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Halo currents generated during disruptions on Alcator C-Mod have been measured with Langmuir "rail" probes. These rail probes are embedded in a lower outboard divertor module in a closely-spaced vertical (poloidal) array. The dense array provides detailed resolution of the spatial dependence (~1 cm spacing) of the halo current distribution in the plasma scrape-off region with high time resolution (400 kHz digitization rate). As the plasma limits on the outboard divertor plate, the contact point is clearly discernible in the halo current data (as an inversion of current) and moves vertically down the divertor plate on many disruptions. These data are consistent with filament reconstructions of the plasma boundary, from which the edge safety factor of the disrupting plasma can be calculated. Additionally, the halo current "footprint" on the divertor plate is obtained and related to the halo flux width. The voltage driving halo current and the effective resistance of the plasma region through which the halo current flows to reach the probes are also investigated. Estimations of the sheath resistance and halo region resistivity and temperature are given. This information could prove useful for modeling halo current dynamics.

The next generation outer divertor target proposed for C-Mod is intended to operate with edge physics behavior that is 'Demo-like i.e., it will be capable of operating at a bulk tile and structure temperature of 600C. The proposed design exposes a vertical cylinder covered with tungsten lamellae tiles to the divertor heat flux. Heat load variation along the height of the cylindrical target has been specified and is being considered in the tile design. The design must allow for differential radial thermal expansion of the cylindrical structure. It is intended to be toroidally continuous with a high tolerance on axisymmetry to improve alignment with the plasma and limit interactions of disruption induced currents with the toroidal field. Inductively driven axisymmetric disruption currents are calculated using electromagnetic transient simulations previously employed for RF antennas and the cryopump. Disruption-induced halo currents are expected to flow though the structure, which have proved troublesome for the old outer divertor structure. The new toroidally continuous structure will be intrinsically strong with respect to axisymmetric mechanical loads, although the support hardware will also need to be robust to resist movement during non-axisymmetric halo loads. Halo current specifications for the outer divertor have been developed, and halo current paths that minimize loading are "forced" with appropriate use of insulation and grounding straps. Radiative energy transfer to other components in the vessel makes sustained operation of the outer divertor at elevated temperatures difficult.

In Alcator C-Mod, we have sought to identify ICRF specific impurity sources and estimate the lifetime of thin, low-Z films in the presence of high power ICRF heating. With ICRF, the erosion rate of low-Z films is estimated to be 15–20 nm/s indicating the eroding species energy is much higher than that normally found in the SOL. Using emissive probes, plasma potential measurements confirm the presence of an enhanced sheath with ICRF when the probe is magnetically linked to the active antenna. Plasma potentials were typically 100–200 V for ∼1.25 MW injected ICRF power. Furthermore, an RF sheath was unexpectedly present with insulating limiters and was dependent on the wall conditions and plasma confinement. These dependencies suggest that other scrape-off layer and plasma-surface characteristics are influencing the resulting RF sheaths.

Movies of edge turbulence at both the outboard midplane and the region outboard of the typical lower X-point location in C-Mod have been obtained using gas-puff-imaging together with fast-framing cameras. Intermittent turbulence structures, typically referred to as blobs or filaments, are observed in both locations. Near the midplane the filaments are roughly circular in cross-section, while in the X-point region they are highly elongated. Filament velocities in this region are ∼3x faster than the radial velocities at the midplane, in a direction roughly outward across the local flux surfaces. The observations are consistent with the picture that the filaments arise in the outboard region and, as a consequence of the rapid parallel diffusion of the potential perturbations, map along field lines. Results from a 3D BOUT turbulence simulation reproduce many of the spatial features observed in the experiment.

Near-sonic parallel plasma flows in the tokamak scrape-off layer (SOL) are modeled with two-dimensional multi-fluid edge plasma transport code UEDGE. As shown, the dominant mechanism driving such flows is fast non-diffusive intermittent cross-field transport (due to blobs) which has strong ballooning-like poloidal asymmetry. The properties of asymmetric-transport-driven flows, recycling, and divertor detachment are studied. The conditions for zonal and circular flow patterns in the SOL are analyzed. The impact of plasma drifts on non-diffusive transport driven flows is modeled and discussed. (© 2008 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

In this chapter we review numerous diagnostics capable of measurements at or near the first wall, many of which contribute information useful for safe operation of a tokamak. Infrared cameras, visible and vacuum ultraviolet cameras, pressure gauges and residual gas analyzers, thermocouples, and erosion and deposition measurements by insertable probes, quartz microbalances, and a rather extensive review of Langmuir probes are discussed. Also discussed are dust measurements by electrostatic detectors, laser scattering, visible and infrared cameras, and manual collection of samples after machine opening. In each case the diagnostic is discussed with a view toward application to a burning plasma machine such as ITER.

