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The tokamak à configuration variable (TCV) continues to leverage its unique shaping capabilities, flexible heating systems and modern control system to address critical issues in preparation for ITER and a fusion power plant. For the 2019–20 campaign its configurational flexibility has been enhanced with the installation of removable divertor gas baffles, its diagnostic capabilities with an extensive set of upgrades and its heating systems with new dual frequency gyrotrons. The gas baffles reduce coupling between the divertor and the main chamber and allow for detailed investigations on the role of fuelling in general and, together with upgraded boundary diagnostics, test divertor and edge models in particular. The increased heating capabilities broaden the operational regime to include T e/T i ∼ 1 and have stimulated refocussing studies from L-mode to H-mode across a range of research topics. ITER baseline parameters were reached in type-I ELMy H-modes and alternative regimes with ‘small’ (or no) ELMs explored. Most prominently, negative triangularity was investigated in detail and confirmed as an attractive scenario with H-mode level core confinement but an L-mode edge. Emphasis was also placed on control, where an increased number of observers, actuators and control solutions became available and are now integrated into a generic control framework as will be needed in future devices. The quantity and quality of results of the 2019–20 TCV campaign are a testament to its successful integration within the European research effort alongside a vibrant domestic programme and international collaborations.

In some conditions, I-mode plasmas can feature pedestal relaxation events (PREs) that transiently enhance the energy reaching the divertor target plates. To shed light on their appearance, characteristics and energy reaching the divertor targets, a comparative study between two tokamaks — Alcator C-Mod and ASDEX Upgrade — is carried out. It is found that PREs appear only in a subset of I-mode discharges, mainly when the plasma is close to the H-mode transition. Also, a growing oscillating precursor before the PRE onset is observed in the region close to the separatrix in both devices, and a discussion on a possible triggering mechanism is outlined. The PRE relative energy loss from the confined region is found to increase with decreasing pedestal top collisionality ν ped * . Similarly, also the relative electron temperature drop at the pedestal top, which is related to the conductive energy loss, rises with decreasing ν ped * . Based on these relations, the PRE relative energy loss in future devices such as DEMO and ARC is estimated. Finally, the divertor peak energy fluence due to the PRE is measured on each device. Those values are then compared to the model introduced in Eich et al (2017 Nucl. Mater. Energy 12 84–90) for type-I edge localized modes. The model is shown to provide an upper boundary for PRE energy fluence data, while a lower boundary is found by dividing the model by three. These two boundaries are used to make projections of the PRE divertor energy fluence to DEMO and ARC.

A new Gas Puff Imaging (GPI) diagnostic has been installed on the TCV tokamak, providing two-dimensional insights into Scrape-Off-Layer (SOL) turbulence dynamics above, at and below the magnetic X-point. A detailed study in L-mode, attached, lower single-null discharges shows that statistical properties have little poloidal variations, while vast differences are present in the 2D behaviour of intermittent filaments. Strongly elongated filaments, just above the X-point and in the divertor far-SOL, show a good consistency in shape and dynamics with field-line tracing from filaments at the outboard midplane, highlighting their connection. In the near-SOL of the outer divertor leg, shortlived, high frequency and more circular (diameter $\sim$15 sound Larmour radii) filaments are observed. These divertor-localised filaments appear born radially at the position of maximum density and display a radially outward motion with velocity $\approx$400\,m/s that is comparable to radial velocities of upstream-connected filaments. Conversely, in these discharges ($B\times\nabla B$ pointing away from the divertor), these divertor filaments' poloidal velocities differ strongly from those of upstream-connected filaments. The importance of divertor-localised filaments upon radial transport and profile broadening is explored using filament statistics and in-situ kinetic profile measurements along the divertor leg. This provides evidence that these filaments contribute significantly to electron density profile broadening in the divertor., This is the version of the article before peer review or editing, as submitted by an author to IOPScience Nuclear Fusion. IOP Publishing Ltd is not responsible for any errors or omissions in this version of the manuscript or any version derived from it. 15 pages, 15 figures

