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Six nominally repeat neon shattered pellet injection (SPI) shutdowns of stable DIII-D Super H-modes are studied to understand the 3D properties of the radiation and impurity transport. The radiation efficiency and radiation peaking determine whether first wall melting is expected following disruption mitigation in ITER. Previous studies make use of axisymmetric approximations to infer radiation efficiencies, but validating the high efficiency required by ITER necessitates improved accuracy, and this work contributes by exploring the 3D radiation and density structures that will inform forward modeling. When the neon shatter plume produced by the SPI reaches the plasma edge, m/n = 3/1 and 2/1 island O-points are observed to align with the injection trajectory in five out of six cases, suggesting that the injected material seeds the island O-points. Field aligned neon structures emitting Ne-I line radiation drift at 1 km s−1 in the ion diamagnetic drift direction during the pre-thermal quench, tracking the motion of the m/n = 2/1 island O-point. Neon fragments penetrate to the q = 2 surface by the time of the thermal quench. Techniques to constrain the 3D emissivity are explored, and one method constrains a 3D flux tube that is consistent with the radiation data, and when mapped to the interferometers, intersects the lasers that measure the highest density. The resulting structure derived from the radiation measurements exists near the 2/1 island X-point. In five repeatable discharges, the peak of the radiation in the toroidal direction exists in a 120° toroidal sector where the injection occurs, in contrast with the outlier discharge where the toroidal peak exists in the complementary 240° toroidal sector far from the injector, and where a 50% lower density rise is observed. The n = 1 phase behavior is markedly different in the outlier discharge, suggesting a possible dependence of the radiation structure and the assimilation efficiency on MHD.

A real-time disruption predictor using random forest was developed for high-density disruptions and used in the plasma control system (PCS) of the EAST tokamak for the first time. The disruption predictor via random forest (DPRF) ran in piggyback mode and was actively exploited in dedicated experiments during the 2019–2020 experimental campaign to test its real-time predictive capabilities in oncoming high-density disruptions. During dedicated experiments, the mitigation system was triggered by a preset alarm provided by DPRF and neon gas was injected into the plasma to successfully mitigate disruption damage. DPRF’s average computing time of ∼250 μs is also an extremely relevant result, considering that the algorithm provides not only the probability of an impending disruption, i.e. the disruptivity, but also the so-called feature contributions, i.e. explainability estimates to interpret in real time the drivers of the disruptivity. DPRF was trained with a dataset of disruptions in which the electron density reached at least 80% of the Greenwald density limit, using the zero-dimensional signal routinely available to the EAST PCS. Through offline analysis, an optimal warning threshold on the DPRF disruptivity signal was found, which allows for a successful alarm rate of 92% and a false alarm rate of 9.9%. By analyzing the false alarm causes, we find that a fraction (∼15%) of the misclassifications are due to sudden transitions of plasma confinement from H- to L-mode, which often occur during high-density discharges in EAST. By analyzing DPRF feature contributions, it emerges that the loop voltage signal is that main cause of such false alarms: plasma signals more apt to characterize the confinement back-transition should be included to avoid false alarms.

Databases of physics events have been used in various fusion research applications, including the development of scaling laws and disruption avoidance algorithms, yet they can be time-consuming and tedious to construct. This paper presents a novel application of the label spreading semi-supervised learning algorithm to accelerate this process by detecting distinct events in a large dataset of discharges, given few manually labeled examples. A high detection accuracy (>85%) for H–L back transitions and initially rotating locked modes is demonstrated on a dataset of hundreds of discharges from DIII-D with manually identified events for which only three discharges are initially labeled by the user. Lower yet reasonable performance (∼75%) is also demonstrated for the core radiative collapse, an event with a much lower prevalence in the dataset. Additionally, analysis of the performance sensitivity indicates that the same set of algorithmic parameters is optimal for each event. This suggests that the method can be applied to detect a variety of other events not included in this paper, given that the event is well described by a set of 0D signals robustly available on many discharges. Procedures for analysis of new events are demonstrated, showing automatic event detection with increasing fidelity as the user strategically adds manually labeled examples. Detections on Alcator C-Mod and EAST are also shown, demonstrating the potential for this to be used on a multi-tokamak dataset.

