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The optimized superconducting stellarator device Wendelstein 7-X (with major radius , minor radius , and plasma volume) restarted operation after the assembly of a graphite heat shield and 10 inertially cooled island divertor modules. This paper reports on the results from the first high-performance plasma operation. Glow discharge conditioning and ECRH conditioning discharges in helium turned out to be important for density and edge radiation control. Plasma densities of with central electron temperatures were routinely achieved with hydrogen gas fueling, frequently terminated by a radiative collapse. In a first stage, plasma densities up to were reached with hydrogen pellet injection and helium gas fueling. Here, the ions are indirectly heated, and at a central density of a temperature of with was transiently accomplished, which corresponds to with a peak diamagnetic energy of and volume-averaged normalized plasma pressure . The routine access to high plasma densities was opened with boronization of the first wall. After boronization, the oxygen impurity content was reduced by a factor of 10, the carbon impurity content by a factor of 5. The reduced (edge) plasma radiation level gives routinely access to higher densities without radiation collapse, e.g. well above line integrated density and central temperatures at moderate ECRH power. Both X2 and O2 mode ECRH schemes were successfully applied. Core turbulence was measured with a phase contrast imaging diagnostic and suppression of turbulence during pellet injection was observed.

In this work we present an analysis of fast-camera imaging of Alcator C-Mod's divertor region and of the plasma cross section in the W7-X stellarator. In C-Mod, the dynamics of intermittent edge plasma filaments are imaged at 397,100 frames per second in an oblique view. An automatic pipeline was developed to reconstruct the filament motion in the R, Z plane of the divertor using EFIT field line projections. Next, we added our fast-framing camera to W7-X's diagnostic suite for the 2017 run campaign, the first with a divertor. It has a tangential view of the machine cross section, including part of a divertor module. The view is coupled to the camera via a coherent fiber bundle. The spatial resolution is 2–4 cm and the typical plasma brightness limits the fastest usable frame rate to 30k frames/s. We observe filaments in the plasma boundary that rotate around the magnetic axis during radiative collapse; filaments were observed in steady state in the limiter configuration of W7-X by Kocsis et al. [1]. We also observe for the first time a 1–2 kHz quasi-coherent fluctuation in divertor target emissions in the W7-X “standard” magnetic configuration. Finally, a rough analysis of edge gas fueling efficiency using fast camera data and S/XB coefficients finds that fueling is less efficient at densities higher than 4 × 1019 m−3. Keywords: C-Mod, W7-X, Filaments, Blobs, Turbulence, Fluctuation

Owing to its high magnetic field, high power, and compact size, the SPARC experiment will operate with divertor conditions at or above those expected in reactor-class tokamaks. Power exhaust at this scale remains one of the key challenges for practical fusion energy. Based on empirical scalings, the peak unmitigated divertor parallel heat flux is projected to be greater than 10 GW m−2. This is nearly an order of magnitude higher than has been demonstrated to date. Furthermore, the divertor parallel Edge-Localized Mode (ELM) energy fluence projections (~11–34 MJ m−2) are comparable with those for ITER. However, the relatively short pulse length (~25 s pulse, with a ~10 s flat top) provides the opportunity to consider mitigation schemes unsuited to long-pulse devices including ITER and reactors. The baseline scenario for SPARC employs a ~1 Hz strike point sweep to spread the heat flux over a large divertor target surface area to keep tile surface temperatures within tolerable levels without the use of active divertor cooling systems. In addition, SPARC operation presents a unique opportunity to study divertor heat exhaust mitigation at reactor-level plasma densities and power fluxes. Not only will SPARC test the limits of current experimental scalings and serve for benchmarking theoretical models in reactor regimes, it is also being designed to enable the assessment of long-legged and X-point target advanced divertor magnetic configurations. Experimental results from SPARC will be crucial to reducing risk for a fusion pilot plant divertor design.

New experiments on Alcator C-Mod reveal that the favorable impurity screening characteristics of the high-field side (HFS) scrape-off layer (SOL), previously reported for single null geometries, is retained in double null configurations, despite the formation of an extremely thin SOL. In balanced double-null, nitrogen injected locally into the HFS SOL is better screened by a factor of 2.5 compared to the same injection into the low field side (LFS) SOL. This result is insensitive to plasma current and Greenwald fraction. Nitrogen injected into the HFS SOL is not as well screened (only a factor of 1.5 improvement over LFS) in unbalanced double-null discharges, when the primary divertor is in the direction of B×∇B. In this configuration, impurity ‘plume’ emission patterns indicate that an opposing E × B drift competes with the parallel impurity flow to the divertor. In balanced double-null plasmas, the dispersal pattern exhibits a dominant E × B motion. Unbalanced discharges with the primary divertor opposite the direction of B×∇B exhibit excellent HFS screening characteristics – a factor of 5 enhancement compared to LFS. These data support the idea that future tokamaks should locate all RF actuators and close-fitting wall structures on the HFS and employ near-double-null magnetic topologies, both to precisely control plasma conditions at the antenna/plasma interface and to maximally mitigate the impact of local impurity sources arising from plasma-material interactions. Keywords: Alcator C-Mod; Impurity screening; Double null; High field side scrape-off layer, United States. Department of Energy (Contract DE-FC02-99ER54512)

A study of long-legged tokamak divertor configurations is performed with the edge transport code UEDGE (Rognlien et al., J. Nucl. Mater. 196, 347, 1992). The model parameters are based on the ADX tokamak concept design (LaBombard et al., Nucl. Fusion 55, 053020, 2015). Several long-legged divertor configurations are considered, in particular the X-point target configuration proposed for ADX, and compared with a standard divertor. For otherwise identical conditions, a scan of the input power from the core plasma is performed. It is found that as the power is reduced to a threshold value, the plasma in the outer leg transitions to a fully detached state which defines the upper limit on the power for detached divertor operation. Reducing the power further results in the detachment front shifting upstream but remaining stable. At low power the detachment front eventually moves to the primary X-point, which is usually associated with degradation of the core plasma, and this defines the lower limit on the power for the detached divertor operation. For the studied parameters, the operation window for a detached divertor in the standard divertor configuration is very small, or even non-existent; under the same conditions for long-legged divertors the detached operation window is quite large, in particular for the X-point target configuration, allowing a factor of 5–10 variation in the input power. These modeling results point to possibility of stable fully detached divertor operation for a tokamak with extended divertor legs., United States. Department of Energy (Contract DE-AC52-07NA27344)

