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Recent research on the Alcator C-Mod tokamak has focused on a range of scientific issues with particular emphasis on ITER needs and on detailed comparisons between experimental measurements and predictive models. Research on ICRF (ion cyclotron range of frequencies) heating emphasized the origins and mitigation of metallic impurities while work on lower hybrid current drive experiments have focused on linear and nonlinear wave interactions that limit efficiency at high densities in regimes with low single pass absorption. Experiments in core turbulence and transport focused on quantitative, multi-field comparisons between nonlinear gyro-kinetics simulations and experimental measurements of profiles, fluxes and fluctuations. Experiments into self-generated rotation observed spontaneous flow reversal at a critical density identical to the transition density between linear ohmic confinement and saturated ohmic confinement regimes. H-mode studies have measured pedestal widths consistent with kinetic-ballooning-mode-like instabilities, while the pedestal heights quantitatively match the EPED code predictions. Experiments with I-mode have increased the operating window for this promising edge-localized-mode-free regime. Extrapolation of I-mode to ITER suggests that the fusion gain Q ~ 10 could be possible in ITER. Investigations into the physics and scaling of the power exhaust channel width in attached enhanced D-alpha H-mode and L-mode plasma showed a direct connection between the midplane pressure-folding length and the outer divertor target footprint. The width was found to scale inversely with IP, while being independent of conducted power, BT or q95 and insensitive to the scrape-off layer connection length - a behaviour that suggests critical-gradient physics sets both pressure and heat-flux profiles.

A dedicated experiment during simultaneous lower hybrid (LH) and ion cyclotron range-of-frequencies (ICRF) operations is carried out to evaluate and understand the effects of ICRF power on the scrape-off-layer (SOL) density profiles and on the resultant LH coupling for a wide range of plasma parameters on Alcator C-Mod. Operation of the LH launcher with the adjacent ICRF antenna significantly degrades LH coupling while operation with the ICRF antenna that is not magnetically connected to the LH launcher minimally affects LH coupling. An X-mode reflectometer system at three poloidal locations adjacent to the LH launcher and a visible video camera imaging the LH launcher are used to measure local SOL density profile and emissivity modifications with the application of LH and LH + ICRF power. These measurements confirm that the density in front of the LH launcher depends strongly on the magnetic field line mapping of the active ICRF antenna. Reflectometer measurements also observe both ICRF-driven and LH-driven poloidal density profile asymmetries, especially a strong density depletion at certain poloidal locations in front of the LH launcher during operation with a magnetically connected ICRF antenna. The results indicate that understanding both LH-driven flows and ICRF sheath driven flows may be necessary to understand the observed density profile modifications and LH coupling results during simultaneous LH + ICRF operation.

This paper describes 2D imaging measurements of plasma turbulence made in the scrape-off layer of the Alcator C-Mod tokamak simultaneously at two different poloidal locations, one near the outer midplane and the other near the divertor X-point region. These images were made with radial and poloidal resolution using two gas puff imaging diagnostics not directly connected along a B field line. The turbulence correlation structure has a significantly different tilt angle with respect to the local flux surfaces for the midplane and X-regions, and a slightly different ellipticity and size. The time-averaged turbulence velocities can be different in the midplane and X-regions, even within the same flux surface in the same shot. The structures are partially consistent with a magnetic flux tube mapping model, and the velocities are compared with various models for turbulence flow.

The edge and scrape-off layer (SOL) region of a tokamak plasma is considered, with emphasis on sheared flow generation and the dynamics of blob-filaments. Both numerical simulations and experimental data analysis are employed. The simulations use the fluid-based two-dimensional (2D) curvature-interchange model embedded in the SOLT code. A blob-tracking algorithm based on 2D time-resolved images from the gas puff imaging diagnostic has also been developed and applied to NSTX, Alcator C-Mod and simulation data. The algorithm is able to track the blob motion and changes in blob structure, such as elliptical deformations, that can be affected by sheared flows. Results of seeded blob simulations and quasi-steady turbulence simulations are compared with the experimental data to determine the role of plasma parameters on the blob tracks and to evaluate the exchange of momentum between the blobs and flows. The simulations are shown to reproduce many qualitative and quantitative features of the data including size, scale-length and direction of perpendicular (approximately poloidal) flows, the inferred Reynolds acceleration and residual stress, poloidal reversal of blob tracks, and blob trapping and/or ejection. Mechanisms related to blob motion, SOL currents and radial inhomogeneity are shown to be sufficient to explain the presence or absence of mean and oscillating zonal sheared flows in selected shots.

Measurements of poloidal variation, nz/⟨nz⟩, in high-Z impurity density have been made using photodiode arrays sensitive to vacuum ultraviolet and soft x-ray emission in Alcator C-Mod plasmas. In/out asymmetries in the range of −0.2 0, of a flux surface is found to be well described by a combination of centrifugal, poloidal electric field, and ion-impurity friction effects. Up/down asymmetries, −0.05 0 corresponding to accumulation opposite the ion ∇B drift direction. Measurements of the up/down asymmetry of molybdenum are found to disagree with predictions from recent neoclassical theory in the trace limit, nzZ2/ni≪1. Non-trace levels of impurities are expected to modify the main-ion poloidal flow and thus change friction-driven impurity density asymmetries and impurity poloidal rotation, vθ,z. Artificially modifying main...