In this study, a full convolutional neural network is trained on a large database of experimental EAST data to classify disruptive discharges and distinguish them from non-disruptive discharges. The database contains 14 diagnostic parameters from the ∼104 discharges (disruptive and non-disruptive). The test set contains 417 disruptive discharges and 999 non-disruptive discharges, which are used to evaluate the performance of the model. The results reveal that the true positive (TP) rate is ∼ 0.827, while the false positive (FP) rate is ∼0.067. This indicates that 72 disruptive discharges and 67 non-disruptive discharges are misclassified in the test set. The FPs are investigated in detail and are found to emerge due to some subtle disturbances in the signals, which lead to misjudgment of the model. Therefore, hundreds of non-disruptive discharges from training set, containing time slices of small disturbances, are artificially added into the training database for retraining the model. The same test set is used to assess the performance of the improved model. The TP rate of the improved model increases up to 0.875, while its FP rate decreases to 0.061. Overall, the proposed data-driven predicted model exhibits immense potential for application in long pulse fusion devices such as ITER.

Disruption database and disruption warning database of the EAST tokamak had been established by a disruption research group. The disruption database, based on Structured Query Language (SQL), comprises 41 disruption parameters, which include current quench characteristics, EFIT equilibrium characteristics, kinetic parameters, halo currents, and vertical motion. Presently most disruption databases are based on plasma experiments of non-superconducting tokamak devices. The purposes of the EAST database are to find disruption characteristics and disruption statistics to the fully superconducting tokamak EAST, to elucidate the physics underlying tokamak disruptions, to explore the influence of disruption on superconducting magnets and to extrapolate toward future burning plasma devices. In order to quantitatively assess the usefulness of various plasma parameters for predicting disruptions, a similar SQL database to Alcator C-Mod for EAST has been created by compiling values for a number of proposed disruption-relevant parameters sampled from all plasma discharges in the 2015 campaign. The detailed statistic results and analysis of two databases on the EAST tokamak are presented.

A disruption prediction algorithm, named DPRF (Disruption Prediction using Random Forests), has run in real-time in the DIII-D plasma control system (PCS) over more than 900 discharges. DPRF naturally provides a probability mapping associated to its predictions, i.e. the disruptivity signal, now incorporated in the DIII-D PCS. This paper discusses disruption prediction accomplishments in terms of shot-by-shot performances, by simulating alarms on each discharge as in the PCS framework. Depending on the optimised performance metric chosen to evaluate DPRF, we find that almost all disruptive discharges are detected in average with few hundred of milliseconds of warning time, but this comes at a high cost of false alarms produced. Performances are not satisfying ITER requirements, where the success rate has to be higher than 95%, but this is not completely unexpected. DPRF is trained on many years of major disruptions occurring during the flattop phase of the plasma current on DIII-D, but without any differentiation by cause. Furthermore, we find that DPRF is affected by a relatively high fraction of false alarms occurring during the first 500 milliseconds from the flattop onset. This subtle effect, more evident on the discharges where DPRF ran in real-time, can be marginalized by taking specific precautions on the validity range of the predictions, and performances do improve. Even if presently burdened by some limitations, DPRF provides an incredible and novel advantage. Thanks to the feature contribution analysis (e.g., the identification of which signals contributed to triggering an alarm), it is possible to interpret and explain DPRF predictions. It is the first time that such interpretability features is exploited by a disruption predictor: by uncovering the disruption events' causes, a better understanding of disruption dynamics is achieved, and a clear path toward the design of disruption avoidance strategies can be provided.

This paper reports on disruption prediction using a shallow machine learning method known as a random forest, trained on large databases containing only plasma parameters that are available in real-time on Alcator C-Mod, DIII-D, and EAST. The database for each tokamak contains parameters sampled ~106 times throughout ~104 discharges (disruptive and non-disruptive) over the last four years of operation. It is found that a number of parameters (e.g. , , , ) exhibit changes in aggregate as a disruption is approached on one or more of these tokamaks. However, for each machine, the most useful parameters, as well as the details of their precursor behaviors, are markedly different. When the prediction problem is framed using a binary classification scheme to discriminate between time slices 'close to disruption' and 'far from disruption', it is found that the prediction algorithms differ substantially in performance among the three machines on a time slice-by-time slice basis, but have similar disruption detection rates (~80%–90%) on a shot-by-shot basis after appropriate optimisation. This could have important implications for disruption prediction and avoidance on ITER, for which development of a training database of disruptions may be infeasible. The algorithm's output is interpretable using a method that identifies the most strongly contributing input signals, which may have implications for avoiding disruptive scenarios. To further support its real-time capability, successful applications in inter-shot and real-time environments on EAST and DIII-D are also discussed.

