becoming peaked. The ITB formed at this time lasts for ten energy confinement times. The central toroidal rotation velocity decreases and changes sign as the density rises. Simi...

Cross-field particle transport in the scrape-off layer (SOL) of Alcator C-Mod [Phys. Plasmas 1, 1511 (1994)] can be characterized by an effective particle diffusivity (Deff) that increases markedly with distance from the separatrix. As a consequence, recycling onto the main-chamber walls is large compared to plasma flows into the divertor volume. The SOL exhibits a two-layer structure: Steep gradients and moderate fluctuation levels are typically found in a ∼5 mm region near the separatrix (near SOL) where parallel electron conduction typically dominates energy losses. Small gradients and larger fluctuation levels with longer correlation times are found outside this region (far SOL). Deff in the near SOL increases strongly with local plasma collisionality normalized to the magnetic connection length. As the discharge density limit is approached, Deff and associated fluctuation levels become large across the entire SOL and cross-field heat convection everywhere exceeds parallel conduction losses, impacting the power balance of the discharge.

High resolution measurements on the Alcator C-Mod tokamak [I. H. Hutchinson et al., Phys. Plasmas 1, 1551 (1994)] of the transport barrier in the “Enhanced Dα” (EDA) regime, which has increased particle transport without large edge localized modes, show steep density and temperature gradients over a region of 2–5 mm, with peak pressure gradients up to 12 MPa/m. Evolution of the pedestal at the L-H transition is consistent with a large, rapid drop in thermal conductivity across the barrier. A quasi-coherent fluctuation in density, potential, and Bpol, with f0∼50–150 kHz and kθ∼4 cm−1, always appears in the barrier during EDA, and drives a large particle flux. Conditions to access the steady-state EDA regime in deuterium include δ>0.35, q95>3.5, and L-mode target density ne>1.2×1020 m−3. A reduced q95 limit is found for hydrogen discharges.

Co-current central impurity toroidal rotation has been observed in Alcator C-Mod plasmas with on-axis ICRF heating. The rotation velocity increases with plasma stored energy and decreases with plasma current. Very similar behaviour has been seen during ohmic H modes, which suggests that the rotation, generated in the absence of an external momentum source, is not mainly an ICRF effect. A scan of ICRF resonance location across the plasma has been performed in order to investigate possible influences on the toroidal rotation. With a slight reduction of toroidal magnetic field from 4.7 to 4.5 T and a corresponding shift of the ICRF resonance from r/a = -0.36 to -0.48, the central toroidal rotation significantly decreased together with the formation of an internal transport barrier (ITB). During the ITB phase, electrons and impurities peaked continuously for |r/a| ≤ 0.5. Comparison of the observed rotation and neoclassical predictions indicates that the core radial electric field changes from positive to negative during the ITB phase. Similar rotation suppression and ITB formation have been observed during some ohmic H mode discharges.

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...

A high spatial resolution CCD based one-dimensional imaging system to measure visible continuum emissivity profiles from Alcator C-Mod tokamak plasmas is described. The instrument has chordal resolution that is better than 1 mm for the edge region of the plasma, where very sharp (1 to 10 mm) gradient lengths in plasma parameters are observed after the formation of the H-mode transport barrier. Each image has up to 2048 pixels, and total spatial coverage goes from 2 cm inside of the magnetic axis to ∼4 cm outside of the last closed flux surface in the ∼22 cm horizontal minor radius plasmas. Time resolution can be varied from 0.21 ms to 4 ms; good signal to noise is achieved with 1 ms integration under typical plasma conditions. The emission over most of the plasma volume is dominated by free–free bremsstrahlung, and can be used to infer local values of the average ion charge (Zeff). Toroidally localized puffing of deuterium, nitrogen, and helium reveals that a significant contribution to the signal in the scrape-off layer at the extreme edge of the plasma can come from diatomic molecular band pseudocontinuum emission.

Two new high resolution detectors have been installed on the Alcator C-Mod tokamak to measure the neutral density and ionization rate at the edge of the main chamber plasma. Using a silicon detector sensitive to UV light, and a very narrow filter with transmission peaking at 1216 A, the Lyman alpha radiation emanating from neutral deuterium (and hydrogen) is measured. The detectors consist of 20 channel arrays which view the plasma tangentially 12.5 cm below the outer midplane, and 10 cm above the inner midplane. The imaging is performed using a 1 mm×3 mm slit, which gives a nominal radial resolution of 2 and 3 mm, respectively. The local emissivity is obtained via a standard Abel inversion technique. Employing well-known branching ratios, and using measured local electron density and temperature, the neutral density and ionization rate are inferred with similar radial resolution. Details of the setup and sensitivity of the results to plasma conditions are discussed.