Ion cyclotron range of frequency (ICRF) heating is expected to provide auxiliary heating for ITER and future fusion reactors where high Z metallic plasma facing components (PFCs) are being considered. Impurity contamination linked to ICRF antenna operation remains a major challenge particularly for devices with high Z metallic PFCs. Here, we report on an experimental investigation to test whether a field aligned (FA) antenna can reduce impurity contamination and impurity sources. We compare the modification of the scrape of layer (SOL) plasma potential of the FA antenna to a conventional, toroidally aligned (TA) antenna, in order to explore the underlying physics governing impurity contamination linked to ICRF heating. The FA antenna is a 4-strap ICRF antenna where the current straps and antenna enclosure sides are perpendicular to the total magnetic field while the Faraday screen rods are parallel to the total magnetic field. In principle, alignment with respect to the total magnetic field minimizes integrated E|| (electric field along a magnetic field line) via symmetry. A finite element method RF antenna model coupled to a cold plasma model verifies that the integrated E|| should be reduced for all antenna phases. Monopole phasing in particular is expected to have the lowest integrated E||. Consistent with expectations, we observed that the impurity contamination and impurity source at the FA antenna are reduced compared to the TA antenna. In both L and H-mode discharges, the radiated power is 20%–30% lower for a FA-antenna heated discharge than a discharge heated with the TA-antennas. However, inconsistent with expectations, we observe RF induced plasma potentials (via gas-puff imaging and emissive probes to be nearly identical for FA and TA antennas when operated in dipole phasing). Moreover, the highest levels of RF-induced plasma potentials are observed using monopole phasing with the FA antenna. Thus, while impurity contamination and sources are indeed reduced with the FA antenna configuration, the mechanism determining the SOL plasma potential in the presence of ICRF and its impact on impurity contamination and sources remains to be understood.

New experimental data from the Alcator C-Mod tokamak are used to benchmark predictive modelling of the edge pedestal in various high-confinement regimes, contributing to greater confidence in projection of pedestal height and width in ITER and reactors. ELMy H-modes operate near stability limits for ideal peeling–ballooning modes, as shown by calculations with the ELITE code. Experimental pedestal width in ELMy H-mode scales as the square root of βpol at the pedestal top, i.e. the dependence expected from theory if kinetic ballooning modes (KBMs) were responsible for limiting the pedestal width. A search for KBMs in experiment has revealed a short-wavelength electromagnetic fluctuation in the pedestal that is a candidate driver for inter-edge localized mode (ELM) pedestal regulation. A predictive pedestal model (EPED) has been tested on an extended set of ELMy H-modes from C-Mod, reproducing pedestal height and width reasonably well across the data set, and extending the tested range of EPED to the highest absolute pressures available on any existing tokamak and to within a factor of three of the pedestal pressure targeted for ITER. In addition, C-Mod offers access to two regimes, enhanced D-alpha (EDA) H-mode and I-mode, that have high pedestals, but in which large ELM activity is naturally suppressed and, instead, particle and impurity transport are regulated continuously. Pedestals of EDA H-mode and I-mode discharges are found to be ideal magnetohydrodynamic (MHD) stable with ELITE, consistent with the general absence of ELM activity. Invocation of alternative physics mechanisms may be required to make EPED-like predictions of pedestals in these kinds of intrinsically ELM-suppressed regimes, which would be very beneficial to operation in burning plasma devices.

We demonstrate that existing theories are insufficient to explain up/down asymmetries of argon x-ray emission in Alcator C-Mod ohmic plasmas. Instead of the poloidal variation, ñ[subscript z]/〈n[subscript z]〉, being of order the inverse aspect ratio, ϵ, and scaling linearly with B[subscript t][superscript _ over n][subscript e]/I[2 over p], it is observed over 0.8 < r/a < 1.0 to be of order unity and exhibits a threshold behaviour between 3.5, United States. Dept. of Energy (Contract DE-FC02-99ER54512), United States. Dept. of Energy (Fusion Research Postdoctoral Research Program)

A swept-frequency X-mode reflectometer has been used to measure the scrape-off-layer (SOL) density profiles with and without lower hybrid (LH) power at three poloidal locations adjacent to the LH launcher for various plasma parameters in order to understand the coupling of LH waves on Alcator C-Mod. LH power has been observed to create significant poloidal SOL density profile asymmetries that are correlated with visible video camera images of emissivity patterns in front of the LH launcher. The observed density profile asymmetries depend on LH power, , magnetic geometry and magnetic field direction. A 2D diffusive?convective model has been used to show that these density profile modifications are consistent with a LH vortex, where LH power drives E???B drifts that then modify the SOL density profile. In particular, the simulations show that the density profile can possibly create a net poloidally averaged density depletion in front of the waveguide rows. A LH slab coupling model is then used to show that the simulated reflection coefficients strongly depend on the poloidal density profile asymmetries. The simulated LH power reflection coefficients agree with the experimental reflection coefficients only after the observed density depletion is included in the model.