This work reports observations of a tokamak plasma that experienced a thermal quench due to a large, transient high-Z influx but avoided a current quench. This is argued to be caused by the presence of lower hybrid range of frequency (LHRF) waves that sustain a non-thermal, current-carrying electron population. In Alcator C-Mod L-mode plasmas at kA, m?3, nearly all of the current can be sustained non-inductively by injecting 700 kW of LHRF power at 4.6 GHz and . A sudden influx of a large amount of tungsten, , triggers a cooling wave that propagates at 2?3 m s?1 all the way into the core, dropping on-axis T e from 3 keV to temperatures less than measurement floor of 50?eV. An off-axis reheat begins after 100?ms, but T e profiles remain hollow for 300?350?ms after the injection. Throughout this temperature evolution, the plasma density, current and shape remain unchanged to within 10%. Following the expulsion of the tungsten, the plasma returns to its baseline conditions and the plasma ends as planned with a controlled current ramp-down. Energy balance analysis shows the LHRF power continues to be absorbed in the plasma after the thermal quench, as a significant fraction of it is needed to be consistent with radiated power measurements. Examination of current relaxation time, , and fast-electron slowing down time, , indicate the LHRF must contribute to driving current, despite the low temperatures, as the current remains nominally stationary despite ms and ms for relativistic electrons. These measurements represent an important existence proof of a possible technique for avoidance of disruptions caused by sudden, unplanned influx of impurities in the form of dust or flakes of high-Z wall material. Implications and suggestions for future experimental and modeling and simulation work are summarized.

In the Alcator C-Mod tokamak, relativistic runaway electron (RE) generation can occur during the flattop current phase of low density, diverted plasma discharges. Due to the high toroidal magnetic field (B-0 = 5.4 T), RE synchrotron radiation is measured by a wide-view camera in the visible wavelength range (lambda approximate to 400-900 nm). In this paper, a statistical analysis of over one thousand camera images is performed to investigate the plasma conditions under which synchrotron emission is observed in C-Mod. In addition, the spatiotemporal evolution of REs during one particular discharge is explored in detail via a thorough analysis of the distortion-corrected synchrotron images. To accurately predict RE energies, the kinetic solver CODE (Landreman et al 2014 Comput. Phys. Commun. 185 847-855) is used to evolve the electron momentum-space distribution at six locations throughout the plasma: the magnetic axis and flux surfaces q = 1, 4/3, 3/2, 2, and 3. These results, along with the experimentally-measured magnetic topology and camera geometry, are input into the synthetic diagnostic SOFT (Hoppe et al 2018 Nucl. Fusion 58 026032) to simulate synchrotron emission and detection. Interesting spatial structure near the surface q = 2 is found to coincide with the onset of a locked mode and increased MHD activity. Furthermore, the RE density profile evolution is fit by comparing experimental to synthetic images, providing important insight into RE spatiotemporal dynamics.

Using data-driven methodology, we exploit the time series of relevant plasma parameters for a large set of disrupted and non-disrupted discharges to develop a classification algorithm for detecting disruptive phases in shots that eventually disrupt. Comparing the same methodology on different devices is crucial in order to have information on the portability of the developed algorithm and the possible extrapolation to ITER. Therefore, we use data from two very different tokamaks, DIII-D and Alcator C-Mod. We focus on a subset of disruption predictors, most of which are dimensionless and/or machine-independent parameters, coming from both plasma diagnostics and equilibrium reconstructions, such as the normalized plasma internal inductance l i and the n = 1 mode amplitude normalized to the toroidal magnetic field. Using such dimensionless indicators facilitates a more direct comparison between DIII-D and C-Mod. We then choose a shallow Machine Learning technique, called Random Forests, to explore the databases available for the two devices. We show results from the classification task, where we introduce a time dependency through the definition of class labels on the basis of the elapsed time before the disruption (i.e. 'far from a disruption' and 'close to a disruption'). The performances of the different Random Forest classifiers are discussed in terms of several metrics, by showing the number of successfully detected samples, as well as the misclassifications. The overall model accuracies are above 97% when identifying a 'far from disruption' and a 'disruptive' phase for disrupted discharges. Nevertheless, the Forests are intrinsically different in their capability of predicting a disruptive behavior, with C-Mod predictions comparable to random guesses. Indeed, we show that C-Mod recall index, i.e. the sensitivity to a disruptive behavior, is as low as 0.47, while DIII-D recall is ~0.72. The portability of the developed algorithm is also tested across the two devices, by using DIII-D data for training the forests and C-Mod for testing and vice versa.