Recent research in scrape-off layer (SOL) and divertor physics is reviewed; new and existing data from a variety of experiments have been used to make cross-experiment comparisons with implications for further research and ITER. Studies of the region near the separatrix have addressed the relationship of profiles to turbulence as well as the scaling of the parallel power flow. Enhanced low-field side radial transport is implicated as driving parallel flows to the inboard side. The medium-n nature of edge localized modes (ELMs) has been elucidated and new measurements have determined that they carry ~10?20% of the ELM energy to the far SOL with implications for ITER limiters and the upper divertor. The predicted divertor power loads for ITER disruptions are reduced while those to main chamber plasma facing components (PFCs) increase. Disruption mitigation through massive gas puffing is successful at reducing PFC heat loads. New estimates of ITER tritium retention have shown tile sides to play a significant role; tritium cleanup may be necessary every few days to weeks. ITER's use of mixed materials gives rise to a reduction of surface melting temperatures and chemical sputtering. Advances in modelling of the ITER divertor and flows have enhanced the capability to match experimental data and predict ITER performance.

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.

This paper reviews measurements of edge plasma turbulence in toroidal magnetic fusion devices with an emphasis on recent results in tokamaks. The dominant feature of edge turbulence is a high level of broadband density fluctuations with a relative amplitude delta n/n - 5-100%, accompanied by large potential and electron temperature. fluctuations. The frequency range of this turbulence is -10 kHz-1 MHz, and the size scale is typically -0.1-10 cm perpendicular to the magnetic field but many metres along the magnetic field, i.e. the structure is nearly that of 2D 'filaments'. Large intermittent bursts or 'blobs' are usually observed in the scrape-off layer. Diagnostic and data analysis techniques are reviewed and the main experimental results are summarized. Recent comparisons of experimental results with edge turbulence theory are discussed, and some directions for future experiments are suggested.

Characteristics of discrete ELMs produced in Alcator C-Mod discharges of low edge collisionality (0.2 4.5 cm. The initial filaments are followed (at intervals of ∼100 μs) by multiple, less perturbing secondary filaments.

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.

Large-Mach-number parallel plasma flows in the single-null SOL of different tokamaks are simulated with multifluid transport code UEDGE. The key role of poloidal asymmetry of cross-field plasma transport as the driving mechanism for such flows is discussed. The impact of ballooning-like diffusive and convective transport and plasma flows on divertor detachment, material migration, impurity flows, and erosion/deposition profiles is studied. The results on well-balanced double null plasma modeling that are indicative of strong asymmetry of cross-field transport are presented.

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.

Many important contributions to the understanding of divertor physics are presented in this review of Alcator C-Mod research. The three regimes of parallel transport, sheath limited, conduction limited, and detached, were identified experimentally, along with their effects on plasma pressure along the magnetic field in the scrape-off layer. The extensive probe and bolometric coverage of the divertor allowed detailed characterization of the physics of detachment. The ability to dissipate ITER-like parallel power densities with extremely high divertor radiation emissivities (>40 MW/m3) was demonstrated under high-recycling and detached divertor conditions. The vertical plate divertor concept, developed and applied first on C-Mod, allowed the effect of divertor geometry to be studied, with the result that the vertical-plate and deep-slot geometries have a lower detachment threshold than the standard, flat-plate divertor. High-density (ne > 1 × 1021 m-3) divertor conditions allowed recombination to be...

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 ...

We describe some of the engineering solutions required to produce a diverted tokamak capable of operation with a toroidal field of 8 T and plasma currents of up to 2 MA. Some design details of the toroidal field magnet, the ohmic heating magnet, the metal plasma-facing components, the rf heating and current drive systems, and the power and liquid nitrogen cooling systems are discussed. Vacuum, vessel bake, boronization, and wall-cleaning systems are also discussed. Finally, disruption research results from Alcator C-Mod are presented.