Plasma fluctuations in the scrape-off layer of the Alcator C-Mod tokamak in ohmic and high confinement modes have been analyzed using gas puff imaging data. In all cases investigated, the time series of emission from a single spatially-resolved view into the gas puff are dominated by large-amplitude bursts, attributed to blob-like filament structures moving radially outwards and poloidally. There is a remarkable similarity of the fluctuation statistics in ohmic plasmas and in edge localized mode-free and enhanced D-alpha high confinement mode plasmas. Conditionally averaged wave forms have a two-sided exponential shape with comparable temporal scales and asymmetry, while the burst amplitudes and the waiting times between them are exponentially distributed. The probability density functions and the frequency power spectral densities are self-similar for all these confinement modes. These results are strong evidence in support of a stochastic model describing the plasma fluctuations in the scrape-off layer as a super-position of uncorrelated exponential pulses. Predictions of this model are in excellent agreement with experimental measurements in both ohmic and high confinement mode plasmas. The stochastic model thus provides a valuable tool for predicting fluctuation-induced plasma-wall interactions in magnetically confined fusion plasmas., 17 pages, 10 figures

This document is the product of a stellarator community workshop, organized by the National Stellarator Coordinating Committee and referred to as Stellcon, that was held in Cambridge, Massachusetts in February 2016, hosted by MIT. The workshop was widely advertised, and was attended by 40 scientists from 12 different institutions including national labs, universities and private industry, as well as a representative from the Department of Energy. The final section of this document describes areas of community wide consensus that were developed as a result of the discussions held at that workshop. Areas where further study would be helpful to generate a consensus path forward for the US stellarator program are also discussed. The program outlined in this document is directly responsive to many of the strategic priorities of FES as articulated in “Fusion Energy Sciences: A Ten-Year Perspective (2015–2025)” [1]. The natural disruption immunity of the stellarator directly addresses “Elimination of transient events that can be deleterious to toroidal fusion plasma confinement devices” an area of critical importance for the US fusion energy sciences enterprise over the next decade. Another critical area of research “Strengthening our partnerships with international research facilities,” is being significantly advanced on the W7-X stellarator in Germany and serves as a test-bed for development of successful international collaboration on ITER. This report also outlines how materials science as it relates to plasma and fusion sciences, another critical research area, can be carried out effectively in a stellarator. Additionally, significant advances along two of the Research Directions outlined in the report; “Burning Plasma Science: Foundations—Next-generation research capabilities”, and “Burning Plasma Science: Long pulse—Sustainment of Long-Pulse Plasma Equilibria” are proposed.

A prototype of an infrared imaging bolometer (IRVB) was successfully tested on the Alcator C-Mod tokamak at the end of its 2016 campaign. The IRVB method interprets the power radiated from the plasma by measuring the temperature rise of a thin, ∼2 μm, Pt absorber that is placed in the torus vacuum and exposed, using a pinhole camera, to the full-spectrum of plasma’s photon emission. The IRVB installed on C-Mod viewed the poloidal cross section of the core plasma and observed Ohmic and ion cyclotron range of frequency (ICRF)-heated plasmas. Analysis of total radiated power and on-axis emissivity from IRVB is summarized, and quantitative comparisons made to data from both resistive bolometers and AXUV diodes. IRVB results are clearly within a factor of two, but additional effort is needed for it to be used to fully support power exhaust research. The IRVB is shown to be immune to electromagnetic interference from ICRF which strongly impacts C-Mod’s resistive bolometers. Results of the bench-top calibration are summarized, including a novel temperature calibration method useful for IRVBs.

A comprehensive study of fully frequency-resolved nonlinear kinetic energy transfer has been performed for the first time in a diverted tokamak, providing new insight into the parametric dependences of edge turbulence transitions. Measurements using gas puff imaging in the turbulent L-mode state illuminate the source of the long known but as yet unexplained "favorable-unfavorable" geometric asymmetry of the power threshold for transition to the turbulence-suppressed H mode. Results from the recently discovered I mode point to a competition between zonal flow (ZF) and geodesic-acoustic modes (GAM) for turbulent energy, while showing new evidence that the I-to-H transition is still dominated by ZFs. The availability of nonlinear drive for the GAM against net heat flux through the edge corresponds very well to empirical scalings found experimentally for accessing the I mode.

Fluctuations in the boundary region of the Alcator C-Mod tokamak have been analyzed using gas puff imaging data from a set of Ohmically heated plasma density scan experiments. It is found that the relative fluctuation amplitudes are modest and close to normally distributed at the separatrix but become increasingly larger and intermittent towards the main chamber wall. The frequency power spectra are nevertheless similar for all radial positions and line-averaged densities. Predictions of a stochastic model, describing the plasma fluctuations as a super-position of uncorrelated pulses, are shown to be in excellent agreement with the measurements. This implies that the pulse duration is the same, while the degree of pulse overlap decreases radially outwards in the scrape-off layer. The model also describes the rate of threshold level crossings, which provide novel predictions of plasma--wall interactions due to transient transport events., 8 figures

In order to more completely demonstrate the I-mode regime as a compelling fusion reactor operating scenario, the first dedicated attempts at I-mode radiative heat exhaust and detachment were carried out on Alcator C-Mod. Results conclusively show that within the parameter space explored, an I/L back-transition is triggered prior to meaningful reductions in parallel heat flux, q||, target temperature, Te;tar, and target pressure, pe;tar, at the outer divertor. The exact mechanism for the I/L trigger remains uncertain, but a multi-diagnostic investigation suggests the pedestal regulation physics is impacted promptly by small amounts of N2 seeded into the private flux region. The time delay between when N2 contacts the plasma and the I/L transition is triggered varied from 30-120 ms, approximately 0.7-3 x tE, and the delay varied inversely with I-mode pedestal-top pressure, pe;95. Power and nitrogen influx scans indicate that the I/L transitions are not linked to excessive bulk-plasma impurity radiation. It is also shown that in the subsequent L-mode following nitrogen seeding, q|| and Te;tar can be reduced by factors of ~10. The I/L transition and L-mode exhaust results using N2 are compared to similar attempts using Ne where such q|| and Te;tar reductions in L-mode are limited to factors of 2-3. Implications for the I-mode regime are discussed, including needs for follow-up experiments on other facilities.