In this paper we summarize the results of recent measurements of the high confinement mode (H-mode) edge transport barrier in Alcator C-Mod. A number of new, high-resolution edge diagnostics provide measurements of plasma parameters (electron density and temperature, and impurity density) in the pedestal region. Pedestal parameters are measured both at the top edge of the plasma and at the midplane. Very narrow (2-5 mm) electron density and temperature and impurity density pedestals are observed. The impurity pedestal is located inside the density and temperature pedestals in the midplane measurements, but coincides with the ne pedestal at the top edge. Correlation of both width and height of the pedestals at the midplane with global plasma parameters and core confinement characteristics is observed. At the same time there are no clear indications of scaling for the electron density and impurity pedestal widths measured at the top of the plasma.

Recent breakthroughs in silicon detector technology now permit measurement of radiated power over a wide range in photon energies. These detectors (also known as AXUV photodiodes) have a flat spectral power response from ultraviolet to x-ray energies, and with a slightly reduced efficiency all the way down to visible wavelengths. Since they can be made small, multichannel detectors allow high spatial resolution to be combined with an intrinsic high temporal resolution, which can reach the microsecond range, depending on the application. Additional features include ease of use and installation, and relatively low cost compared to other techniques. A combination of two multichannel toroidally viewing systems has been recently installed on the Alcator C-Mod tokamak. The first array, which is composed of 16 channels, sees tangentially the outer-half of the plasma at the midplane, and is used to measure the total power radiated. The second array, also located at the midplane, consists of 19 channels and views ...

A series of experiments, examining the confinement properties of ion cyclotron range of frequencies (ICRF) heated H mode plasmas, has been carried out on the Alcator C-Mod tokamak. Alcator C-Mod is a compact tokamak that operates at high particle, power and current densities at toroidal fields up to 8 T. Under these conditions the plasma is essentially thermal with very little contribution to the stored energy from energetic ions (typically no more than 5%) and with Ti~Te. Most of the data were taken with the machine in a single null `closed' divertor configuration with the plasma facing components clad in molybdenum tiles. The data include those taken both before and after the first wall surfaces were coated with boron, with emphasis on the latter. H modes obtained from plasmas run on boronized walls typically had a lower impurity content and radiated power and attained a higher stored energy than those run on bare molybdenum. Confinement enhancement, the energy confinement time normalized to L mode scaling, for discharges with boronized walls, ranged from 1.6 to 2.4. The unique operating regime of the Alcator C-Mod device provided a means for extending the tests of global scaling laws to parameter ranges not previously accessible. For example, the Alcator C-Mod edge localized mode (ELM)-free data were found to be 1.1 to 1.6 times the ITERH93 scaling and the ELMy data almost 2.0 to 2.8 times the ITERH92 ELMy scaling law, suggesting that the size scaling in both scalings may be too strong. While both ELM-free and ELMy discharges were produced, the ELM characteristics were not easily compared with observations on other devices. No large, low frequency ELMs were seen despite the very high edge pressure and temperature gradients that were attained. For all of our H mode discharges, a clear linear relationship between the edge temperature pedestal and the temperature gradient in the core plasma was observed; the discharges with the `best' transport barriers also showing the greatest improvement in core c

Highly localized direct electron heating [full width at half-maximum (FWHM)≅0.2a] via mode converted ion Bernstein waves has been observed in the Alcator C-Mod Tokamak [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)]. Electron heating at or near the plasma center (r/a⩾0.3) has been observed in H(3He) discharges at B0=(6.0–6.5) T and ne(0)≅1.8×1020 m−3. [Here, the minority ion species is indicated parenthetically.] Off-axis heating (r/a⩾0.5) has also been observed in D(3He) plasmas at B0=7.9 T. The concentration of 3He in these experiments was in the range of n3He/ne≅(0.2–0.3) and the locations of the mode conversion layer and electron heating peak could be controlled by changing the 3He concentration or toroidal magnetic field (B0). The electron heating profiles were deduced using a rf modulation technique. Detailed comparisons with one-dimensional and toroidal full-wave models in the ion cyclotron range of frequencies have been carried out. One-dimensional full-wave code predictions were found to ...