H-mode research on Alcator C-Mod is described, with a focus on the edge transport barrier (ETB). ETB pedestals are characterized using several diagnostics, leading to a thorough description of prof...

Parametric decay instabilities (PDI) appear to be an ubiquitous feature of lower hybrid current drive (LHCD) experiments at high density. In density ramp experiments in Alcator C-Mod and other machines the onset of PDI activity has been well correlated with a decrease in current drive efficiency and production of fast electron bremsstrahlung. However whether PDI is the primary cause of the 'density limit', and if so by exactly what mechanism (beyond the obvious one of pump depletion) has not been clearly established. In order to further understand the connection, the frequency spectrum of PDI activity occurring during Alcator C-Mod LHCD experiments has been explored in detail by means of a number of RF probes distributed around the periphery of the C-Mod tokamak including a probe imbedded in the inner wall. The results show that (i) the excited spectra consists mainly of a few discrete ion cyclotron (IC) quasi-modes, which have higher growth than the ion sound branch; (ii) PDI activity can begin either at the inner or outer wall, depending on magnetic configuration; (iii) the frequencies of the IC quasi-modes correspond to the magnetic field strength close to the low-field side (LFS) or high-field side separatrix; and (iv) although PDI activity may initiate near the inner separatrix, the loss in fast electron bremsstrahlung is best correlated with the appearance of IC quasi-modes characteristic of the magnetic field strength near the LFS separatrix. These data, supported by growth rate calculations, point to the importance of the LFS scrape-off layer (SOL) density in determining PDI onset and degradation in current drive efficiency. By minimizing the SOL density it is possible to extend the core density regime over which PDI can be avoided, thus potentially maximizing the effectiveness of LHCD at high density. Increased current drive efficiency at high density has been achieved in FTU and EAST through lithium coating and special fuelling methods, and in recent C-Mod experiments by operating at higher plasma current. Another approach would be to locate the launcher in the inner wall with double null operation. This would reduce the SOL density by an order of magnitude or more and greatly mitigate the effects of PDI as well as other parasitic losses., United States. Department of Energy (DE-FC02-99ER54512)

Launching radio frequency (RF) waves from the high field side (HFS) of a tokamak offers significant advantages over low field side (LFS) launch with respect to both wave physics and plasma material interactions (PMI). For lower hybrid (LH) waves, the higher magnetic field opens the window between wave accessibility (n∥≡ck∥/ω>1−ω2pi/ω2+ω2pe/ω2ce−−−−−−−−−−−−−−−−√+ωpe/∣∣ωce∣∣) and the condition for strong electron Landau damping (n∥∼30/Te−−−−−√ with Te in keV), allowing LH waves from the HFS to penetrate into the core of a burning plasma, while waves launched from the LFS are restricted to the periphery of the plasma. The lower n∥ of waves absorbed at higher Te yields a higher current drive efficiency as well. In the ion cyclotron range of frequencies (ICRF), HFS launch allows for direct access to the mode conversion layer where mode converted waves absorb strongly on thermal electrons and ions, thus avoiding the generation of energetic minority ion tails. The absence of turbulent heat and particle fluxes on the HFS, particularly in double null configuration, makes it the ideal location to minimize PMI damage to the antenna structure. The quiescent SOL also eliminates the need to couple LH waves across a long distance to the separatrix, as the antenna can be located close to plasma without risking damage to the structure. Improved impurity screening on the HFS will help eliminate the long-standing issues of high Z impurity accumulation with ICRF. Looking toward a fusion reactor, the HFS is the only possible location for a plasma-facing RF antenna that will survive long-term. By integrating the antenna into the blanket module it is possible to improve the tritium breeding ratio compared with an antenna occupying an equatorial port plug. Blanket modules will require remote handling of numerous cooling pipes and electrical connections, and the addition of transmission lines will not substantially increase the level of complexity. The obvious engineering challenges associated with locating antenna structures on the HFS can be overcome if HFS antennas are incorporated in the overall experimental design from the start. The Advanced Divertor and radio frequency eXperiment(ADX) will include LH and ICRF antennas located on the HFS. Compact antenna designs based on proven technologies (e.g. multi-junction and “4-way splitter” antennas) fit within the available space on the HFS of ADX. Field aligned ICRF antennas are also located on the HFS. The ADX vacuum vessel design includes dedicated space for transmission lines, pressure windows, and vacuum feedthrus for accessing the HFS wall.