For the first time, the power deposition of lower hybrid RF waves into the edge plasma of a diverted tokamak has been systematically quantified. Edge deposition represents a parasitic loss of power that can greatly impact the use and efficiency of Lower Hybrid Current Drive (LHCD) at reactor-relevant densities. Through the use of a unique set of fast time resolution edge diagnostics, including innovative fast-thermocouples, an extensive set of Langmuir probes, and a Lyα ionization camera, the toroidal, poloidal, and radial structure of the power deposition has been simultaneously determined. Power modulation was used to directly isolate the RF effects due to the prompt (t1.0×10[superscript 20] (m[superscript −3])). Results will be shown addressing the distribution of power within the SOL, including the toroidal symmetry and radial distribution. These characteristics are important for deducing the cause of the reduced LHCD efficiency at high density and motivate the tailoring of wave propagation to minimize SOL interaction, for example, through the use of high-field-side launch., United States. Department of Energy. Office of Fusion Energy Sciences (Award No. DE-FC02-99ER54512-CMOD)

We study long time series of the ion saturation current and floating potential, sampled by Langmuir probes dwelled in the outboard mid-plane scrape off layer and embedded in the lower divertor baffle of Alcator C-Mod. A series of ohmically heated L-mode plasma discharges is investigated with line-averaged plasma density ranging from n_e/n_G = 0.15 to 0.42, where n_G is the Greenwald density. All ion saturation current time series that are sampled in the far scrape-off layer are characterized by large-amplitude burst events. Coefficients of skewness and excess kurtosis of the time series obey a quadratic relationship and their histograms coincide partially upon proper normalization. Histograms of the ion saturation current time series are found to agree well with a prediction of a stochastic model for the particle density fluctuations in scrape-off layer plasmas. The distribution of the waiting times between successive large-amplitude burst events and of the burst amplitudes are approximately described by exponential distributions. The average waiting time and burst amplitude are found to vary weakly with the line-averaged plasma density. Conditional averaging reveals that the radial blob velocity, estimated from floating potential measurements, increases with the normalized burst amplitude in the outboard mid-plane scrape-off layer. For low density discharges, the conditionally averaged waveform of the floating potential associated with large amplitude bursts at the divertor probes has a dipolar shape. In detached divertor conditions the average waveform is random, indicating electrical disconnection of blobs from the sheaths at the divertor targets., 45 pages, 20 figures

The mitigation of divertor heat fluxes is an active topic of investigation on existing tokamaks. One approach uses radiation, both inside and outside the last closed flux surface (LCFS), to convert plasma thermal energy, usually directed towards dedicated plasma facing components, to soft X-ray and ultraviolet radiation, spread over a much larger surface area. Recent enhanced D-α H-mode experiments on Alcator C-Mod varied the ICRF input power and radiative power losses via impurity seeding to demonstrate that normalized energy confinement depends strongly on the difference between input power and the radiated power inside the LCFS. These investigations also show that when seeded with either Ne or N2, a factor of two and higher reduction in outer divertor heat flux is achieved while maintaining H98,y2 ∼ 1.0. Conversely, when seeding with Ar, confinement is limited to H98,y2 ∼ 0.8 for a similar level of exhaust power.

New data from Alcator C-Mod have extended the range of heat flux measurements and scalings to poloidal magnetic fields above ITER-level. Knowledge of how the scrape-off layer heat flux width (λ q) scales with machine parameters is crucial for designing fusion reactors and developing a power exhaust solution. An international database indicated that λ q scaled approximately inversely with the poloidal magnetic field (B p) and had no other significant dependencies. However, reactor-class tokamaks are expected to have at least 50% higher B p than the maximum of that database (0.8 T). Alcator C-Mod has been the only diverted tokamak capable of operating at reactor-level B p, up to ~1.3 T. A major focus of the final experimental campaign on Alcator C-Mod was to characterize λ q over a wide range of conditions, utilizing a unique array of heat flux sensors with unprecedented spatial resolution and heat flux dynamic range. The heat flux width scaling is found to extend up to B p ~ 1.3 T in H-mode. Looking across confinement regimes we find the remarkable result that λ q exhibits a unified dependence on volume-averaged core plasma pressure (). Within a standard deviation of about 20%, the heat flux width in any of the C-Mod plasmas studied (L-, I-, and H-mode) is proportional to the inverse square root of . It is also found that the standard prescription of representing the target plate heat flux profile as a convolution of exponential and Gaussian functions does not capture the heat flux profile measured in the private zone; a purely exponential decay fits the data better in this region to >3 orders of magnitude in heat flux dynamic range.

The heat flux distributions on divertor targets in H-mode plasmas are serious concerns for future devices. We seek to simulate the tokamak boundary plasma turbulence and heat transport in the edge localized mode-suppressed regimes. The improved BOUT++ model shows that not only Ip but also the radial electric field Er plays an important role on the turbulence behavior and sets the heat flux width. Instead of calculating Er from the pressure gradient term (diamagnetic Er), it is calculated from the plasma transport equations with the sheath potential in the scrape-off layer and the plasma density and temperature profiles inside the separatrix from the experiment. The simulation results with the new Er model have better agreement with the experiment than using the diamagnetic Er model: (1) The electromagnetic turbulence in enhanced Dα H-mode shows the characteristics of quasi-coherent modes (QCMs) and broadband turbulence. The mode spectra are in agreement with the phase contrast imaging data and almost has no change in comparison to the cases which use the diamagnetic Er model; (2) the self-consistent boundary Er is needed for the turbulence simulations to get the consistent heat flux width with the experiment; (3) the frequencies of the QCMs are proportional to Er, while the divertor heat flux widths are inversely proportional to Er; and (4) the BOUT++ turbulence simulations yield a similar heat flux width to the experimental Eich scaling law and the prediction from the Goldston heuristic drift model.