ICRF heating experiments have been carried out in the Alcator C‐Mod tokamak at power levels up to 3.5 MW. Features of Alcator C‐Mod include high density operation, molybdenum plasma facing components, and a closed divertor configuration. The heating is accomplished with two two‐strap antennas each run with dipole phasing at 80 MHz in deuterium plasmas with a hydrogen minority resonant at 5.3 T. Plasmas with Ti=4 keV and Te=5 keV at ne=1×1020 m−3 have been produced with 3.5 MW of heating power. The heating has been shown to be strongest with a low minority concentration (1–5%) and the resonance on axis. For ne≳2×1020 m−3, the central Zeff remains below 1.5 with more than 2.5 MW of applied rf power. L‐mode confinement scaling is observed both for ohmic and ICRF‐heated plasmas with some deterioration at higher densities. H‐mode transitions have been produced with the threshold for ELM‐free H‐modes at or below the ASDEX/DIII‐D scaling. Enhanced confinement with strongly peaked density profiles has been achi...

A peak fusion power production of 9.3±0.7 MW has been achieved on the Tokamak Fusion Test Reactor (TFTR) in deuterium plasmas heated by co and counter injected deuterium and tritium neutral beams with a total power of 33.7 MW. The ratio of fusion power output to heating power input is 0.27. At the time of the highest neutron flux the plasma conditions are: Te(0)=11.5 keV, Ti(0)=44 keV, ne(0)=8.5×1019 m−3, and 〈Zeff〉=2.2 giving τE=0.24 s. These conditions are similar to those found in the highest confinement deuterium plasmas. The measured D‐T neutron yield is within 7% of computer code estimates based on profile measurements and within experimental uncertainties. These plasmas have an inferred central fusion alpha fraction of 0.2% and central fusion power density of 2 MW/m3 similar to that expected in a fusion reactor. Even though the alpha velocity exceeds the Alfven velocity throughout the time of high neutron output in most high power plasmas, MHD activity is not substantially different from that in co...

Reduced-model scrape-off layer turbulence (SOLT) simulations of an enhanced D-alpha (EDA) H-mode shot observed in the Alcator C-Mod tokamak were conducted to compare with observed variations in the scrape-off-layer (SOL) width of the parallel heat flux profile. In particular, the role of the competition between sheath- and conduction-limited parallel heat fluxes in determining that width was studied for the turbulent SOL plasma that emerged from the simulations. The SOL width decreases with increasing input power and with increasing separatrix temperature in both the experiment and the simulation, consistent with the strong temperature dependence of the parallel heat flux in balance with the perpendicular transport by turbulence and blobs. The particularly strong temperature dependence observed in the case analyzed is attributed to the fact that these simulations produce SOL plasmas which are in the conduction-limited regime for the parallel heat flux. A persistent quasi-coherent (QC) mode dominates the SOLT simulations and bears considerable resemblance to the QC mode observed in C-Mod EDA operation. The SOLT QC mode consists of nonlinearly saturated wave-fronts located just inside the separatrix that are convected poloidally by the mean flow, continuously transporting particles and energy and intermittently emitting blobs into the SOL.

H-mode power thresholds (Pth) are examined in Alcator C-Mod lower single null plasmas with standard vertical-plate and slot-type divertor configuration (BT = 5.4 T, Ip = 0.9 MA, q95 ∼ 4.0 and 〈ne〉 = (1.3–1.6) × 1020 m−3). The required power to access H-mode is found to be significantly reduced when the plasma is operated with a slot divertor. The lowest Pth achieved is ∼0.7 MW, about 1/3 of the normal level and 40% of the multi-machine scaling law prediction. Pth does not explicitly correlate with plasma shaping parameters (X-point locations, elongation, triangularities, etc), but somehow shows an overall inverse correlation with outer divertor leg length (the poloidal distance between X-point and outer strike point) and/or the outer divertor connection length. This implies that another, still unknown parameter hidden in the vicinity of the separatrix outer leg might be responsible for the observed difference in H-mode access power. Despite the large differences in Pth, the edge electron temperature (Te) and density (ne) profiles prior to L–H transition are not strongly affected by the divertor configuration. Values of Te,95 (i.e. Te at 95% normalized poloidal flux surface) at the L–H transition are similar for the two divertor configurations.

A coordinated effort to measure divertor heat flux characteristics in fully attached, similarly shaped H-mode plasmas on C-Mod, DIII-D, and NSTX was carried out in 2010 in order to construct a predictive scaling relation applicable to next step devices including ITER, FNSF, and DEMO. Few published scaling laws are available and those that have been published were obtained under widely varying conditions and divertor geometries, leading to conflicting predictions for this critically important quantity. This study was designed to overcome these deficiencies. Analysis of the combined data set reveals that the primary dependence of the parallel heat flux width is robustly inverse with Ip, which all three tokamaks independently demonstrate. An improved Thomson scattering system on DIII-D has yielded very accurate scrape off layer (SOL) profile measurements from which tests of parallel transport models have been made. It is found that a flux-limited model agrees best with the data at all collisionalities, while...