High pressure noble gas injection is a promising technique to mitigate the effect of disruptions in tokamaks. In this paper, results of mitigation experiments with low-Z massive gas injection (helium) on the EAST tokamak are reported. A fast valve has been developed and successfully implemented on EAST, with valve response time ≤150 μs, capable of injecting up to particles, corresponding to 300 times the plasma inventory. Different amounts of helium gas were injected into stable plasmas in the preliminary experiments. It is seen that a small amount of helium gas () can not terminate a discharge, but can trigger MHD activity. Injection of 40 times the plasma inventory impurity () can effectively radiate away part of the thermal energy and make the electron density increase rapidly. The mitigation result is that the current quench time and vertical displacement can both be reduced significantly, without resulting in significantly higher loop voltage. This also reduces the risk of runaway electron generation. As the amount of injected impurity gas increases, the gas penetration time decreases slowly and asymptotes to (~7 ms). In addition, the impurity gas jet has also been injected into VDEs, which are more challenging to mitigate that stable plasmas.

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.

Energetic minority protons with ~100 keV effective temperature are routinely created in Alcator C-Mod plasmas with the application of ion-cyclotron-range-of-frequency (ICRF) heating. A multichannel compact neutral particle analyzer (CNPA) is used to make measurements of these distributions in Alcator C-Mod's unique and reactor-relevant operating space via active and passive charge-exchange techniques. A radially injected 50 keV diagnostic hydrogen neutral beam is used for active analysis. Using a detailed model that accounts for beam, halo and impurity electron donors, core proton temperatures of ~30–120 keV are directly measured for the first time in lower density (ne0 ~ (0.8–1.5) × 1020 m−3) Alcator C-Mod plasmas with up to only ~0.5 MW of ICRF power. The model found that the minority proton temperatures are peaked spatially away from r/a = 0, even for an on-axis resonance. Additionally, noticeable phase-space anisotropy is seen as expected for ICRF heating. The measured effective temperatures also scale approximately with the Stix parameter. The CNPA temperature measurements are compared with several leading simulation packages such as the TORIC/FPPRF and AORSA/CQL3D full-wave/Fokker–Planck (FW/FP) solvers. Preliminary comparisons with the AORSA/CQL3D code which include results from a new synthetic diagnostic show good agreement and demonstrate that accurate tracking of the minority distribution during iterations of the FW and FP solvers is required to simulate Alcator C-Mod's energetic minority populations with accuracy. Physically, poor wave focusing and preferential heating of trapped energetic protons are found to move the fast proton temperature profiles off-axis. These FW/FP analyses represent the first comparison between predictions of these detailed codes and core minority tail experimental measurements on Alcator C-Mod.

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)

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.

Anomalous momentum transport has been observed in Alcator C-Mod tokamak plasmas. The time evolution of core impurity toroidal rotation velocity profiles has been measured with a tangentially viewing crystal x-ray spectrometer array. Following the L-mode to EDA (enhanced Dα) H-mode transition in both Ohmic and ion cyclotron range of frequencies heated discharges, the ensuing co-current toroidal rotation velocity, which is generated in the absence of any external momentum source, is observed to propagate in from the edge plasma to the core with a timescale of the order of the observed energy confinement time, but much less than the neo-classical momentum confinement time. The ensuing steady state toroidal rotation velocity profiles in EDA H-mode plasmas are relatively flat, with V ~ 50 km s−1, and the momentum transport can be simulated using a simple diffusion model. Assuming that the L–H transition produces an instantaneous edge source of toroidal torque (which disappears at the H- to L-mode transition), the momentum transport may be characterized by a diffusivity, with values of ~0.07 m2 s−1 during EDA H-mode and ~0.2 m2 s−1 in L-mode. These values are large compared to the calculated neo-classical momentum diffusivities, which are of the order of 0.003 m2 s−1. Velocity profiles of ELM-free H-mode plasmas are centrally peaked (with V(0) exceeding 100 km s−1 in some cases), which suggests the presence of an inward momentum pinch; the observed profiles are consistent with simulations including an edge inward convection velocity of ~10 m s−1. In EDA H-mode discharges which develop internal transport barriers, the velocity profiles become hollow in the centre, indicating the presence of a negative radial electric field well in the vicinity of the barrier foot. Upper single null diverted and inner wall limited L-mode discharges exhibit strong counter-current rotation (with V(0)~−60 km s−1 in some cases), which may be related to the observed higher H-mode power threshold in these configurations. For plasmas with locked modes, the toroidal rotation is observed to cease (V ≤ 5 km s−1).