Enhanced confinement modes up to a toroidal field of BT=8 T have been studied with up to 3.5 MW of radiofrequency (rf) heating power in the ion cyclotron range of frequencies (ICRF) at 80 MHz. H-mode is observed when the edge temperature exceeds a threshold value. The high confinement mode (H-mode) with higher confinement enhancement factors (H) and longer duration became possible after boronization by reducing the radiated power from the main plasma. A quasi-steady state with high confinement (H=2.0), high normalized beta (βN=1.5), low radiated power fraction (Pradmain/Ploss=0.3), and low effective charge (Zeff=1.5) has been obtained in Enhanced Dα H-mode. This type of H-mode has enhanced levels of continuous Dα emission and very little or no edge localized mode (ELM) activity, and reduced core particle confinement time relative to ELM-free H-mode. The pellet enhanced performance (PEP) mode is obtained by combining core fueling with pellet injection and core heating. A highly peaked pressure profile with a central value of 8 atmospheres was observed. The steep pressure gradient drives off-axis bootstrap current, resulting in a shear reversed safety factor (q) profile. Suppression of sawteeth appears to be important in maintaining the highly peaked pressure profile. Lithium pellets were found to be more effective than deuterium pellets in raising q0.

A neutral particle analyzer, based on the time-of-flight technique is scheduled to be installed on the Alcator C-Mod tokamak. The instrument was originally designed and used in the ZT-40 experiment, and later, briefly used on the CTX experiment at Los Alamos National Laboratory. The design consists of a chopper wheel mounted on a turbomolecular pump, a ∼2 m long flight tube, a Cu–Be secondary electron emitting surface, and an electron multiplier, which together are used as the neutral particle detector. Among the changes introduced in the system from the original design, of particular interest is the location of the instrument. The instrument is mounted behind a magnetic-type neutral particle analyzer and shares the same line of sight. Both of them can be scanned, poloidally down to the X point and tangentially to R/R0=0.7. This will enable us to compare neutral flux through two different techniques with an overlap in energies from ∼0.5–4 keV, with the instrument capable of detecting neutral particles wit...

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.

Results of ICRF heating experiments in Alcator C-Mod during the November 1994 to June 1995 campaign are summarized. Efficient heating of high-density plasmas was demonstrated with high power densities (up to volume averaged, surface averaged). These experiments were carried out with RF powers up to 3.5 MW at 80 MHz, at magnetic fields up to 8 T and plasma currents up to 1.2 MA. For on-axis hydrogen minority heating at 5.3 T, near complete absorption is achieved. Energy confinement in L-mode plasmas was found to be consistent with the ITER89-P scaling. H-mode is routinely observed when the ion drift is directed toward the X-point. The H-mode power threshold was found to scale as low as , which is a factor of two lower than the scaling observed on other tokamaks. PEP modes with highly peaked density and ion temperature profiles and highly enhanced fusion reactivity were obtained with Li pellet injection followed by on-axis ICRF heating at both (H minority heating) and 8 T ( minority heating). In H - plasmas at T highly localized direct electron heating by the mode converted ion Bernstein wave was observed. Nearly complete absorption by electrons in a small volume resulted in an extremely high electron heating power density of with .

The sign of the toroidal magnetic field has a major effect on the divertor in Alcator C-Mod, determining whether the higher recycling is on the inboard or outboard side and leading to inboard/outboard temperature and density differences of up to a factor of ten. A recently published paper (Hutchinson et al 1995 Plasma Phys. Control. Fusion 37 1389) reported in detail on these observations, which are an indication of the importance of plasma particle drifts for understanding and modelling the divertor. The conference presentation covered these published results and also some additional results, documented here, in which the plasma flows associated with the field-direction-dependent asymmetries are measured using a Mach probe and a localized impurity puffing plume analysis technique.

The regime of high particle and energy confinement known as the H mode [Phys. Rev. Lett. 49, 1408 (1982)] has been extended to a unique range of operation for divertor tokamaks up to toroidal fields of nearly 8 T, line‐averaged electron densities of 3×1020 m−3, and surface power densities of nearly 0.6 MW/m2 in the compact high‐field tokamak Alcator C Mod [Phys. Plasmas 1, 1511 (1994)]. H modes are achieved in Alcator C Mod with Ion Cyclotron Resonant Frequency (ICRF) heating and with Ohmic heating alone without boronization of the all molybdenum tiled first wall. Large increases in charge exchange flux are observed during the H mode over the entire range of energies from 2 to 10 keV. There appears to be an upper limit to the midplane neutral pressure, of about 0.08 Pa above which no H modes have been observed. The plasmas with the best energy confinement have the lowest midplane neutral pressures, below 0.01 Pa. There is an edge electron temperature threshold such that Te≥280 eV ±40 eV for sustaining the...