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

Two 7-mm diameter, remote-head, visible, charge-coupled device (CCD) cameras are used on the Alcator C-Mod tokamak to generate two-dimensional emissivity profiles of deuterium line radiation (typically at Balmer /spl alpha/ and Balmer /spl gamma/) from the divertor. These emissivity profiles are obtained by reconstructing the measured brightnesses using the thin-chord approximation and assuming toroidal symmetry of the emission. The regions of the divertor with a high ratio of Balmer /spl gamma/ to Balmer /spl alpha/ emission are indicative of a plasma that is recombining. These recombining regions in the divertor are seen to shift from the inner to the outer divertor as the core density in increased.

Power requirements for maintaining sufficiently high confinement (i.e. normalized energy confinement time H 98 ⩾ 1) in H-mode and its relation to H-mode threshold power scaling, P th, are of critical importance to ITER. In order to better characterize these power requirements, recent experiments on the Alcator C-Mod tokamak have investigated H-mode properties, including the edge pedestal and global confinement, over a range of input powers near and above P th. In addition, we have examined the compatibility of impurity seeding with high performance operation, and the influence of plasma radiation and its spatial distribution on performance. Experiments were performed at 5.4 T at ITER relevant densities, utilizing bulk metal plasma facing surfaces and an ion cyclotron range of frequency waves for auxiliary heating. Input power was scanned both in stationary enhanced Dα (EDA) H-modes with no large edge localized modes (ELMs) and in ELMy H-modes in order to relate the resulting pedestal and confinement to the amount of power flowing into the scrape-off layer, P net, and also to the divertor targets. In both EDA and ELMy H-mode, energy confinement is generally good, with H 98 near unity. As P net is reduced to levels approaching that in L-mode, pedestal temperature diminishes significantly and normalized confinement time drops. By seeding with low-Z impurities, such as Ne and N2, high total radiated power fractions are possible, along with substantial reductions in divertor heat flux (>4×), all while maintaining H 98 ∼ 1. When the power radiated from the confined versus unconfined plasma is examined, pedestal and confinement properties are clearly seen to be an increasing function of P net, helping to unify the results with those from unseeded H-modes. This provides increased confidence that the power flow across the separatrix is the correct physics basis for ITER extrapolation. The experiments show that P net/P th of one or greater is likely to lead to H 98 ⩾ 1 operation, and also that such a condition can be made compatible with a low-Z radiative impurity solution for reducing divertor heat loads to levels acceptable for ITER.

We report on the status of a set of ITPA-coordinated experiments between the Alcator C-Mod, MAST and NSTX devices to compare the characteristics and access conditions of discharges with small edge-localized modes (ELMs). The small ELMs in C-Mod, MAST and one of the two small ELM types in NSTX exist when approached 10–15%, although the lower/upper limits of the operational windows differ. These small ELM regimes appear in diverted configurations very close to balanced double-null in each device. We classify these small ELMs as type II, based on the published characteristics from a number of previous studies. In addition, these type II ELMs in each device had multiple filaments with propagation in the co-I p or ion diamagnetic drift direction. Moreover, we conclude that these type II ELMs are distinct from the type V ELMs routinely observed in NSTX, which have one or two filaments and propagate in the electron diamagnetic drift direction.

Experiments in Alcator C-Mod in (Enhanced D-alpha) EDA H-modes with extrinsic impurity seeding (N2, Ne, and Ar) have demonstrated a direct correlation between plasma energy confinement and edge power flow, achieving values of H98 ≥ 1 for edge power flows only marginally exceeding the scaled power for access to H-mode confinement in these conditions. For lower Z impurity seeding (N2 and Ne), plasmas with high energy confinement are obtained with a radiative power fraction of 85% or larger and a reduction of the peak heat flux at the divertor by more than a factor of 5 compared to similar attached conditions. The H-mode plasmas thus achieved in Alcator C-Mod meet or exceed the requirements both in terms of divertor heat flux handling and energy confinement for ITER QDT = 10 operation and with an edge power flow only marginally above the H-mode threshold power (by 1.0–1.4) as expected in ITER.

We report extended studies of the I-mode regime [Whyte et al., Nucl. Fusion 50, 105005 (2010)] obtained in the Alcator C-Mod tokamak [Marmar et al., Fusion Sci. Technol. 51(3), 3261 (2007)]. This regime, usually accessed with unfavorable ion B × ∇B drift, features an edge thermal transport barrier without a strong particle transport barrier. Steady I-modes have now been obtained with favorable B × ∇B drift, by using specific plasma shapes, as well as with unfavorable drift over a wider range of shapes and plasma parameters. With favorable drift, power thresholds are close to the standard scaling for L–H transitions, while with unfavorable drift they are ∼ 1.5–3 times higher, increasing with Ip. Global energy confinement in both drift configurations is comparable to H-mode scalings, while density profiles and impurity confinement are close to those in L-mode. Transport analysis of the edge region shows a decrease in edge χeff, by typically a factor of 3, between L- and I-mode. The decrease correlates with ...