Using recently installed scrape-off layer diagnostics on the tokamak à configuration variable, we characterise the poloidal and parallel properties of turbulent filaments. We access both attached and detached divertor conditions across a wide range of core densities (f G ∈ [0.09, 0.66]) in diverted L-mode plasma configurations. With a gas puff imaging (GPI) diagnostic at the outer midplane we observed filaments with a monotonic increase in radial velocity (from 390 m s−1 to 800 m s−1) and cross-field radii (from 8.5 mm to 13.4 mm) with increasing core density. Interpreting the filament behaviour in the context of the two-region model by Myra et al (2006 Phys. Plasmas 13 112502), we find that they populate the ideal-interchange regime (C i) in discharges at very low densities, and the resistive X (RX)-point regime for all other discharges. The scaling of filament velocity versus size shows good agreement with this interpretation. These results are discussed and compared with previous probe-based measurements for similar conditions, which mostly placed filaments in TCV in the resistive ballooning (RB) regime (Tsui et al 2018 Phys. Plasmas 25 072506). In addition, for the first time in TCV, the parallel filament extension is studied by magnetically aligning the GPI measurements at the outboard midplane with a reciprocating probe in the divertor. In agreement with the filaments being in the ideal-interchange and the RX-point regimes, they are found to extend beyond the X-point into the outer divertor leg.

A synthetic gas puff imaging (GPI) diagnostic is applied to a gyrokinetic turbulence simulation of Alcator C-Mod with the objective of assessing the impact of turbulent plasma fluctuations on the GPI neutral density field. We find that these effects are significant for much of the run, resulting in qualitatively different light emission patterns relative to the idealized limit of a fixed neutral profile. Spatial and temporal variations in the local neutral density within the GPI gas cloud are shown to be comparable in magnitude to those in the plasma, which would affect the interpretation of plasma fluctuations made using GPI. Assessing the experimental implications of these results will require additional study using validated turbulence simulations. Our findings are compatible with previous theoretical investigations concluding that GPI may unreliably diagnose plasma turbulence more than one neutral mean free path away from the gas source. Although we used a comprehensive, time dependent simulation to arrive at these results, our analysis shows that a frozen-plasma treatment of the neutrals is adequate in this case. That is, we can assume that the neutral transport is too fast to be affected by temporal fluctuations in the turbulent plasma structures. Keywords: Synthetic diagnostic, Gas puff imaging, Neutral transport simulation, Plasma turbulence simulation, Alcator C-Mod

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.

This work presents new evidence that the heat flux width, λ q , in the Alcator C-Mod tokamak scales with the edge electron pressure, as observed in the ASDEX Upgrade (AUG) tokamak (Silvagni et al 2020 Plasma Phys. Control. Fusion 62 045015), but the scaling with volume-averaged pressure, p ¯ , from the plasma stored energy, found by Brunner et al (2018 Nucl. Fusion 58 094002), is a better predictor of λ q in Alcator C-Mod than the edge electron pressure. These previous studies, which find that λ q decreases with increasing plasma pressure, imply that a high performance core at high pressure will lead to challenging heat and particle exhaust due to very small λ q . This concern has led to our significant enlargement of the C-Mod database with the electron density, temperature, and pressure profile data from the Thomson scattering and electron cyclotron emission diagnostics. Using the C-Mod database augmented with new profile data, we find that λ q decreases with increasing edge electron pressure as λ q ∝ p e , 95 − 0.26 , similar to results from AUG, and showing the strength of cross-machine comparisons. We also find that λ q ∝ p e , c o r e − 0.56 , consistent with the original finding from C-Mod that the heat flux width scales as p ¯ − 0.48 (Brunner et al 2018 Nucl. Fusion 58 094002). The scalings of λ q with separatrix pressure and gradient scale length are found to match the AUG results qualitatively. The C-Mod scalings with edge plasma quantities have more scatter than the p ¯ scaling, and, importantly, show different trends for H-modes relative to L- and I-mode. Investigating the source of this discrepancy presents an opportunity for further study that may improve our ability to predict the heat flux width in different confinement scenarios in the pursuit of optimizing core-edge performance in future reactors.

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

As part of an effort to validate the edge plasma model in the SOLPS-ITER code suite under ITER-relevant divertor plasma and neutral conditions, we report on progress in the modeling of the Alcator C-Mod divertor plasma with the new code. We perform simulations with a complete drifts model and kinetic neutrals, including effects of neutral viscosity, ion-molecule collisions and Lyα-opaque conditions, but assuming a pure deuterium plasma. Through a series of simulations with varying divertor geometries, we show the importance of including neutal leakage paths through the divertor substructure on the divertor plasma solution. Moreover, the impact of drifts on inner-outer target asymmetries is assessed. Including both effects, we achieve excellent agreement between simulations and upstream and outer target Langmuir Probe data. In absence of strong volumetric losses due to e.g. impurity radiation in our simulations, the strong inner target detachment observed experimentally remains elusive in our modeling at present. Keywords: SOLPS-ITER, Alcator C-Mod, Divertor modeling