© 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 efficiency of LHCD on Alcator C-Mod drops exponentially with line average density. At reactor relevant densities (> 1 · 1020 [m−3]) no measurable current is driven. While a number of causes have been suggested, no specific mechanism has been shown to be responsible for the loss of current drive at high density. Fast modulation of the LH power was used to isolate and quantify the LHCD deposition within the plasma. Measurements from these plasmas provide unique evidence for determining a root cause. Modulation of LH power in steady plasmas exhibited no correlated change in the core temperature. A correlated, prompt response in the edge suggests that the loss in efficiency is related to a edge absorption mechanism. This follows previous results which found the generation of n||-independent SOL currents. Multiple Langmuir probe array measurements of the conducted heat conclude that the lost power is deposited near the last closed flux surface. The heat flux induced by LH waves onto the outer divertor is c...

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.

Experiments on Alcator C-Mod with toroidally and poloidally localized divertor nitrogen injection have been modeled using the three-dimensional edge transport code EMC3-EIRENE to elucidate the mechanisms driving measured toroidal asymmetries. In these experiments five toroidally distributed gas injectors in the private flux region were sequentially activated in separate discharges resulting in clear evidence of toroidal asymmetries in radiated power and nitrogen line emission as well as a ∼50% toroidal modulation in electron pressure at the divertor target. The pressure modulation is qualitatively reproduced by the modeling, with the simulation yielding a toroidal asymmetry in the heat flow to the outer strike point. Toroidal variation in impurity line emission is qualitatively matched in the scrape-off layer above the strike point, however kinetic corrections and cross-field drifts are likely required to quantitatively reproduce impurity behavior in the private flux region and electron temperatures and densities directly in front of the target., United States. Department of Energy (Contract DE-AC05-00OR22725), United States. Department of Energy (Contract DE-FC02-99ER54512)

Near-sonic plasma flows along magnetic field lines have been observed in the far scrape-off layer of some tokamaks. Such flows are simulated with multi-fluid edge transport code UEDGE and the results for Alcator C-Mod and NSTX tokamaks are presented. The important role of asymmetric intermittent cross-field transport and magnetic configuration in generating near-sonic flows is highlighted. The effect of large parallel flows on divertor detachment and main chamber recycling is discussed.

Alcator C-Mod uses high power radio frequency (rf) power for auxiliary plasma heating at the ion cyclotron frequency. The operation of the rf antennas is limited at neutral pressures which fall far below the minimum pressure for glow discharge as given by the rf Paschen curve. Recent experiments have found that this operational pressure limit is due to multipactor discharges in the vacuum transmission lines. At elevated neutral pressures, the multipactor discharge induces a glow discharge that causes a nearly complete reflection of rf power. On C-Mod, in presence of a magnetic field, this discharge forms at pressures of 0.5 and 1mTorr, which are the same pressures as the observed operational limit. Experimental results are supported by the Coaxial Multipactor Experiment.

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.

Scrape-off layer (SOL) data from DIII-D and C-Mod have been acquired and analysed for radial particle transport based on a particle balance model. This has allowed a detailed comparison for L-mode plasmas. The inferred radial particle flux, Γ⊥(r), is parameterized in terms of diffusive [Deff(r) ≡ Γ⊥(r)/∇n(r)] and convective particle transport [veff(r) ≡ Γ⊥(r)/n(r)]. The magnitude of the inferred Deff or veff increases across the SOL for both tokamaks. The inferred Deff or veff in the 'far' SOL (one density e-folding length from the separatrix and beyond) are essentially unchanged by changes in core density by factors of 2–3. This corresponds to changes in the far SOL density, collisionality (ν*) and radial fluxes of a factor of 10 or more. Thus ν* does not appear to be an important parameter in determining the radial particle transport in that region.The dimensionlessly-scaled SOL plasma profiles from the two tokamaks overlay for similar dimensionless plasma parameters. The SOL density profile near the separatrix is steeper than in the 'far' SOL. The scaled Deff and veff are slightly larger on DIII-D than C-Mod. This difference appears to be within experimental uncertainties.Neutral ionization in the SOL does not appear to affect radial transport but may be related to the observed flattening of the density profiles with increasing .

