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This paper reports the latest lower hybrid current drive (LHCD) experimental and modeling investigations conducted on Alcator C-Mod, which further elucidates the role of the scrape-off layer (SOL) in determining the magnitude and profile of wave power deposition. In the latest C-Mod LHCD experiment, with a proper management of the SOL plasma, a recovery of efficient current drive is observed in a diverted configuration at densities exceeding the previously identified LH density limit of n¯e ≈ 1.0x1020 m−3. By operating at higher current, i.e., at lower Greenwald fraction, the SOL width is reduced and the turbulence level is minimized. Under this condition, parasitic wave power losses occurring in the edge/SOL region are found to be minimized. In our modeling study, a shift or spread in the poloidal mode number by the SOL turbulence is investigated in order to better match the experimental power deposition profile with an available ray-tracing/Fokker-Planck model. The approach is to introduce an additional spectral broadening mechanism by launching the rays with a non-negligible poloidal wavenumber component while preserving the launched n// power spectrum. This approach contrasts to spectral broadening achieved in the n// space. An initial modeling investigation on the C-Mod plasma reproduces an experimental on-axis power deposition profile.

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)

The research program of the TCV tokamak ranges from conventional to advanced-tokamak scenarios and alternative divertor configurations, to exploratory plasmas driven by theoretical insight, exploiting the device's unique shaping capabilities. Disruption avoidance by real-time locked mode prevention or unlocking with electron-cyclotron resonance heating (ECRH) was thoroughly documented, using magnetic and radiation triggers. Runaway generation with high-Z noble-gas injection and runaway dissipation by subsequent Ne or Ar injection were studied for model validation. The new 1 MW neutral beam injector has expanded the parameter range, now encompassing ELMy H-modes in an ITER-like shape and nearly non-inductive H-mode discharges sustained by electron cyclotron and neutral beam current drive. In the H-mode, the pedestal pressure increases modestly with nitrogen seeding while fueling moves the density pedestal outwards, but the plasma stored energy is largely uncorrelated to either seeding or fueling. High fueling at high triangularity is key to accessing the attractive small edge-localized mode (type-II) regime. Turbulence is reduced in the core at negative triangularity, consistent with increased confinement and in accord with global gyrokinetic simulations. The geodesic acoustic mode, possibly coupled with avalanche events, has been linked with particle flow to the wall in diverted plasmas. Detachment, scrape-off layer transport, and turbulence were studied in L- and H-modes in both standard and alternative configurations (snowflake, super-X, and beyond). The detachment process is caused by power 'starvation' reducing the ionization source, with volume recombination playing only a minor role. Partial detachment in the H-mode is obtained with impurity seeding and has shown little dependence on flux expansion in standard single-null geometry. In the attached L-mode phase, increasing the outer connection length reduces the in-out heat-flow asymmetry. A doublet plasma, featuring an internal X-point, was achieved successfully, and a transport barrier was observed in the mantle just outside the internal separatrix. In the near future variable-configuration baffles and possibly divertor pumping will be introduced to investigate the effect of divertor closure on exhaust and performance, and 3.5 MW ECRH and 1 MW neutral beam injection heating will be added., This work was supported in part by the Swiss National Science Foundation.

This paper presents the first experimental analysis of polarized synchrotron emission from relativistic runaway electrons (REs) in a tokamak plasma. Importantly, we show that the polarization information of synchrotron radiation can be used to diagnose spatially-localized RE pitch angle distributions. Synchrotron-producing REs were generated during low density, Ohmic, diverted plasma discharges in the Alcator C-Mod tokamak. The ten-channel Motional Stark Effect diagnostic was used to measure spatial profiles of the polarization angle $\theta_{\mathrm{pol}}$ and the fraction $\mathrm{f}_{\mathrm{pol}}$ of detected light that was linearly-polarized. Spatial transitions in $\theta_{\mathrm{pol}}$ of 90$\deg$---from horizontal to vertical polarization and vice versa---are observed in experimental data and are well-explained by the gyro-motion of REs and high directionality of synchrotron radiation. Polarized synchrotron emission is modeled with the synthetic diagnostic SOFT; its output Green's (or detector response) functions reveal a critical RE pitch angle at which $\theta_{\mathrm{pol}}$ flips by 90$\deg$ and $\mathrm{f}_{\mathrm{pol}}$ is minimal. Using SOFT, we determine the dominant RE pitch angle which reproduces measured $\theta_{\mathrm{pol}}$ and $\mathrm{f}_{\mathrm{pol}}$ values. The spatiotemporal evolutions of $\theta_{\mathrm{pol}}$ and $\mathrm{f}_{\mathrm{pol}}$ are explored in detail for one C-Mod discharge. For channels viewing REs near the magnetic axis and flux surfaces $q$ = 1 and 4/3, disagreements between synthetic and experimental signals suggest that the sawtooth instability may be influencing RE dynamics. Furthermore, other sources of pitch angle scattering, not considered in this analysis, could help explain discrepancies between simulation and experiment.