Parametric dependences of the heat flux footprint on the outer divertor target plate are explored in EDA H-mode and ohmic L-mode plasmas over a wide range of parameters with attached plasma conditions. Heat flux profile shapes are found to be independent of toroidal field strength, independent of power flow along magnetic field lines and insensitive to x-point topology (single-null versus double-null). The magnitudes and widths closely follow that of the “upstream” pressure profile, which are correlated to plasma thermal energy content and plasma current. Heat flux decay lengths near the strike-point in H- and L-mode plasmas scale approximately with the inverse of plasma current, with a diminished dependence at high collisionality in L-mode. Consistent with previous studies, pressure gradients in the boundary scale with plasma current squared, holding the magnetohydrodynamic ballooning parameter approximately invariant at fixed collisionality—strong evidence that critical-gradient transport physics plays a key role in setting the power exhaust channel.

Plasma edge turbulence of Alcator C-Mod tokamak is studied with a fast camera in different density regimes. The statistical properties of the fluctuations, as well as the behaviour of the blobs, are characterized in plasma discharges at different normalized densities, studying the link between the edge turbulence and the Greenwald limit. It is shown that approaching the Greenwald density limit, the edge velocity field measured with the cross-correlation technique changes and the strong fluctuations, which for standard discharges develop mainly outside the separatrix, extend also in the radial region inside the last closed flux surface. At the same time, the blobs cover a larger radial region, suggesting a strong impact of the edge turbulence and transport on the Greenwald limit.

Vacuum ultraviolet spectroscopy is used on the Alcator C-Mod tokamak to study the physics of impurity transport and provide feedback on impurity levels to assist experimental operations. Sputtering from C-Mod's all metal (Mo+W) plasma facing components and ion cyclotron range of frequency antenna and vessel structures (sources for Ti, Fe, Cu, and Ni), the use of boronization for plasma surface conditioning and Ar, Ne, or N(2) gas seeding combine to provide a wealth of spectroscopic data from low-Z to high-Z. Recently, a laser blow-off impurity injector has been added, employing CaF(2) to study core and edge impurity transport. One of the primary tools used to monitor the impurities is a 2.2 m Rowland circle spectrometer utilizing a Reticon array fiber coupled to a microchannel plate. With a 600 lines/mm grating the 80

The steep gradient edge region and scrape-off-layer (SOL) on the low-field-side of Alcator C-Mod [I. H. Hutchinson, R. Boivin, F. Bombarda et al., Phys. Plasmas 1, 1511 (1994)] tokamak plasmas are studied using gas-puff-imaging diagnostics. In L-mode plasmas, the region extending ∼2 cm inside the magnetic separatrix has fluctuations showing a broad, turbulent spectrum, propagating in the electron diamagnetic drift direction, whereas features in the open field line region propagate in the ion diamagnetic drift direction. This structure is robust against toroidal field strength, poloidal null-point geometry, plasma current, and plasma density. Global parameter dependence of spectral and spatial structure of the turbulence inside the separatrix is explored and characterized, and both the intensity and spectral distributions are found to depend strongly on the plasma density normalized to the tokamak density limit. In H-mode discharges the fluctuations at and inside the magnetic separatrix show fundamentally ...

Alcator C-MOD has compared plasma performance with plasma-facing components (PFCs) coated with boron to all-metal PFCs to assess projections of energy confinement from current experiments to next-generation burning tokamak plasmas. Low-Z coatings reduce metallic impurity influx and diminish radiative losses leading to higher H-mode pedestal pressure that improves global energy confinement through profile stiffness. RF sheath rectification along flux tubes that intersect the RF antenna is found to be a major cause of localized boron erosion and impurity generation. Initial lower hybrid current drive (LHCD) experiments (PLH < 900?kW) in preparation for future advanced-tokamak studies have demonstrated fully non-inductive current drive at Ip ~ 1.0?MA with good efficiency, Idrive = 0.4 PLH/neoR (MA, MW, 1020?m?3,m). The potential to mitigate disruptions in ITER through massive gas-jet impurity puffing has been extended to significantly higher plasma pressures and shorter disruption times. The fraction of total plasma energy radiated increases with the Z of the impurity gas, reaching 90% for krypton. A positive major-radius scaling of the error field threshold for locked modes (Bth/B ? R0.68?0.19) is inferred from its measured variation with BT that implies a favourable threshold value for ITER. A phase contrast imaging diagnostic has been used to study the structure of Alfv?n cascades and turbulent density fluctuations in plasmas with an internal transport barrier. Understanding the mechanisms responsible for regulating the H-mode pedestal height is also crucial for projecting performance in ITER. Modelling of H-mode edge fuelling indicates high self-screening to neutrals in the pedestal and scrape-off layer (SOL), and reproduces experimental density pedestal response to changes in neutral source, including a weak variation of pedestal height and constant width. Pressure gradients in the near SOL of Ohmic L-mode plasmas are observed to scale consistently as , and show a significant dependence on X-point topology. Fast camera images of intermittent turbulent structures at the plasma edge show they travel coherently through the SOL with a broad radial velocity distribution having a peak at about 1% of the ion sound speed, in qualitative agreement with theoretical models. Fast D? diagnostics during gas puff imaging show a complex behaviour of discrete ELMs, starting with an n ? 10 precursor oscillation followed by a rapid primary ejection as the pedestal crashes and then multiple, slower secondary ejections.