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)

A rich variety of field-aligned fluctuations has been revealed using fast imaging of Dα emission from Alcator C-Mod's lower X-point region. Field-aligned filamentary fluctuations are observed along the inner divertor leg, within the Private-Flux-Zone (PFZ), in the Scrape-Off Layer (SOL) outside the outer divertor leg, and, under some conditions, at or above the X-point. The locations and dynamics of the filaments in these regions are strikingly complex in C-Mod. Changes in the filaments’ generation appear to be ordered by plasma density and magnetic configuration. Filaments are not observed for plasmas with n/nGreenwald ≲ 0.12 nor are they observed in Upper Single Null configurations. In a Lower Single Null with 0.12 ≲ n/nGreenwald ≲ 0.45 and Bx∇B directed down, filaments typically move up the inner divertor leg toward the X-point. Reversing the field direction results in the appearance of filaments outside of the outer divertor leg. With the divertor targets “detached”, filaments inside the LCFS are seen. These studies were motivated by observations of filaments in the X-point and PFZ regions in MAST, and comparisons with those observations are made. Keywords: Alcator C-Mod, Turbulence, Divertor, X-point, Filaments

Extremely intense power exhaust channels are projected for tokamak-based fusion power reactors; a means to handle them remains to be demonstrated. Advanced divertor configurations have been proposed as potential solutions. Recent modelling of tightly baffled, long-legged divertor geometries for the divertor test tokamak concept, ADX, has shown that these concepts may access passively stable, fully detached regimes over a broad range of parameters. The question remains as to how such divertors may perform in a reactor setting. To explore this, numerical simulations are performed with UEDGE for the long-legged divertor geometry proposed for the ARC pilot plant conceptual design—a device with projected heat flux power width ( ) of 0.4 mm and power exhaust of 93 MW—first for a simplified Super-X divertor configuration (SXD) and then for the actual X-point target divertor (XPTD) being proposed. It is found that the SXD, combined with 0.5% fixed-fraction neon impurity concentration, can produce passively stable, detached divertor regimes for power exhausts in the range of 80–108 MW—fully accommodating ARC’s power exhaust. The XPTD configuration is found to reduce the strike-point temperature by a factor of ∼10 compared to the SXD for small separations (∼1.4 ) between main and divertor X-point magnetic flux surfaces. Even greater potential reductions are identified for reducing separations to ∼1 or less. The power handling response is found to be insensitive to the level of cross-field convective or diffusive transport assumed in the divertor leg. By raising the separatrix density by a factor of 1.5, stable fully detached divertor solutions are obtained that fully accommodate the ARC exhaust power without impurity seeding. To our knowledge, this is the first time an impurity-free divertor power handling scenario has been obtained in edge modelling for a tokamak fusion power reactor with of 0.4 mm.

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.

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.

This article is part of Ralph Kube's doctoral thesis, which is available in Munin at http://hdl.handle.net/10037/7088 Gas-puff imaging data is used to investigate the dynamics of blob-like structures in the scrape-off layer of Alcator C-Mod. A blob-tracking algorithm is developed and used to compute blob propagation velocities and cross-field sizes for a series of discharges where the line-averaged particle density View the MathML source is varied. We find that the mean radial blob velocity increases from View the MathML source for View the MathML source to View the MathML source for View the MathML source, where is the nG the Greenwald density. The cross-field size of the blobs is independent of the line-averaged particle densities. With increasing View the MathML source the observed blob velocities become larger than predicted by simple blob models for sheath-connected plasma filaments.

Radio frequency (RF) sheath rectification is a leading mechanism suspected of causing anomalously high erosion of plasma facing materials in RF-heated plasmas on Alcator C-Mod. An extensive experimental survey of the plasma potential (Φ[subscript P]) in RF-heated discharges on C-Mod reveals that significant Φ[subscript P] enhancement (>100 V) is found on outboard limiter surfaces, both mapped and not mapped to active RF antennas. Surfaces that magnetically map to active RF antennas show Φ[subscript P] enhancement that is, in part, consistent with the recently proposed slow wave rectification mechanism. Surfaces that do not map to active RF antennas also experience significant Φ[subscript P] enhancement, which strongly correlates with the local fast wave intensity. In this case, fast wave rectification is a leading candidate mechanism responsible for the observed enhancement., United States. Dept. of Energy (DE-FC02-99ER54512)

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.