High-pressure noble gas jet injection is a mitigation technique which potentially satisfies the requirements of fast response time and reliability, without degrading subsequent discharges. Previously reported gas jet experiments on DIII-D showed good success at reducing deleterious disruption effects. In this paper, results of recent gas jet disruption mitigation experiments on Alcator C-Mod and DIII-D are reported. Jointly, these experiments have greatly improved the understanding of gas jet dynamics and the processes involved in mitigating disruption effects. In both machines, the sequence of events following gas injection is observed to be quite similar: the jet neutrals stop near the plasma edge, the edge temperature collapses and large MHD modes are quickly destabilized, mixing the hot plasma core with the edge impurity ions and radiating away the plasma thermal energy. High radiated power fractions are achieved, thus reducing the conducted heat loads to the chamber walls and divertor. A significant (2 × or more) reduction in halo current is also observed. Runaway electron generation is small or absent. These similar results in two quite different tokamaks are encouraging for the applicability of this disruption mitigation technique to ITER.

Mechanisms determining the structure of edge temperature and density pedestals, which are associated with edge transport barrier (ETB) formation in tokamaks, are investigated on Alcator C-Mod. Experiments suggest a strong role for critical gradient behaviour in setting profile characteristics of edge plasma. The maximum pressure gradient scales as the square of plasma current in both H-modes without edge-localized modes, and in the near scrape-off layer in Ohmic discharges. In either case, the pressure gradient obtained, normalized to the square of plasma current, is a function of local collisionality, hinting that common physics may contribute to setting profile gradients in both confinement regimes. Varying the neutral fuelling source has little effect on density gradient scale lengths in the ETB and a relatively weak impact on the height of the density pedestal, even during aggressive deuterium puffing. Strong screening of neutrals in the ETB are observed, creating a challenge for fuelling H-modes in C-Mod beyond their 'natural' density. A simple pedestal fuelling model does not reproduce the typically clamped density gradients seen in experiment during H-mode puffing. These results suggest that a simple diffusive model for plasma transport is deficient, and that a critical gradient assumption for transport may be essential for pedestal modelling.

Theoretical and computational investigations of boundary-plasma microturbulence which take into account important effects of the geometry of diverted tokamaks—in particular, the effect of X-point magnetic shear and the termination of field lines on divertor plates—are presented. We first generalize our previous 'heuristic boundary condition' which describes, in a lumped model, the closure of currents in the vicinity of the X-point region to encompass three current-closure mechanisms. We then use this boundary condition to derive the dispersion relation for low-beta flute-like modes in the divertor-leg region under the combined drives of curvature, sheath impedance and divertor tilt effects. The results indicate the possibility of strongly growing instabilities, driven by sheath boundary conditions, and localized in either the private or common flux region of the divertor leg depending on the radial tilt of divertor plates. We revisit the issue of X-point effects on blobs, examining the transition from blobs terminated by X-point shear to blobs that extend over both the main SOL and divertor legs. We find that, for a main-SOL blob, this transition occurs without a free-acceleration period as previously thought, with X-point termination conditions applying until the blob has expanded to reach the divertor plate. We also derive propagation speeds for divertor-leg blobs. Finally, we present fluid simulations of the C-Mod tokamak from the BOUT edge fluid turbulence code, which show main-SOL blob structures with similar spatial characteristics to those observed in the experiment, and also simulations which illustrate the possibility of fluctuations confined to divertor legs.

The range of ‘edge plasma states’ (i.e., the combination of densities, temperatures and gradients in the scrape-off layer near the separatrix) accessible to a tokamak are highly restricted. Experiments in Alcator C-Mod suggest that this restriction is a consequence of electromagnetic fluid drift turbulence (EMFDT), which sets the cross-field transport levels; plasma states are constrained to lie in a narrow region of EMFDT ‘phase-space’ defined by poloidal beta gradient ( α MHD ) and normalized plasma collisionality. Recent experiments have investigated this behavior over an increased parameter range in lower- and upper-null (LSN/USN) discharges. L-mode edge pressure gradients are found to exhibit clear sensitivity to magnetic topology, favoring higher α MHD for LSN compared to USN. These variations may be caused, in part, by the different edge plasma flow patterns in LSN versus USN. Such flows arise from a ballooning-like transport drive and can produce a topology-dependent modulation in the toroidal plasma rotation.

We report the results of operation of Alcator C-Mod with all high- Z molybdenum plasma facing component (PFC) surfaces. Without boron-coated PFCs energy confinement was poor ( H ITER,89 ∼ 1) due to high core molybdenum ( n Mo / n e ⩽ 0.1%) and radiation. After applying boron coatings, n Mo / n e was reduced by a factor of 10–20 with H ITER,89 approaching 2. Results of between-discharge boronization, localized at various major radii, point towards important molybdenum source regions being small, outside the divertor, and due to RF-sheath-rectification. Boronization also has a significant effect on the plasma startup phase lowering Z eff , radiation, and lowering the runaway electron damage. The requirement of low- Z coatings over at least a fraction of the Mo PFCs in C-Mod for best performance together with the larger than expected D retention in Mo, give impetus for further high- Z PFC investigations to better predict the performance of un-coated tungsten surfaces in ITER and beyond.

This paper reviews the physics and technology of wave-particle-interaction experiments in the ion cyclotron range of frequencies (ICRF) and the lower hybrid (LH) range of frequencies (LHRF) on the ...

The high neutral densities and short neutral mean-free-paths in theAlcator C-Mod divertor have provided a unique testing ground for our understanding of the role of neutrals in a tokamak. The high neutral pressures found in the C-Mod divertor can be reproduced in models only by including such processes as ion-neutral and neutral-neutral collisions and neutral viscosity, as well as taking into account the plasma in the private flux region. After detachment, when the divertor plate ion flux has dropped by more than an order of magnitude, the divertor pressure still remains high. High neutral collisionality and the plasma in the private flux region again help keep neutrals in the divertor along with the large source of neutrals due to recombination. Likewise, diffusive neutrals are the explanation for the divertor neutral pressure's insensitivity to strike point position. Closure of neutral leakage pathways did not lead to a decrease in neutral pressures in the region outside the divertor-the main chamber. This observation prompted further research, which showed that ion fluxes to main chamber surfaces rival those reaching the divertor plates; the main chamber pressure can be primarily determined by the level of ion transport perpendicular to the magnetic field. This finding has spawned a host of studies (active and passive) both at C-Mod and other tokamaks to understand how radial transport can be so large.