The high-frequency range of power spectra of turbulent fluctuating quantities measured at the edge of magnetized plasmas displays a variety of trends: from power laws with different spectral indices to exponential. We propose a model able to account for the whole phenomenology simply by tuning the distribution in the duration of the signal spikes. Comparisons with data from RFX-mod and Alcator C-Mod experiments are performed. An attempt to relate the statistics of the bursts with their generating mechanism is made.

This paper describes quantitative comparisons between turbulence measured in the scrape-off layer (SOL) of Alcator C-Mod [S. Scott, A. Bader, M. Bakhtiari et al., Nucl. Fusion 47, S598 (2007)] and three dimensional computations using electromagnetic gyrofluid equations in a two-dimensional tokamak geometry. These comparisons were made for the outer midplane SOL for a set of inner-wall limited, near-circular Ohmic plasmas. The B field and plasma density were varied to assess gyroradius and collisionality scaling. The poloidal and radial correlation lengths in the experiment and computation agreed to within a factor of 2 and did not vary significantly with either B or density. The radial and poloidal propagation speeds and the frequency spectra and poloidal k-spectra also agreed fairly well. However, the autocorrelation times and relative Da fluctuation levels were higher in the experiment by more than a factor of 2. Possible causes for these disagreements are discussed. 2009 American Institute of Physics.

The visibility of carbon dust particles in fusion plasmas with fast framing cameras is evaluated using computer and theoretical modelling. Dust heating and ablation in the plasmas, ionization dynamics of the ablation cloud around the dust particle, thermal radiation of the dust and line radiation of the cloud are considered. The minimal size of carbon dust particles visible with the cameras is calculated as a function of the plasma parameters and the distance from the dust to the camera objective. The relative contributions of thermal radiation from the grain and line radiation from the cloud to the total dust radiation are analysed.

A review of TFTR plasma transport studies is presented. Parallel transport and the confinement of suprathermal ions are found to be relatively well described by theory. Cross-field transport of the thermal plasma, however, is anomalous with the momentum diffusivity being comparable to the ion thermal diffusivity and larger than the electron thermal diffusivity in neutral beam heated discharges. Perturbative experiments have studied nonlinear dependencies in the transport coefficients and examined the role of possible nonlocal phenomena. The underlying turbulence has been studied using microwave scattering, beam emission spectroscopy and microwave reflectometry over a much broader range in k perpendicular to than previously possible. Results indicate the existence of large-wavelength fluctuations correlated with enhanced transport.

It has been known for a long time that microscopic dust appears in plasmas in fusion devices. Recently it was shown that dust can be responsible for the termination of long- discharges. Also, in ITER-scale experiments dust can pose safety problems related to its chemical activity, tritium retention and radioactive content. In particular, the presence of dust in the vacuum chamber of ITER is one of the main concerns of the ITER licensing process. Here we review recent progress in the understanding of different experimental and theoretical aspects of the physics of dust dynamics and transport in fusion plasmas and discuss the remaining issues.

Recent experiments have led to a fundamental shift in our view of edge transport physics; transport near the last-closed flux surface may be more appropriately described in terms of a critical gradient phenomenon rather than a diffusive and/or convective paradigm. Edge pressure gradients, normalized by the square of the poloidal magnetic field strength, appear invariant in plasmas with the same normalized collisionality, despite vastly different currents and magnetic fields—a behavior that connects with first-principles electromagnetic plasma turbulence simulations. Near-sonic scrape-off layer (SOL) flows impose a cocurrent rotation boundary condition on the confined plasma when B×∇B points toward the active x-point, suggesting a link to the concomitant reduction in input power needed to attain high-confinement modes. Indeed, low-confinement mode plasmas are found to attain higher edge pressure gradients in this configuration, independent of the direction of B, evidence that SOL flows may affect transport and “critical gradient” values in the edge plasma.

This paper reports on studies of the edge transport barrier and transition threshold of the high confinement (H) mode of operation on the Alcator C-Mod tokamak [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)], over a wide range of toroidal field (2.6–7.86T) and plasma current (0.4–1.7MA). The H-mode power threshold and edge temperature at the transition increase with field. Barrier widths, pressure limits, and confinement are nearly independent of field at constant current, but the operational space at high B shifts toward higher temperature and lower density and collisionality. Experiments with reversed field and current show that scrape-off-layer flows in the high-field side depend primarily on configuration. In configurations with the B×∇B drift away from the active X-point, these flows lead to more countercurrent core rotation, which apparently contributes to higher H-mode thresholds. In the unfavorable case, edge temperature thresholds are higher, and slow evolution of profiles indicates a red...

A novel Doppler spectrometer is currently being used for ion or neutral velocity and temperature measurements on the Alcator C-Mod tokamak. The spectrometer has an f∕# of ∼3.1 and is appropriate for visible light (3500–6700A). The linewidth from a line emitting calibration source has been measured to be as small as 0.4A. The ultimate time resolution is line brightness light limited and on the order of milliseconds. A new photon efficient charge coupled device detector is being used at C-Mod. Time resolution is achieved by moving the camera during a plasma discharge in a perpendicular direction through the dispersion plane of the spectrometer, causing a vertical streaking across the camera face. Initial results from C-Mod as well as previous measurements from the Compact Toroid Injection Experiment are presented.