The observed toroidal rotation in Alcator C-Mod Ohmic L-mode plasmas has been found to depend strongly on the magnetic configuration. For standard discharges with a lower single null and the ion B ? ?B drift downward, and with BT = 5.4?T, IP = 0.8?MA and ne = 1.4 ? 1020?m?3, the core toroidal rotation is measured to be in the range of 10?20?km?s?1 (counter-current). Similar plasmas with upper single null (USN) have significantly stronger counter-current rotation, in the range 30?50?km?s?1. The rotation depends very sensitively on the distance between the primary and secondary separatrices in near double null plasmas, with changes of ~25?km?s?1 occurring over a variation of a few millimetres in this distance. Application of ICRF power has been found to increase the rotation in the co-current direction. The transition to H-mode is seen to occur in these standard plasmas when the core rotation reaches a characteristic value, near 0?km?s?1; hence higher input power is needed to induce the transition in the USN configuration.

We report on impurity transport modeling using the DIVIMP code which is able to qualitatively reproduce the poloidal variation of non-recycling impurity penetration factor (PF NR ) found in C-Mod experiments: a lower PF NR is computed at the inboard (3.6%) and divertor target locations (0.7%) than at the outboard (11%). By artificially increasing the modeled inner SOL plasma flow to correspond to measured values, a better quantitative agreement between modeled and measured PFs is achieved. We have also roughly reproduced the observed penetration factor for recycling impurities both in time dependence and magnitude. The model has shown that under attached conditions, the majority of recycling impurity ions flow into the confined plasma through the outboard side separatrix. For detached conditions the impurity influx across the separatrix is more concentrated near the divertor.

We evaluate the limits of the spectroscopic measurement of scrape-off layer flows utilizing HeII excitation emission (486.6 nm) in the inboard, high-field, scrape-off region. Velocities are inferred from the Doppler shift of the HeII emission line for lower and upper single-null discharges, exhibiting differences of up to 20 km/s between the two. Unfortunately at high temperatures (>30 eV), the short ionization time, compared to momentum transfer times, makes the interpretation difficult for those plasma regimes. The inferred variations of plasma flows near the inner wall correlates with Langmuir Mach probe measurements in its dependence on magnetic topology. However the flow magnitudes are different. The discrepancy between probe and spectroscopic measurements is likely due a combination of two effects: the magnitude of gas injection needed to achieve a reasonable signal is found to be too perturbative and loss of high-energy He ions from the viewing volume.

Three-dimensional Monte Carlo neutral transport simulations of gas flow through the Alcator C-Mod sub-divertor yield conductances comparable to those found in dedicated experiments. All are significantly smaller than the conductance found with the previously used axisymmetric geometry. A benchmarking exercise of the code against known conductance values for gas flow through a simple pipe provides a physical basis for interpreting the comparison of the three-dimensional and experimental C-Mod conductances.

A novel, magnetically-driven swing probe was recently installed near the midplane on the high-field side SOL in Alcator C-Mod. The probe collects plasma from co- and counter-current directions during its respective 0–90° and 90–180° of motion, thus providing profiles of density, electron temperature and plasma flow parallel to magnetic field lines (Mach number, M∥) up to the separatrix. Results are reported from discharges with different magnetic topologies: lower single-null, upper single-null, and double-null. In single-null, a strong parallel flow (∣M∥∣ ∼ 1) is detected, which is always directed from the low- to high-field SOL. In double-null discharges, e-folding lengths in the high-field SOL are a factor of ∼4 shorter than the low-field SOL. Thus, plasma appears to ‘fill-in’ the high-field SOL in single-null plasmas, not by cross-field transport but by parallel flow from the low-field SOL – a picture consistent with a very strong ballooning-like component to the cross-field transport.