The divertor of the Alcator C‐Mod tokamak [Phys. Plasmas 1, 1511 (1994)] routinely radiates a large fraction of the power entering the scrape‐off layer. This dissipative divertor operation occurs whether the divertor is detached or not, and large volumetric radiative emissivities, up to 60 MW m−3 in ion cyclotron range of frequency (ICRF) heated discharges, have been measured using bolometer arrays. An analysis of both Ohmic and ICRF‐heated discharges has demonstrated some of the relative merits of detached divertor operation versus high‐recycling divertor operation. An advantage of detached divertor operation is that the power flux to the divertor plates is decreased even further than its already low value. Some disadvantages are that volumetric losses outside the separatrix in the divertor region are decreased, the neutral compression ratio is decreased, and the penetration efficiency of impurities increases.

The quality of plasma produced in a magnetic confinement fusion device is influenced to a large extent by the neutral gas surrounding the plasma. The plasma is fueled by the ionization of neutrals, and charge exchange interactions between edge neutrals and plasma ions are a sink of energy and momentum. Here we describe a diagnostic capable of measuring the spatial distribution of neutral gas in a magnetically confined fusion plasma. A high intensity (5 MW/cm(2)), narrow bandwidth (0.1 cm(-1)) laser is injected into a hydrogen plasma to excite the Lyman β transition via the simultaneous absorption of two 205 nm photons. The absorption rate, determined by measurement of subsequent Balmer α emission, is proportional to the number of particles with a given velocity. Calibration is performed in situ by filling the chamber to a known pressure of neutral krypton and exciting a transition close in wavelength to that used in hydrogen. We present details of the calibration procedure, including a technique for identifying saturation broadening, measurements of the neutral density profile in a hydrogen helicon plasma, and discuss the application of the diagnostic to plasmas in the DIII-D tokamak.

A series of transport experiments has been carried out in Alcator‐C‐Mod. [Phys Plasmas 1, 1511 (1994)]. Data from both Ohmic and ICRF (ion cyclotron range of frequencies) heated plasmas can be fitted with an L‐mode (low mode) scaling law. The Ohmic τE’s show no scaling with density in any regime and can reach values of 2–3 times neo‐Alcator. Impurity confinement has been studied with the laser blow‐off technique with τI showing nearly linear scaling with plasma current. Ohmic and ICRF H modes are obtained over a wide range of discharge parameters, extending the range in the international database for nB, by almost a factor of 10. The power threshold for ELM‐free (edge localized mode) discharges is in rough agreement with the scaling P/S=0.044nB. Energy diffusivities of Ohmic and ICRF heated plasmas have been measured from local analysis of plasma profiles and power fluxes. The same analysis produces a value for plasma resistivity which lies between the Spitzer and neoclassical calculations. Analysis of pl...

Detailed measurements and transport analysis of divertor conditions in Alcator C‐Mod [Phys. Plasmas 1, 1511 (1994)] are presented for a range of line‐averaged densities, 0.7

A design to extend the unique divertor Thomson scattering system on DIII-D to allow measurements of electron temperature and density in high triangularity plasmas is presented. Access to this region is selectable on a shot-by-shot basis by redirecting the laser beam of the existing divertor Thomson system inboard - beneath the lower floor using a moveable, high-damage threshold, in-vacuum mirror - and then redirecting again vertically. The currently measured divertor region remains available with this mirror retracted. Scattered light is collected from viewchords near the divertor floor using in-vacuum, high temperature optical elements and relayed through the port window, before being coupled into optical fiber bundles. At higher elevations from the floor, measurements are made by dynamically re-focusing the existing divertor system collection optics. Nd:YAG laser timing, analysis of the scattered light spectrum via polychromators, data acquisition, and calibration are all handled by existing systems or methods of the current multi-pulse Thomson scattering system. Existing filtered polychromators with 7 spectral channels are employed to provide maximum measurement breadth (Te in the range of 0.5 eV-2 keV, ne in the range of 5 × 1018-1 × 1021 m3) for both low Te in detachment and high Te measurement up beyond the separatrix.