In this paper we compare the structure and motion of edge turbulence observed in L-mode vs. H-mode plasmas in the National Spherical Torus Experiment (NSTX) [M. Ono, M. G. Bell, R. E. Bell et al., Plasma Phys. Controlled Fusion 45, A335 (2003)]. The radial and poloidal correlation lengths are not significantly different between the L-mode and the H-mode in the cases examined. The poloidal velocity fluctuations are lower and the radial profiles of the poloidal turbulence velocity are somewhat flatter in the H-mode compared with the L-mode plasmas. These results are compared with similar measurements Alcator C-Mod [E. Marmar, B. Bai, R. L. Boivin et al., Nucl. Fusion 43, 1610 (2003)], and with theoretical models.

Factor of ∼2 higher power thresholds for low- to high-confinement mode transitions (L-H) with unfavorable x-point topologies in Alcator C-Mod [Phys. Plasmas 1, 1511 (1994)] are linked to flow boundary conditions imposed by the scrape-off layer (SOL). Ballooning-like transport drives flow along magnetic field lines from low- to high-field regions with toroidal direction dependent on upper/lower x-point balance; the toroidal rotation of the confined plasma responds, exhibiting a strong counter-current rotation when B×∇B points away from the x point. Increased auxiliary heating power (rf, no momentum input) leads to an L-H transition at approximately twice the edge electron pressure gradient when B×∇B points away. As gradients rise prior to the transition, toroidal rotation ramps toward the co-current direction; the H mode is seen when the counter-current rotation imposed by the SOL flow becomes compensated. Remarkably, L-H thresholds in lower-limited discharges are identical to lower x-point discharges; SOL...

The recently upgraded phase-contrast imaging (PCI) diagnostic is used to characterize the transition from the low (L) to the enhanced Dα (EDA) high (H) confinement mode in Alcator C-Mod [I. H. Hutchinson, R. Boivin, F. Bombarda et al., Phys. Plasmas 1, 1511 (1994)] plasmas. PCI yields information on line integrated density fluctuations along vertical chords. The number of channels has been increased from 12 to 32 and the sampling rate from 1 MHz to 10 MHz. This expansion of diagnostic capabilities is used to study broadband turbulence in L and EDA H mode and to analyze the quasicoherent (QC) mode associated with EDA H mode. Changes in broadband turbulence at the transition from L to EDA H mode can be interpreted as an effect of the Doppler rotation of the bulk plasma. Additional fluctuation measurements of Dα light and the poloidal magnetic field show features correlated with PCI in two different frequency ranges at the transition. The backtransition from EDA H to L mode, the so-called enhanced neutron (E...

An eight-channel beam-emission-spectroscopy (BES)1 system has been installed on the Alcator C-Mod tokamak, intended for use with a diagnostic neutral hydrogen beam (DNB). Capable of localized measurements from the plasma edge to the plasma core, the BES diagnostic collects light from the first Balmer transition (Hα) resultant from beam/plasma collisions. The Hα line splits into several components whose central wavelengths depend on the viewing geometry, the magnetic field, and the beam energy. This is due to the Doppler shifts from viewing the beam off perpendicular, the different velocities of the three mass components of the beam (H, H2, H3), and the large motional Stark effect. Optimal signal-to-noise requires collecting these components while attenuating all other emission: primarily bremsstrahlung and Dα radiation (from plasma D0/e− collisions). Tunable bandpass filters are thus required. A BES simulation code has been developed that calculates the brightnesses (bremsstrahlung, Dα, Hα) versus wavelen...

Spectral line shifts have been observed in the Dα transition of neutral deuterium in the Alcator C-mod [I. H. Hutchinson, Proceedings of the IEEE 13th Symposium on Fusion Engineering (IEEE, Piscataway, 1990), Vol. 1, p. 13] tokamak and interpreted as Doppler shifts due to a neutral flow of approximately 5×103 m/s. The emission originates from plasma at the inner wall, as indicated by the Zeeman patterns of the transition. The direction of the flow changes when the plasma transitions from a limited to a diverted configuration. In all cases, the flows are directed toward the point of contact with the wall or divertor plate. Narrow spectral profiles have been observed for the Dα transition near the midplane and within the divertor. Temperatures not higher than 0.86 eV are indicated for the divertor measurements. A difference in plasma conditions for separate locations along a single line of sight has been observed.

The Rydberg series up to n = 14 of helium-like chlorine, argon and sulfur have been observed in Alcator C-Mod plasmas. High-n satellites to these lines of the form 1s22s-1s2snp and 1s22p-1s2pnp with 3 n12 have also been seen for chlorine and argon. Accurate wavelengths of these satellites have been obtained, comparison has been made with the theoretical predictions from the atomic structure codes RELAC and MZ, and the agreement is good. Measured line intensities have also been compared with collisional radiative modelling that includes the contributions from dielectronic recombination and inner-shell excitation rates to each line's emission, again with good agreement.

Plasma heating by directly launched ion Bernstein waves is investigated in the Alcator C tokamak at densities n-bar/sub e/approx. >1 x 10/sup 20/ m/sup -3/. For injected rf powers P/sub rf/approx. 2 x 10/sup 20/ m/sup -3/), the heating ceases to be efficient.