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.

Internal transport barrier (ITB) plasmas with peak pressures of 0.25 MPa and pressure gradients as large as 2.5 MPa m−1 have been produced in Alcator C-Mod using off-axis ion cyclotron range of frequencies (ICRF) heating. The onset of the ITBs is apparent when the pressure gradient exceeds 1.0 MPa m−1, which is similar to the JET criterion of ρs/L≥0.014. Concommitant with the peaking of the core pressure as the ITB develops is a drop of the toroidal rotation velocity profile inside of the barrier foot; the maximum of the velocity gradient coincides with the peak in the pressure gradient. The quasi-coherent (QC) mode, associated with the enhanced Dα (EDA) H-mode plasmas which evolve the ITBs, breaks up and disappears as the barriers develop, even though the measured edge pedestal parameters remain fixed. The position of the ITB foot has been moved over a range of (1/3) of the plasma minor radius by varying the toroidal magnetic field. The peak in the calculated bootstrap current density profile has correspondingly been regulated over a similar range in plasma minor radius. The location of the density profile foot is found to expand as the toroidal magnetic field is reduced and the ICRF frequency is lowered. The density foot radius is relatively independent of q95, however, in a scan of the plasma current at fixed BT and wave frequency.

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

Since Alcator C-Mod normally operates at high toroidal field and high collisionality, β limiting instabilities are rarely observed. Under some conditions, however, when operating at low collisionality and high input power (PICRF≤5 MW), large amplitude (5×10-5 0.52 and increased in amplitude with increasing β until a rollover or collapse in β occurred. The largest amplitude modes with m = 2,n = 1 strongly degraded momentum and energy confinement when the modes coupled across the plasma core and locked to the wall, bringing the plasma ion toroidal rotation to zero, within experimental errors, about 50 ms after mode locking. MHD stability was calculated with the linear resistive toroidal MHD code MARS for a discharge with a large m = 2,n = 1 mode. Comparisons with neoclassical tearing mode (NTM) theory and with NTMs found on other tokamaks indicate that these modes may be driven by a combination of resistive, neoclassical, and error field effects.

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.

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

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

Measurements of profiles of soft X ray emissivity with 1.5 mm radial resolution are combined with high resolution electron density and temperature measurements in the edge region of the Alcator C-Mod tokamak to facilitate transport analysis of medium-Z impurities during H modes. Results from detailed modelling of the radiation and transport of fluorine are compared with experimental measurements, yielding information about the transport coefficients in the H mode transport barrier region. Evidence is found for a strong inward impurity pinch just inside the separatrix. The region of strong inward pinch agrees very well with the region of strong electron density gradient, suggesting that the inward pinch could be driven by the ion density gradient, as predicted by neoclassical theory. Simulations using the neoclassical impurity convection profile agree very well with experiments. Transport modelling shows that the X ray pedestal width is largely determined by the diffusion coefficient in the transport barrier. This allows diagnosis of changes in the edge diffusion coefficient on the basis of observations of X ray pedestal width changes. Significant differences in the edge diffusion coefficient are seen between different types of H mode. Several scalings for the edge diffusion coefficient in the enhanced Dα H mode are also identified. This may help elucidate the physical processes responsible for this attractive confinement mode.

The physics of sawtooth oscillations in plasmas produced by the Alcator C-Mod machine is discussed by identifying the regimes that characterize these oscillations, which occur in the central value of the electron temperature, as measured by X ray and/or ECE diagnostics. Sawteeth are generally present in ohmic as well as in RF heated plasmas. It is shown by numerical analysis that these plasmas are stable to ideal internal, pressure driven modes. Hence, the modes responsible for the crash phase of the sawteeth have to be resistive in nature. This is confirmed by numerical computations using a resistive code. The role of energetic ions, produced by ion cyclotron heating, is also discussed.