An overview of the diagnostics installed on the Alcator C-Mod tokamak is presented. Approximately 25 diagnostic systems are being operated on C-Mod. The compact design of the machine and the cryostat enclosing the vacuum vessel and magnetic field coils make access challenging. Diagnostics are used to study four focus areas: transport, plasma boundary, waves, and macrostability. There is significant overlap between these topics, and they all contribute toward the burning plasma and advanced tokamak thrusts. Several advanced and novel diagnostics contribute to the investigation of C-Mod plasmas, e.g., electron cyclotron emission, phase-contrast imaging, gas puff imaging, probe measurements, and active magnetohydrodynamic antennas.

Global and local transport experiments in ohmic, L-mode and H-mode regimes on the Alcator C-Mod tokamak are summarized. For ohmic plasmas, earlier results derived for energy confinement scaling in ...

© 2015 IOP Publishing Ltd. We carry out a quantitative comparison between gas puff imaging (GPI) turbulence measurements in Alcator C Mod inner wall limited discharges (Zweben et al 2009 Phys. Plasmas 18 082505) and 3D flux driven drift reduced Braginskii turbulence simulations of scrape off layer dynamics. The comparison is carried out for a series of inner wall limited discharges where the magnetic field and the density are varied. The comparison between GPI data and non linear simulations yields overall good agreement for several observables such as the Da emission levels and intermittency the radial and poloidal correlation lengths and propagation velocities and the power and frequency spectral density.

The guiding-center kinetic neoclassical transport code, XGC0, [C.S. Chang et. al, Phys. Plasmas 11, 2649 (2004)] is used to compute the heat fluxes and the heat-load width in the outer divertor plates of Alcator C-Mod and DIII-D tokamaks. The dependence of the width of heat-load fluxes on neoclassical effects, neutral collisions and anomalous transport is investigated using the XGC0 code. The XGC0 code includes realistic X-point geometry, a neutral source model, the effects of collisions, and a diffusion model for anomalous transport. It is observed that width of the XGC0 neoclassical heat-load is approximately inversely proportional to the total plasma current $I_{\rm p}$. The scaling of the width of the divertor heat-load with plasma current is examined for an Alcator C-Mod discharge and four DIII-D discharges. The scaling of the divertor heat-load width with plasma current is found to be weaker in the Alcator C-Mod discharge compared to scaling found in the DIII-D discharges. The effect of neutral collisions on the $1/I_{\rm p}$ scaling of heat-load width is shown not to be significant. Although inclusion of poloidally uniform anomalous transport results in a deviation from the $1/I_{\rm p}$ scaling, the inclusion of the anomalous transport that is driven by ballooning-type instabilities results in recovering the neoclassical $1/I_{\rm p}$ scaling. The Bohm or Gyro-Bohm scalings of anomalous transport does not strongly affect the dependence of the heat-load width on plasma current. The inclusion of anomalous transport, in general, results in widening the width of neoclassical divertor heat-load and enhances the neoclassical heat-load fluxes on the divertor plates. Understanding heat transport in the tokamak scrape-off layer plasmas is important for strengthening the basis for predicting divertor conditions in ITER., Comment: Submitted to Physics of Plasmas

he inboard limiters for ITER were initially designed on the assumption that the parallel heat flux density in the scrape-off layer (SOL) could be approximated by a single exponential with decay length λq. This assumption was found not to be adequate in 2012, when infra-red (IR) thermography measurements on the inner column during JET limiter discharges clearly revealed the presence of a narrow heat flux channel adjacent to the last closed flux surface. This near-SOL decay occurs with λq ~ few mm, much shorter than the main SOL λq, and can raise the heat flux at the limiter apex a factor up to ~4 above the value expected from a single, broader exponential. The original logarithmically shaped ITER inner wall first wall panels (FWPs) would be unsuited to handling the power loads produced by such a narrow feature. A multi-machine study involving the C-Mod, COMPASS, DIII-D and TCV tokamaks, employing inner wall IR measurements and/or inner wall reciprocating probes, was initiated to investigate the narrow limiter SOL heat flux channel. This paper describes the new results which have provided an experimental database for the narrow feature and presents an ITER inner wall FWP toroidal shape optimized for a double-exponential profile with λq = 4 (narrow feature) and 50 mm (main-SOL), the latter also derived from a separate multi-machine database constituted recently within the International Tokamak Physics Activity. It is shown that the new shape allows the power handling capability of the original shape design to be completely recovered for a wide variety of limiter start-up equilibria in the presence of a narrow feature, even taking assembly tolerances into account. It is, moreover, further shown that the new shape has the interesting property of both mitigating the impact of the narrow feature and resulting in only a very modest increase in heat load, compared to the current design, if the narrow feature is not eventually found on ITER.

Damaging effects of disruptions are a major concern for Alcator C-Mod, ITER and future tokamak reactors. High-pressure noble gas jet injection is a mitigation technique which potentially satisfies the operational requirements of fast response time and reliability, while still being benign to subsequent discharges. Disruption mitigation experiments using an optimized gas jet injection system are being carried out on Alcator C-Mod to study the physics of gas jet penetration into high pressure plasmas, as well as the ability of the gas jet impurities to convert plasma energy into radiation on timescales consistent with C-Mod's fast quench times, and to reduce halo currents given C-Mod's high-current density. The dependence of impurity penetration and effectiveness on noble gas species (He, Ne, Ar, Kr) is also being studied. It is found that the high-pressure neutral gas jet does not penetrate deeply into the C-Mod plasma, and yet prompt core thermal quenches are observed on all gas jet shots. 3D MHD modelling of the disruption physics with NIMROD shows that edge cooling of the plasma triggers fast growing tearing modes which rapidly produce a stochastic region in the core of the plasma and loss of thermal energy. This may explain the apparent effectiveness of the gas jet in C-Mod despite its limited penetration. The higher-Z gases (Ne, Ar, Kr) also proved effective at reducing halo currents and decreasing thermal deposition to the divertor surfaces. In addition, noble gas jet injection proved to be benign for plasma operation with C-Mod's metal (Mo) wall, actually improving the reliability of the startup in the following discharge.