Alcator C-Mod was recently upgraded with multiple diagnostics to measure the heat flux footprint on the outer divertor: IR thermography, plasma conditions from Langmuir probes (LPs), and embedded calorimeters to measure local energy deposition. In addition to providing consistency checks of the inferred heat flux profiles, these measurements allow an assessment of the local sheath heat flux transmission coefficient (). We report measurements of the sheath heat-flux transmission coefficient at multiple locations in the scrape-off-layer (SOL) during an EDA H-mode discharge. The LP data are consistent with the presence of a heat flux “tail” measured by the IR system, extending into the far SOL. Measured are found to deviate from theoretical values by up to a factor of 3, depending on position in the SOL, pointing to some as yet unresolved physics. Calorimeter-inferred energy deposition profiles are found to be within 20% of the IR-inferred values across the footprint.

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.

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.

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.

High-pressure gas-jet injection disruption mitigation is investigated on Alcator C-Mod. The experimental results are encouraging for ITER. Gas-jets delivers reliable and substantial disruption mitigation of thermal loads and halo currents at the ITER field and pressure. The gas injection is compatible with a metallic wall. The gas-jets provide sufficient radiative energy dissipation on a timescale even faster than required for ITER. Runaway electrons are suppressed, however a better understanding of this suppression is still required for ITER. Present 0-D calculations of the gas delivery and global plasma response match important experimental data such as overall quench time and the current quench L/R time. Frictional dissipation may be important for gas delivery down the long tubes that will probably be required for ITER. However deep gas and/or impurity penetration is not required for mitigation, relieving technical requirements for the gas pressure in ITER. Initial studies with the non-ideal MHD simulation NIMROD show that radiation-induced MHD plays a critical role in thermal quench particle and energy dynamics. The application of a newly available version of NIMROD, with more complete atomic physics, shows promise in matching the experimental trends and could be an invaluable tool for increasing our confidence in designing efficient disruption mitigation for ITER.

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.

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.

The Rydberg series (1s2 to 1snp) up to n = 14 of helium-like argon (Z = 18) has been observed from Alcator C-Mod plasmas using a high-resolution X-ray spectrometer array. High-n satellites to these...

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 ...

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...

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.

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.

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.

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 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.

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.

The BOUT++ code has been exploited in order to improve the understanding of the role of turbulent modes in controlling edge transport and resulting scaling of the scrape-off layer (SOL) heat flux width. For the C-Mod enhanced (EDA) H-mode discharges, BOUT++ six-field two-fluid nonlinear simulations show a reasonable agreement of upstream turbulence and divertor target heat flux behavior: (a) the simulated quasi-coherent modes show consistent characteristics of the frequency versus poloidal wave number spectra of the electromagnetic fluctuations when compared with experimental measurements: frequencies are around 60–120 kHz (experiment: about 70–110 kHz), are around 2.0 which is similar to the phase contrast imaging data; (b) linear spectrum analysis is consistent with the nonlinear phase relationship calculation which indicates the dominance of resistive-ballooning modes and drift-Alfven wave instabilities; (c) the SOL heat flux width versus current I p scaling is reproduced by turbulent transport: the simulations yield similar to experimental measurements within a factor of 2. However the magnitudes of divertor heat fluxes can be varied, depending on the physics models, sources and sinks, sheath boundary conditions, or flux limiting coefficient; (d) Simple estimate by the '2-point model' for is consistent with simulation. Moreover, blobby turbulent spreading is confirmed for these relatively high B p shots.

The Alcator C-Mod team has recently developed a feedback system to measure and control surface heat flux in real-time. The system uses real-time measurements of surface heat flux from surface thermocouples and a pulse-width modulated piezo valve to inject low-Z impurities (typically N2) into the private flux region. It has been used in C-Mod to mitigate peak surface heat fluxes >40 MW m−2 down to 1. While the system works quite well under relatively steady conditions, use of it during transients has revealed important limitations on feedback control of impurity seeding in conventional vertical target plate divertors. In some cases, the system is unable to avoid plasma reattachment to the divertor plate or the formation of a confinement damaging x-point MARFE. This is due to the small operational window for mitigated heat flux in the parameters of incident plasma heat flux, plasma density, and impurity density as well as the relatively slow response of the impurity gas injection system compared to plasma transients. Given the severe consequences for failure of such a system to operate reliably in a reactor, there is substantial risk that the conventional vertical target plate divertor will not provide an adequately controllable system in reactor-class devices. These considerations motivate the need to develop passively stable, highly compliant divertor configurations and experimental facilities that can test such possible solutions.