Rapid changes in the loading resistance of fast wave antennas can limit high power operations of heating systems in the ion cyclotron range of frequencies (ICRF). Although novel matching techniques are being developed to reduce their effects, understanding the physics involved in these variations is of interest to guide and facilitate the design effort. We have studied the dependence of the loading resistance upon plasma parameters for the three ICRF antennas in the Alcator C-Mod tokamak. In contrast with similar studies in JET and Tore Supra, the evanescent decay term was not found to play an important role. The dominant variations could be related to changes in the shape of the electron density profiles in the propagating region. In H-mode, the loading resistance decreases as the density at the top of the pedestal is increased, and increases for higher scrape-off-layer (SOL) densities. This dependence on global plasma parameters is generally identical for the three antennas, up to a proportionality constant, while local changes in front of an individual antennas could explain the residual discrepancy. To link the observations with theory, the surface impedance at the Faraday shield was calculated by solving the wave equation in a slab geometry using experimental radial density profiles. This approach leads to a good agreement with measurements over a wide range of operating conditions in L-mode, ELM-free and EDA H-mode plasmas, and it can be interpreted qualitatively in terms of impedance transformation in the SOL region. Implications for more complex modelling approaches are also discussed.

The physics of impurity transport in response to a local gas injection in the scrape-off layer (SOL) of Alcator C-Mod is investigated. Carbon 'plumes' are formed at variable locations in the SOL—up to the separatrix—by puffing deuterated ethylene gas (C2D4) through the end of a reciprocating fast-scanning probe. CCD cameras are used to simultaneously record CII and CIII emission patterns from two near-perpendicular views. The plume dispersal patterns are found to yield direct qualitative information about plasma flow, including the direction of vE × B near the separatrix. Impurity transport and plasma–surface interaction physics implicit in the three-dimensional plume structure are explored in detail using a Monte Carlo impurity transport code, with the aim of extracting background plasma-flow quantities. A number of important local effects involving plasma–probe interactions have been identified: a vertical E × B drift near the probe surface, a parallel electric field above the probe tip arising from plasma recycling off the probe surface, and sputtering of a carbon layer that dynamically forms on the probe surface. The emission patterns are also found to yield important information on flows in the SOL: the radial electric field (Er) in the near SOL and volume-averaged values of the parallel Mach number (M∥) in the far SOL. The Er values obtained from plume data compare favourably with estimates of Er based on the poloidal propagation velocity of edge plasma fluctuations measured using the scanning probe. Emission patterns obtained in EDA H-mode suggest that the radial electric field changes sign at the location of the associated quasi-coherent mode layer. Comparisons between parallel Mach numbers obtained from the plume data and probe measurements indicate that the probe over-estimates the parallel flow towards the divertor in the far SOL. This result supports the picture of particle balance in the SOL of Alcator C-Mod being dominated by main-chamber recycling, with weak plasma flow into the divertor.

Measurements from an inner wall mounted scanning Langmuir probe and modelling of Dα emission show a sharp fall-off of plasma density in the inboard private flux region. Dα emission is modelled using a one-dimensional space, two-dimensional velocity kinetic neutral code in order to track the changes in the density profile while the magnetic topology is changed from single to double null. A sharp density decay is observed to track with the location of the secondary separatrix. The data presented are consistent with a lower level of cross-field transport on field lines in the private flux region restricted to the high field side when compared to cross-field transport on field lines in the low field side scrape-off layer.

Fast intermittent transport has been observed in the scrape-off layer (SOL) of major tokamaks including Alcator C-Mod, DIII-D, and NSTX. This kind of transport is not diffusive but rather convective. It strongly increases plasma flux to the chamber walls and enhances the recycling of neutral particles in the main chamber. We discuss the anomalous cross-field convection (ACFC) model for impurity and main plasma ions and its relation to intermittent transport events, i.e. plasma density blobs and holes in the SOL. Along with plasma diffusivity coefficients, our transport model introduces time-independent anomalous convective velocities across the confining magnetic field. In the discharge modeling, diffusivity coefficients and ACFC velocity profiles are adjusted to match a set of representative experimental data. We use this model in the edge plasma physics code UEDGE to simulate the multi-fluid two-dimensional transport for these three tokamaks. We present simulation results suggesting the dominance of anomalous convection in the far SOL transport. These results are consistent with the hypothesis that the chamber wall is an important source of impurities and that different impurity charge states have different magnitude and sign of their anomalous radial convective velocities.

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.