Comparisons of H-mode regimes were carried out on the Alcator C-Mod and JFT-2M tokamaks. Shapes were matched apart from aspect ratio, which is lower on C-Mod. The high recycling steady H-mode on JFT-2M and enhanced D-alpha (EDA) regime on C-Mod, both of which feature very small or no ELMs, are found to have similar access conditions in q 95 - v* space, occurring for pedestal collisionality v* ? 1. Differences in edge fluctuations were found, with lower frequencies but higher mode numbers on C-Mod. In both tokamaks an attractive regime with small ELMs on top of an enhanced D α baseline was obtained at moderate v* and higher pressure. The JFT-2M shape favoured the appearance of ELMs on C-Mod and also resulted in the appearance of a lower frequency component of the quasicoherent mode during EDA.

Core impurity transport has been investigated for a variety of confinement regimes in Alcator C-Mod plasmas from x-ray emission following injection of medium and high Z materials. In ohmic L-mode discharges, impurity transport is anomalous (D[subscript eff] ≫ D[subscript nc]) and changes very little across the LOC/SOC boundary. In ion cyclotron range of frequencies (ICRF) heated L-mode plasmas, the core impurity confinement time decreases with increasing ICRF input power (and subsequent increasing electron temperature) and increases with plasma current. Nearly identical impurity confinement characteristics are observed in I-mode plasmas. In enhanced D[subscript α] H-mode discharges the core impurity confinement times are much longer. There is a strong connection between core impurity confinement time and the edge density gradient across all confinement regimes studied here. Deduced central impurity density profiles in stationary plasmas are generally flat, in spite of large amplitude sawtooth oscillations, and there is little evidence of impurity convection inside of r/a = 0.3 when averaged over sawteeth., United States. Department of Energy (Contract DE-FC02-99ER54512), United States. Dept. of Energy. Fusion Energy Postdoctoral Research Program (Oak Ridge Institute for Science and Education)

Plasma profiles across the separatrix and scrape-off layer (SOL) in Alcator C-Mod are examined for a range of plasma densities, currents and magnetic fields in Ohmic L-mode discharges and for a subset of conditions in Ohmic H-mode discharges. In all plasmas, electron pressure gradient scale lengths ( ) exhibit a minimum value just outside the separatrix (i.e. in the near SOL), forming the base of a weak (strong) pedestal in L-mode (H-mode) plasmas. Over a wide range of conditions in Ohmic L-mode discharges, at this location are found to track with a monotonic function of electron collision frequency, when this quantity is normalized according to the framework of electromagnetic fluid drift turbulence (EMFDT) theory. Moreover, at fixed values of normalized collisionality parameter (characterized as the 'diamagnetic parameter', ?d), electron pressure gradients in the near SOL increase with plasma current squared, holding the MHD ballooning parameter, ?MHD, unchanged. Thus, the state of the near SOL is restricted to a narrow region within this two-parameter phase-space. An implication is that cross-field heat and particle transport are strong functions of these parameters. Indeed, as ?d is decreased below ~0.3, cross-field heat convection increases sharply and competes with parallel heat conduction along open field lines, making high plasma density regions of ?MHD??d space energetically inaccessible. These observations are consistent with the idea that the operational space of the edge plasma, including boundaries associated with the tokamak density limit, is controlled by EMFDT.

Research on the Alcator C-Mod tokamak has emphasized RF heating, self-generated flows, momentum transport, scrape-off layer (SOL) turbulence and transport and the physics of transport barrier transitions, stability and control. The machine operates with P-RF up to 6 MW corresponding to power densities on the antenna of 10 MW m(-2). Analysis of rotation profile evolution, produced in the absence of external drive, allows transport of angular momentum in the plasma core to be computed and compared between various operating regimes. Momentum is clearly seen diffusing and convecting from the plasma edge on time scales similar to the energy confinement time and much faster than neo-classical transport. SOL turbulence and transport have been studied with fast scanning electrostatic probes situated at several poloidal locations and with gas puff imaging. Strong poloidal asymmetries are found in profiles and fluctuations, confirming the essential ballooning character of the turbulence and transport. Plasma topology has a dominant effect on the magnitude and direction of both core rotation and SOL flows. The correlation of self-generated plasma flows and topology has led to a novel explanation for the dependence of the H-mode power threshold on the del B drift direction. Research into internal transport barriers has focused on control of the barrier strength and location. The foot of the barrier could be moved to larger minor radius by lowering q or B-T. The barriers, which are produced in C-Mod by off-axis RF heating, can be weakened by the application of on-axis power. Gyro-kinetic simulations suggest that the control mechanism is due to the temperature dependence of trapped electron modes which are destabilized by the large density gradients. A set of non-axisymmetric coils was installed allowing intrinsic error fields to be measured and compensated. These also enabled the determination of the mode locking threshold and, by comparison with data from other machines, provided the first direct measurement of size scaling for the threshold. The installation of a new inboard limiter resulted in the reduction of halo currents following disruptions. This effect can be understood in terms of the change in plasma contact with the altered geometry during vertical displacement of the plasma column. Unstable Alfven eigenmodes (AE) were observed in low-density, high-power ICRF heated plasmas. The damping rate of stable AEs was investigated with a pair of active MHD antennae.

Measurements of toroidal flow velocities in the midplane of the inner and outer scrape-off layers (SOL) of Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] plasmas are made using a high-resolution spectrograph. Doppler shifts of the line emissions from either He+ at 468.6 nm or neutral deuterium at 656.1 nm from two local gas puffs are observed along radially separated, toroidally viewing chords to obtain the radial profiles of the toroidal ion and neutral flow velocities and of the temperatures in the midplane of the inner and outer SOL’s. In these experiments, the ion vertical drift due to the toroidal magnetic field gradient (ion B×∇B drift) is directed towards the lower X-point, and an outer SOL plasma flow with a velocity in the range of 5–10 km/s in the direction of the plasma current is observed at the midplane far above the divertor plates. Plasma flow in a similar direction with similar velocities is observed in the inner SOL midplane. This observation suggests a toroidal rot...

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.