Just above the H-mode threshold power at low density, small amplitude high repetition rate (1-5 kHz) type III edge-localized modes (ELMs) are observed. At higher input power, the H-modes become ELM-free. At high density and high input power (, MW), the plasma often enters a new H-mode regime now called the enhanced H-mode, which is a steady-state H-mode often without distinct ELMs but with high levels of emission. Enhanced H-modes are accompanied by persistent high-frequency broadband (50-400 kHz) density and magnetic fluctuations. Above about , the enhanced H-modes may also exhibit bursting or continuous low m, n coherent magnetic fluctuations with frequencies from 60-90 kHz. In the H-mode, steep edge gradients are observed with a nine-channel electron cyclotron emission (ECE) polychromator and a 38-channel, high-resolution, edge soft-x-ray array. The soft-x-ray pedestal width remains around 3 to 4 mm in the ELM-free H-mode and between type III ELMs. The pedestal collapses within during an ELM. Calculations indicate that the plasma in the H-mode is often near ideal ballooning first stability limits, but no type I ELMs have yet been identified. Edge micro-tearing modes destabilized by a steep temperature gradient are also considered as a possible explanation of the observed fluctuations during the enhanced H-mode.

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

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

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

H-modes exhibiting improved confinement above the L-mode are achieved in Alcator C-Mod with ICRF and with ohmic heating alone without boronization. Both ELM-free and ELMy H-modes are obtained with total input power from 0.75 to 4.2 MW over a range of densities (0.8 to ) and toroidal fields (3 to 8 T). Type III ELMs are often observed to have coherent, high m and n precursor oscillations with frequencies of 100 - 160 kHz. The threshold power required to achieve the H-mode increases with density and toroidal field, in rough agreement with scalings derived from other tokamaks. The power densities and density times toroidal field products are an order of magnitude larger than in other tokamaks, in the range of values expected for ITER. The L - H and H - L transitions occur at approximately the same edge electron temperature. A low density limit to the H-mode is found at about . A high midplane neutral pressure limit of about 0.6 mTorr is also observed.

Disruptions in Alcator C-Mod can generate large eddy currents in the highly conducting vacuum vessel and internal structures, including a significant poloidal component due to halo currents. In order to understand better the stresses arising from the resulting J*B forces, Alcator C-Mod has been fitted with a comprehensive set of sensors to measure the spatial distribution and temporal behaviour of the halo currents. It is found that they are toroidally asymmetric, with a typical peaking factor of 2. The asymmetric pattern usually rotates toroidally at a few kilohertz, thus ruling out first wall non-uniformities as the cause of the asymmetry. Analysis of the information compiled in the C-Mod disruption database indicates that the maximum halo current during a disruption scales roughly as either Ip2/Bphi or Ip/q95, but that there is a large amount of variation that is not yet understood

This work discusses sources of uncertainty in the validation of lower hybrid wave current drive simulations against experiments, by evolving self-consistently the magnetic equilibrium and the heating and current drive profiles, calculated with a combined toroidal ray tracing code and 3D Fokker–Planck solver. The simulations indicate a complex interplay of elements, where uncertainties in the input plasma parameters, in the models and in the transport solver combine and—in some cases—compensate each other. It is concluded that ray-tracing calculations should include a realistic representation of the density and temperature in the region between the confined plasma and the wall, which is especially important in regimes where the LH waves are weakly damped and undergo multiple reflections from the plasma boundary. Uncertainties introduced in the processing of diagnostic data as well as uncertainties introduced by model approximations are assessed. It is shown that, by comparing the evolution of the plasma parameters in self-consistent simulations with available data, inconsistencies can be identified and limitations in the models or in the experimental data assessed.

Changes in the core toroidal rotation profiles following injection of lower hybrid (LH) waves have been documented in Alcator C-Mod plasmas. Shot by shot scans of LH input power have been performed at fixed magnetic field and electron density for several plasma currents. For sawtoothing target plasmas, if the input power is low enough that the central safety factor q 0 remains below 1, the change in the core rotation is in the counter-current direction, consistent in sign, magnitude and LH power scaling with direct momentum input from the LH waves. If the power level is high enough that there are significant changes to the q profile, including the termination of sawtooth oscillations, the change in the toroidal rotation is in the co-current direction, consistent with changes in the momentum flux through its dependence on the current density profile. The direction of the rotation changes depends on whether q 0 is below or above unity, and seemingly not on the magnetic shear, nor the Ohmic confinement regime of the target plasma.