For steady state high-confinement-mode (H-mode) operation, a relaxation mechanism is required to limit build-up of the edge gradient and impurity content. Alcator C-Mod [Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] sees two such mechanisms—EDA (enhanced D-alpha H mode) and grassy ELMs (edge localized modes), but not large type I ELMs. In EDA the edge relaxation is provided by an edge localized quasicoherent (QC) electromagnetic mode that exists at moderate pedestal temperature T 3.5, and does not limit the buildup of the edge pressure gradient. The q boundary of the operational space of the mode depends on plasma shape, with the q95 limit moving down with increasing plasma triangularity. At high edge pressure gradients and temperatures the mode is replaced by broadband fluctuations ( f

The intermittent turbulent transport in the scrape-off-layer (SOL) of Alcator C-Mod [I.H. Hutchinson, R. Boivin, P.T. Bonoli et al., Nucl. Fusion 41, 1391 (2001)] is studied experimentally by imaging with a very high density of spatial measurements. The two-dimensional structure and dynamics of emission from a localized gas puff are observed, and intermittent features (also sometimes called “filaments” or “blobs”) are typically seen. The characteristics of the spatial structure of the turbulence and their relationship to the time-averaged SOL profiles are discussed and compared with those measured on the National Spherical Torus Experiment [M. Ono, S. M. Kaye, Y.-K. M. Pong et al., Nucl. Fusion 40, 557 (2000)]. The experimental observations are compared also with three-dimensional nonlinear numerical simulations of edge turbulence. Radial profiles of the poloidal wave number spectra and the poloidal scale length from the simulations are in reasonable agreement with those obtained from the experimental ima...

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.

Accurate measurements of plasma flows in the scrape-off layer (SOL) are a necessary requirement for understanding the physics of tokamak edge plasmas. A system is being developed on Alcator C-Mod for inferring flows parallel (v 1 ) and perpendicular (v E×B ) to local magnetic field lines from impurity emission patterns ('plumes') generated by local gas injection. Carbon plumes are generated at variable location in the SOL by puffing deuterated ethylene gas (C 2 D 4 ) through the end of a reciprocating fast-scanning probe. Two intensified CCD cameras are used to record C -1 and C +2 emission patterns simultaneously from near-perpendicular views. Plumes are modeled using a Monte Carlo impurity transport code, from which values for the background flows may be extracted. The sensitivity of the plume structure is investigated for a number of code inputs, including radial electric field (E r ) and the neutral launch dynamics. Initial modeling results indicate discrepancies between values of v || and E r extracted from the plumes and measurements obtained from probe data.

Parallel and E × B plasma flows near the separatrix on the outside midplane of Alcator C-Mod are investigated with a scanning probe for a range plasma densities, currents, and magnetic fields. Strong parallel flows (up to Mach 0.6) are found to peak ∼2 mm into the scrape-off layer (SOL), reverse nearly symmetrically with magnetic field reversal, and decrease in magnitude with increasing line-averaged density normalized to the Greenwald density. E × B flows in the SOL inferred from the poloidal propagation velocity of plasma fluctuations appear to compensate these parallel flows and scale similarly, i.e., the dominant flow pattern is a pure toroidal rotation. E × B flows inferred by probe-sheath potentials are generally smaller, exhibit more scatter, and do not scale the same, perhaps indicating a less reliable measurement. These measurements suggest a residual poloidal flow (along field lines) of ∼0.2 to ∼1 times the electron diamagnetic velocity, depending on plasma conditions.

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.

The nature of intermittency, long observed in magnetic fusion devices, was revisited lately [G. Antar et al., Phys. Rev. Lett. 87, 065001 (2001)]. It was shown that intermittency is caused by large-scale events with high radial velocity reaching about 1/10th of the sound speed. These type of structures were named “avaloids.” In the present article, the universality of convective turbulence in magnetically confined plasmas is investigated. Turbulence properties in the scrape-off layer of four different magnetic fusion devices are compared. Namely, the Tore Supra tokamak [Tore Supra Team, Nuclear Fusion, 40, 1047 (2000)] with circular cross-section limiter-bounded plasma, the Alcator C-Mod tokamak [B. LaBombard et al., Phys. Plasmas 8, 2107 (2001)] which is a divertor device, the Mega-Ampere Spherical Tokamak (MAST) [A. Sykes et al., Phys. Plasmas 8, 2101 (2001)] with vacuum chamber walls far from the plasma last closed flux surface and the PISCES linear plasma device [D. Geobel et al., Rev. Sci. Istrum. 56...