Impurity retention is a key role of a divertor in a tokamak. One of the measurements of retention performance is the impurity compression ratio c z , defined as the ratio of the impurity neutral density in the divertor to the impurity ion density in the core plasma. DIVIMP Monte-Carlo modeling of argon recycling impurity compression in C-Mod has been performed to explore the physics behind impurity compression. It is found that the frictional force is dominant over the temperature gradient forces in determining the impurity compression in C-Mod divertor.

The gas puff imaging (GPI) diagnostic can be used to study the turbulence present at the edge of magnetically confined plasmas. In this diagnostic the instantaneous two-dimensional (2D) radial vs poloidal structure of the turbulence is measured using fast-gated cameras and discrete fast chords. By imaging a controlled neutral gas puff, of typically helium or deuterium, the brightness and contrast of the turbulent emission fluctuations are increased and the structure can be measured independently of natural gas recycling. In addition, recent advances in ultrafast framing cameras allow the turbulence to be followed in time. The gas puff itself does not perturb the edge turbulence and the neutral gas does not introduce fluctuations in the emission that could possibly arise from a nonsmooth (turbulent) neutral gas puff. Results from neutral transport and atomic physics simulations using the DEGAS 2 code are discussed showing that the observed line emission is sensitive to modulations in both the electron density and the electron temperature. The GPI diagnostic implementation in the National Spherical Torus Experiment (NSTX) and Alcator C-Mod tokamak is presented together with example results from these two experiments.

The EIRENE Monte-Carlo neutral code is used to calculate neutral pressures throughout the divertor region of the Alcator C-Mod tokamak. A semi-empirical onion-skin method (OSM) is used to calculate the hydrogenic background plasma for a moderate density discharge, 〈 n e 〉=1.5×10 20 m −3 , with a partially detached inner target and an attached outer target. The model parameters are adjusted until there is agreement with the Langmuir probe and spectroscopic data for the divertor region. The measured divertor pressure is 30 mTorr, and for C-Mod divertor dimensions this implies the transition flow regime between free molecular and viscous neutral transport. Therefore, a non-linear BGK-model Boltzmann collision term is included in EIRENE to allow for interactions between neutral particles. For the standard model parameters, the calculated divertor pressure is 7 mTorr. The model parameters are adjusted to reduce the discrepancy, and the plausibility of each variation is discussed.

Simultaneous high resolution measurements of the soft x-ray emission at the top and outboard edge of the Alcator C-Mod tokamak [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] are presented. Impurity density profiles are derived from these measurements. The impurity density varies considerably on a flux surface in the high confinement mode (H-mode) pedestal region. The soft x-ray and impurity density pedestals are not on the same flux surface at the top and outboard edge. The pedestal widths are consistently larger at the outboard edge in the Enhanced D-alpha (EDA) H-mode, suggesting a ballooning-like character of the EDA quasi-coherent mode. The usual separation of time scales between radial and parallel impurity transport does not hold in the H-mode pedestal region of Alcator C-Mod. Thus, one should not necessarily expect that the impurity density be constant on a flux surface in this region.

New results on the I-mode regime of operation on the Alcator C-Mod tokamak are reported. This ELM-free regime features high energy confinement and a steep temperature pedestal, while particle confinement remains at L-mode levels, giving stationary density and avoiding impurity accumulation. I-mode has now been obtained over nearly all of the magnetic fields and currents possible in this high field tokamak (I-p 0.55-1.7 MA, B-T 2.8-8 T) using a configuration with B x del B drift away from the X-point. Results at 8 T confirm that the L-I power threshold varies only weakly with B-T, and that the power range for I-mode increases with B-T; no 8 T discharges transitioned to H-mode. Parameter dependences of energy confinement are investigated. Core transport simulations are giving insight into the observed turbulence reduction, profile stiffness and confinement improvement. Pedestal models explain the observed stability to ELMs, and can simulate the observed weakly coherent mode. Conditions for I-H transitions have complex dependences on density as well as power. I-modes have now been maintained in near-DN configurations, leading to improved divertor power flux sharing. Prospects for I-mode on future fusion devices such as ITER and ARC are encouraging. Further experiments on other tokamaks are needed to improve confidence in extrapolation.

Gas-puff imaging techniques are employed to determine the far SOL region radial electric field and the plasma potential in ICRF heated discharges in the Alcator C-Mod tokamak. The 2-dimensional velocity fields of the turbulent structures, which are advected by RF-induced $ \mathbf{E\times B} $ flows, are obtained via the time-delay estimation (TDE) techniques. Both the magnitude and radial extension of the radial electric field $ E_r $ are observed to increase with the toroidal magnetic field strength $ B_\varphi $ and the ICRF power. In particular, the RF-induced $ E_r $ extends from the vicinity of the ICRF antenna to the separatrix when $ B_\varphi=7.9\,\mathrm{T} $ and $ P_{\mathrm{ICRF}} \gtrsim 1\,\mathrm{MW} $. In addition, low-Z impurity seeding near the antenna is found to substantially reduce the sheath potential associated with ICRF power. The ICRF-induced potentials are also estimated in different antenna configurations: (1) conventional toroidally-aligned (TA) antenna versus field-aligned (FA) antenna; (2) FA monopole versus FA dipole. Results show that FA and TA antennas produce similar magnitude of plasma potentials, and the FA monopole induced greater potential than the FA dipole phasing.

The impurity screening response of the high-field side (HFS) scrape-off layer (SOL) to localized nitrogen injection is investigated on Alcator C-Mod for magnetic equilibria spanning lower-single-null, double-null and upper-single-null configurations under otherwise identical plasma conditions. L-mode, EDA H-mode and I-mode discharges are investigated. HFS impurity screening is found to depend on magnetic flux balance and the direction of B??×?? B relative to the most active divertor. Impurity 'plume' emission patterns indicate that both parallel and perpendicular (E??×??B) flows in the SOL contribute to the 'flushing' of impurities towards the active divertor, thereby affecting the overall impurity screening behavior. Despite the fact that the HFS SOL is extremely narrow in near-double-null configurations, this SOL is able to screen locally injected nitrogen at least as effectively as the low-field side (LFS) SOL?up to a factor of 10 more effective, depending on specific plasma conditions and whether the magnetic geometry produces parallel flows that work with or against E??×??B flows. For situations in which the E??×??B drift of the impurity ions opposes parallel flow toward the primary divertor, HFS impurity screening is found to be least effective. When E??×??B drifts assist parallel flow toward the primary divertor, HFS impurity screening is found to be very effective. These data support the idea of placing RF actuators and close-fitting wall components on the HFS of the tokamak. With this configuration, near-double-null magnetic topologies may be used for active control of plasma parameters at the antenna/plasma interface for optimal RF coupling, mitigate the generation of local impurities from plasma?material interactions and, taking advantage of favorable plasma flows and good screening properties of the HFS SOL, further minimize the impact of wall-born impurity sources on the plasma core.