Ion cyclotron resonance heating is the primary auxiliary heating on the Alcator C-Mod tokamak and is commonly used on other devices, and is planned for use on ITER. The RF-power density on C-Mod is above 5 MW m−3 providing for a unique opportunity to study wave–particle effects in the high RF power per particle regime. Minority heating produces a highly energetic tail in the minority distribution function which is measured using a compact neutral particle analyser. In this paper, we present the measurements of the fast-ion spectrum between 200 and 2 MeV, compiled over an entire experimental campaign. We also estimate the effective tail temperatures for the fast-ion distribution. We find that the fast-ion distribution is less energetic and less dense with increasing electron density; is more energetic with increasing plasma current; and is more dense but has no measurable change in energy with increasing RF power. Some possible explanations for these findings are discussed., United States. Dept. of Energy (Award DE-FC02-99ER54512)

Early results from the Alcator C-Mod tokamak indicate that ohmic energy confinement does not obey the Neo-Alcator scaling. Instead 'Ohmic L-mode' confinement, close to ITER89-P, is obtained, which substantially exceeds Neo-Alcator under some circumstances. ICRF-heated plasmas follow the same scaling. H-modes are observed both in ohmic and ICRF heated plasmas, with confinement enhancements of about 2 for ELM-free conditions. Divertor operation into the closed chamber shows different regimes, depending on the density. Of particular interest is the progressive detachment at moderate to high densities. Power to the divertor plates is generally low, reflecting the influence of strong radiation in the divertor. The scaling of parameters from Alcator's growing edge database is briefly summarized.

Ohmic H modes have been achieved in ALCATOR C-MOD with toroidal fields up to BT ≈ 5.25 T and line averaged densities up to ne ≈ 1.4 × 1020 m-3 with less than 1 MW of input power. In ELM-free cases at lower toroidal fields, the energy confinement increases by at least 60% and the particle confinement increases by more than a factor of two. The ELM-free H modes appear to fit the threshold power scaling observed on other tokamaks, P/S (MW/m2)=4.4 × 10-3 neBT (1019 m-3.T), up to power densities of about 0.15 MW/m2. At higher toroidal fields, ELMy H modes were observed after the injection of lithium pellets at about half of the scaled threshold power density. All of these results were obtained with molybdenum plasma facing components and the ELM-free H modes were obtained without boronization or other wall coatings

The collapse phase of sawtooth oscillations in tokamaks is believed to be due to the rapid growth of an m = 1 instability. We show here that the onset of the rapid growth is spontaneous and cannot be described by the increase of the linear growth rate arising from development of the plasma equilibrium. This result calls for a reappraisal of the sawtooth problem.

Simulations of massive gas injection for disruption mitigation in DIII-D are carried out to compare the toroidal peaking of radiated power for the cases of one and two gas jets. The radiation toroidal peaking factor (TPF) results from a combination of the distribution of impurities and the distribution of heat flux associated with the mode. When ignoring the effects of strong uni-directional neutral beam injection and rotation present in the experiment, the injected impurities are found to spread helically along field lines preferentially toward the high-field-side, which is explained in terms of a nozzle equation. Therefore when considering the plasma rest frame, reversing the current direction also reverses the toroidal direction of impurity spreading. During the pre-thermal quench phase of the disruption, the toroidal peaking of radiated power is reduced in a straightforward manner by increasing from one to two gas jets. However, during the thermal quench phase, reduction in the TPF is achieved only for a particular arrangement of the two gas valves with respect to the field line pitch. In particular, the relationship between the two valve locations and the 1/1 mode phase is critical, where gas valve spacing that is coherent with 1/1 symmetry effectively reduces TPF.

The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)-a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (>= 6.5 T, 1.5 MA), high power density facility (P/S similar to 1.5 MW m(-2)) will test innovative divertor ideas, including an 'X-point target divertor' concept, at the required performance parameters-reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region-while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magneticfield side-the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination-advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions-will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept (affordable, robust, compact) (Sorbom et al 2015 Fusion Eng. Des. submitted (arXiv: 1409.3540)) that makes use of high-temperature superconductor technology-a high-field (9.25 T) tokamak the size of the Joint European Torus that produces 270 MW of net electricity.