Transport of trace, non-recycling, injected impurities has been studied on the Alcator C tokamak. Changes of impurity confinement times with varying plasma density, current, toroidal field, majority ion species mass, impurity charge and mass, Zeff, and major and minor radius have been delineated. An empirical scaling is developed from these results and compared with the results of similar transport studies undertaken on other tokamak devices. The agreement is reasonable. A computer model simulating the transport is utilized to compare several models with the empricial results. With the possible exception of low-density, high-Zeff discharges, the transport is not consisten with the predictions of neoclassical theory, but can be well described by simple spreading diffusion with a diffusion coefficient ranging from 1 to 5 × 103 cm2s−1, depending on plasma parameters. This model yields good agreement both with the time histories of single-chord measurements of various ionization states, and with radial soft-X-ray emission profiles. Increased impurity transport with the onset of strong MHD oscillations has also been observed, with the effective diffusion coefficient scaling approximately as (ΔB)4.

The effectiveness of plasma heating by electron Landau interaction in the lower hybrid range of frequencies in tokamak plasmas is demonstrated. Upon injection of 850 kW of rf power at a density of n-barsubeapprox. =1.4 x 10sup14 cmsup-3, an electron temperature increase of 1.0 keV and an ion temperature increase of 0.8 keV was achieved. These results are compared with transport and ray-tracing code predictions.

The influx of heavy impurities in the Alcator C tokamak is determined as a function of plasma parameters from observations of intrinsic impurities, in conjunction with an empirically derived anomalous impurity diffusion model. The influx of molybdenum as a function of electron density is found to decrease dramatically as the electron density is raised above 1 × 1014 cm−3. Sputtering (by neutrals, ions and impurities) is probably the dominant molybdenum release mechanism in Ohmically heated discharges.

Energy confinement is studied in lower hybrid current driven (LHCD) plasmas in Alcator C in the density range e = (1?8) ? 1013 cm?3. In LHCD plasmas, the stored energy in the electron tail, Wtail, can be a significant fraction of the total stored energy, Wtot, especially at lower densities. At sufficiently low densities, the confinement time of the high energy electrons is expected to become shorter than their collisional slowing down time, and direct energy losses from the electron tail can become important in the overall power balance. The global energy confinement time, defined as is found to be comparable to or exceed that in ohmically heated (OH) plasmas at low densities e 3 ? 1013 cm?3, where a steady state current can be maintained with relatively low RF power. However, at higher densities where substantially more RF power is needed (relative to the Ohmic power required to maintain a similar plasma), a deterioration of relative to Ohmic confinement, similar in magnitude to that predicted by the neutral beam heated L-mode scaling, is observed. Theoretical modelling with the aid of a ray tracing Fokker-Planck transport code suggests that the deteriorated confinement in this high density, high power regime may be attributed to an enhanced bulk electron thermal diffusivity. In a combined OH-LHCD plasma, a value of greater than the Ohmic value is obtained as long as the applied RF power does not significantly exceed the Ohmic power.

Observations of impurity behaviour are presented from ICRF heating experiments at 180 MHz performed for a variety of conditions on the Alcator C tokamak, using graphite limiters and stainless steel antenna Faraday shields. Spectroscopic observations revealed significant increases in metal impurity concentrations during the RF pulse, with iron levels increasing by as much as a factor of twelve at the highest RF powers (about 350–400 kW). Analysis of the inferred iron source rates shows an approximately linear dependence on RF power up to 400 kW, with no clear dependence on resonance conditions or bulk plasma parameters. However, a sharp temperature increase in the limiter shadow region was observed during the ICRF pulse, which was well correlated with the iron influx rate. From this and other evidence it is concluded that physical sputtering of the Faraday shield due to an elevated sheath potential is the primary source of metal impurities during ICRF heating on Alcator C. The same process, occurring at the graphite limiter, is believed to be the dominant source of carbon and oxygen. The calculated sputtering yields obtained from an edge erosion code demonstrate the plausibility of this model.

A series of pellet-fueling experiments has been carried out on the Alcator $C$ tokamak. High-speed hydrogen pellets penetrate to within a few centimeters of the magnetic axis, raise the plasma density, and produce peaked density profiles. Energy confinement is observed to increase over similar discharges fueled only by gas puffing. In this manner record values of electron density, plasma pressure, and Lawson number ($n\ensuremath{\tau}$) have been achieved.

Recent results of the neutral-beam heating experiments (NBI) on the Heliotron E plasma are reported. The target plasma for NBI is produced by electron cyclotron resonance heating (ECRH), using two gyrotrons (53.2 GHz, 200 kW × 2). The neutral beams of 2.9 MWmax are injected 20 to 40 ms afterwards in order to produce a high-energy currentless plasma. At a magnetic field of 1.9 T, a maximum ion temperature of 1.0 keV is obtained at a density of 1.9 × 1013 cm−3. The maximum eτE is 2.4 × 1012 cm −3⋅s. By reducing the magnetic field to 0.94 T and using the second-harmonic resonance of microwaves to produce a target plasma, a high-beta plasma is produced by NBI. The maximum central-beta and volume-average beta values are 3.6% and 2%, respectively. The paper describes the results of NBI heating experiments and the confinement and MHD activity properties of the plasmas in question.