A computational study of long-legged tokamak divertor configurations is performed with the edge transport code UEDGE. Several divertor configurations are considered, with radially or vertically extended, tightly baffled, outer divertor legs and with or without a secondary X-point in the divertor leg volume. For otherwise identical conditions, a scan of the input power from the core plasma is performed. As the power is reduced to a threshold value, the plasma in the outer leg transitions to a fully detached state, which defines the upper limit on the power for detached divertor operation. Reducing the power further results in the detachment front shifting upstream but remains stable. At low power, the detachment front eventually moves all the way to the primary X-point, which is usually associated with degradation of the core plasma, and this defines the lower limit on the power for the detached divertor operation. For the studied parameters, for long-legged divertors, the detached operation window is quit...

Analysis of the time series obtained with the O-Mode reflectometer [Rhodes et al., 1997 Plasma Phys. Controlled Fusion 40, 493-510 (1998)] and the gas puff imaging [Cziegler et al., Phys. Plasmas 17(5), 056120 (2010)] systems on the Alcator C-Mod tokamak reveals that the turbulent edge density fluctuations are chaotic. Supporting evidence for this conclusion includes the observation of exponential power spectra (which is associated with Lorentzian-shaped pulses in the time series), the population of the corresponding Bandt-Pompe probability distributions [Bandt and Pompe, Phys. Rev. Lett. 88, 174102 (2002)], and the location of the signal on the complexity-entropy plane (C-H plane) [Rosso et al., Phys. Rev. Lett. 99, 154102 (2007)]. The classification of edge turbulence as chaotic opens the door for further work to understand the underlying process and the impact on turbulent transport., United States. Department of Energy (Grant DE-FC02-99ER54512), United States. Department of Energy (Grant DE-FC02-07ER54918-011)

Dedicated experiments in ion cyclotron range heated enhanced D-alpha (EDA) H-mode and I-mode plasmas have been performed on Alcator C-Mod to identify the location of edge fluctuations inside the pedestal and to determine their plasma frame phase velocity. For this purpose, measurements from gas puff imaging (GPI) and gas puff charge exchange recombination spectroscopy (GP-CXRS) have been collected using the same optical views. The data suggest that the EDA H-mode-specific quasi-coherent mode (QCM) is centered near the radial electric field (E r) well minimum and propagates along the ion diamagnetic drift direction in the plasma frame. The weakly coherent mode (WCM) and the geodesic acoustic mode observed in I-mode, on the other hand, are found to be located around the outer shear layer of the E r well. This results in a weak plasma frame phase velocity mostly along the electron diamagnetic drift direction for the WCM. The findings in these EDA H-mode plasmas differ from probe measurements in ohmic EDA H-mode (LaBombard et al 2014 Phys. Plasmas 21 056108), where the QCM was identified as an electron drift-wave located several mm outside the E r well minimum in a region of positive E r. To explore if instrumental effects of the optical diagnostics could be the cause of the difference, a synthetic diagnostic for GPI is introduced. This diagnostic reproduces amplitude ratios and relative radial shifts of the mode profiles determined from poloidally and toroidally oriented optics and, if instrumental effects related to GP-CXRS are also included, indicates that the measured location of the QCM and WCM relative to the E r well reported here is only weakly affected by instrumental effects.

The effect of divertor baffling on impurity neutral and impurity ion compression in ohmic discharges in the Alcator C-Mod tokamak is explored. Experiments are performed using a novel divertor bypass which allows in situ variations to the divertor baffling. The results indicate that divertor baffling in Alcator C-Mod plays an important role in impurity neutral compression, but little role in impurity ion compression. The apparent separation between neutral compression and ion compression is surprising, but is consistent with a model in which the neutral compression (both in deuterium and argon) is determined by a leakage flux through the baffle structure from the divertor region to the main chamber which is independent of the leakage conductance. Instead, the leakage fluxes appear to be determined by the divertor plasma conditions (including impurity and fuel ions), which are not influenced by the baffle structure.

Research on the Alcator C-Mod tokamak is focused on exploiting compact high density plasmas to understand core transport and heating, the physics of the H mode transport barrier, and the dynamics of the scrape-off layer and divertor. Rapid toroidal acceleration of the plasma core is observed during ohmic heated H modes and indicates a momentum pinch or similar transport mechanism. Core thermal transport observations support a critical gradient interpretation, but with gradients that disagree with present theoretical values. High resolution measurements of the H mode barrier have been obtained, including impurity and neutral densities, and the instability apparently responsible for the favourable `enhanced D alpha' regime has been identified. Divertor bypass dynamic control experiments have directly addressed the important questions surrounding main chamber recycling and the effect of divertor closure on impurities and confinement. Future plans include quasi-steady-state advanced tokamak plasmas using lower hybrid current drive.

Magnetically-broadened hydrogenic line profiles are used, in conjunction with a non-local thermodynamic equilibrium (NLTE) code, to model an optically thick phenomenon in the Alcator C-Mod tokamak fusion experiment. The line profiles, including magnetic field effects as well as Doppler and Stark broadening, have been calculated using Beline and are used in Cretin, a multi-dimensional NLTE radiation transfer code, to study the influence of a magnetic field on emergent spectrum. A midplane Alcator C-Mod multifaceted axisymmetric radiation from the edge (MARFE), in which the magnetic field strength was changed dynamically from 5.4 T down to 2.7 T , is simulated and compared with experimental data. Assuming hydrodynamic and radiative equilibrium, MARFE simulations are performed in one-dimension with fixed plasma density and temperature profiles. Magnetic broadening effects on the brightness of optically thick Lyman series lines are quantified.

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.