The beam-based plasma diagnostics on ITER require high accuracy and reliability which accordingly put challenging requirements on the collisional–radiative (CR) models used for description of the excited states of beam atoms. These states are known to play an essential role in implementation and interpretation of neutral beam-based diagnostics, such as the motional Stark effect (MSE), charge-exchange recombination spectroscopy (CXRS), beam-emission spectroscopy (BES) and in the dynamics of beam penetration. The latest analyses demonstrate that the widely used assumption of statistical populations among the beam excited states is questionable for high beam energies and high magnetic fields. Here we report an empirical verification of the recently developed non-statistical nkm-resolved CR model (Marchuk et al 2010 J. Phys. B: At. Mol. Opt. Phys. 43 011002) and the short extrapolation to the relevant parameter range for ITER. The experiment was performed on the Alcator C-Mod tokamak, which operates in a unique range of parameters well suited for testing CR models for ITER beams. Beam emission spectra are collected for a selected range of plasma parameters. The line ratios σ1/σ0, π4/π3 and Σσ/Σπ are measured and compared to the n-resolved statistical population model and the new non-statistical nkm-resolved CR model. The measured ratios show clear deviations from the statistical model and a good agreement with the non-statistical results. The difference between the experimental values and nkm-resolved simulations for the most part is within 10% for all three line ratios. The largest deviation here is for the σ1/σ0 ratio that reaches up to 14% for the lowest electron density of 0.6 × 1020 m−3. In contrast, the difference between experiment and the n-resolved (statistical) model is within 13–27% for the 50 keV/u Alcator C-Mod beam and expected to be up to 32% for 100 keV/u and up to 43% for 500 keV/u ITER beams. The discussed effect must be taken into account for the proposed CXRS/BES and MSE diagnostics for ITER.

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

Massive gas injection rapid shutdown experiments have been conducted on the Alcator C-Mod tokamak using two toroidally separated gas injectors, in order to investigate the effect of multiple gas injection locations on the toroidal asymmetry in the radiated power. Toroidal radiation asymmetry is diagnosed by an array of six single-channel photodiodes mounted on the vessel wall. The presence of magnetohydrodynamic (MHD) activity is diagnosed using an array of magnetic pickup (Mirnov) coils, mounted on stalks on the vessel wall. Scans were conducted of the relative timing between the two jets, of the 95th percentile safety factor, and of the plasma elongation. It is observed that firing the two gas jets so that the injected impurities arrive at the plasma at nearly the same time produced an increase in the toroidal radiation asymmetry. In addition, the radiation asymmetry in the thermal quench phase correlates with the growth rate of low toroidal mode number MHD modes, indicating that these mode(s) are playing a role in setting the radiation asymmetry.

Long-lived (1, 1) ?snake? modes were discovered nearly three decades ago, but basic questions regarding their formation, stability, and superb particle confinement?shown by surviving tens to hundreds of sawtooth cycles?have remained unanswered. High-resolution spectroscopic imaging diagnostics permit studies of heavy-impurity-ion snakes with unprecedented temporal and spatial resolution, making it possible to positively identify the SXR signals with specific ion charge states and to infer, for the first time, the perturbed impurity density, Zeff, and resistivity at the centre of these long-lived helical modes. The results show a new scenario for the formation of heavy-impurity-ion snakes, which can begin as a broad 1/1 kink asymmetry of the central impurity-ion density, that grows and undergoes a seamless transition to a large crescent-shaped helical island-like structure inside q?1, with a regularly sawtoothing core. This type of formation departs strongly from the nonlinear island model based on a modified Rutherford equation proposed originally to describe the pellet-induced snakes and expanded further to account for the impurity effects (e.g. and ). These new high-resolution observations show details of their evolution and the accompanying sawtooth oscillations that suggest important differences between the density and temperature dynamics, ruling out a purely pressure-driven process. Instead, many features arise naturally from nonlinear interactions in a 3D MHD model that separately evolves the plasma density and temperature.

Direction reversals of intrinsic toroidal rotation have been observed in Alcator C-Mod ohmic L-mode plasmas following modest electron density or toroidal magnetic field ramps. The reversal process occurs in the plasma interior, inside of the q = 3/2 surface. For low density plasmas, the rotation is in the co-current direction, and can reverse to the counter-current direction following an increase in the electron density above a certain threshold. Reversals from the co- to counter-current direction are correlated with a sharp decrease in density fluctuations with k R ⩾ 2 cm−1 and with frequencies above 70 kHz. The density at which the rotation reverses increases linearly with plasma current, and decreases with increasing magnetic field. There is a strong correlation between the reversal density and the density at which the global ohmic L-mode energy confinement changes from the linear to the saturated regime.