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.

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

Recent progress in understanding and mitigating parasitic wave absorption in edge plasmas is presented. Experimental observations collected on Alcator C-Mod suggest multiple physics mechanisms are involved in such losses. Localized measurement of parametric decay instabilities (PDIs) has been performed using RF Langmuir probes. The divertor heat flux due to LH and ionization power loss have been evaluated quantitatively. We observe that the LHCD efficiency can be recovered when the SOL density profile is controlled by operating the tokamak at high current. The experimental progresses motivated a re-examination of the LHCD simulation model based on the ray-tracing/Fokker-Planck code (GENRAY/CQL3D). The effect of introducing a relatively small wave number broadening in the launched power spectrum and using 2D SOL density and temperature profiles was investigated. Comparison with C-Mod experiment indicates that the new model can explain the experimental trend over a wider density range including the density regime where disagreement was seen previously, suggesting that including realistic SOL geometry is a key to improve the simulation accuracy.

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)

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.

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

Upgrades to the motional Stark effect (MSE) diagnostic on Alcator C-mod have enabled accurate measurement of the current profile using pitch angle constrained magnetic equilibrium reconstructions. The MSE diagnostic utilizes an intrashot calibration technique along with kinetic profiles to constrain the reconstruction throughout the discharge. The system was used to study the current profile in plasmas with substantial Lower Hybrid Current Drive (LHCD), including fully non-inductive discharges sustained for several current relaxation times. The current profile evolution is reconstructed using the MSE data with 100ms time resolution as LHCD is applied to the plasma and as the plasma current profile relaxes back to inductively driven after LHCD ceases. The LH driven current is observed to be off-axis and significantly broadens the plasma current profile, resulting in qo above 1 and sawtooth suppression. In fully non-inductive discharges the reconstructed q profile is flat or non-monotonic. In plasmas with s...

Application of lower hybrid range of frequencies (LHRF) waves can induce both co- and counter-current directed changes in toroidal rotation in Alcator C-Mod plasmas, depending on the target plasma current, electron density, confinement regime and magnetic shear. For ohmic L-mode discharges with good core LH wave absorption, and significant current drive at a fixed LH power near 0.8 MW, the interior (r/a q95/11.5, and in the co-current direction if ne(1020 m−3) 1, indicating a good correlation with driven current fraction, unifying the results observed on various tokamaks. For high density (ne ≥ 1.2 × 1020 m−3) L-mode target discharges, where core LH wave absorption is low, the rotation change is in the co-current direction, but evolves on a shorter momentum transport time scale, and is seen across the entire spatial profile. For H-mode target plasmas, both co- and counter-current direction increments have been observed with LHRF. The H-mode co-rotation is correlated with the pedestal temperature gradient, which itself is enhanced by the LH waves absorbed in the plasma periphery. The H-mode counter-rotation increment, a flattening of the peaked velocity profile in the core, is consistent with a reduction in the momentum pinch correlated with a steepening of the core density profile. Most of these rotation changes must be due to indirect transport effects of LH waves on various parameters, which modify the momentum flux.

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.

In the absence of an internal particle source, plasma turbulence will impose an intrinsic relationship between an inwards pinch and an outwards diffusion resulting in a stationary density profile. The Alcator C-mod tokamak utilizes RF heating and current drive so that fueling only occurs in the vicinity of the separatrix. Discharges that transition from L-mode to I-mode are seen to maintain a self-similar stationary density profile as measured by Thomson scattering. For discharges with negative magnetic shear, an observed rise of the safety factor in the vicinity of the magnetic axis appears to be accompanied by a decrease of electron density, qualitatively consistent with the theoretical expectations. © 2012 American Institute of Physics., United States. Department of Energy. Office of Fusion Energy Sciences

New experimental observations are reported on the structure and dynamics of short-lived periodic (1, 1) “fishbone”-like oscillations that appear during radio frequency heating and current-drive experiments in tokamak plasmas. For the first time, measurements can directly relate changes in the high energy electrons to the mode onset, saturation, and damping. In the relatively high collisionality of Alcator C-Mod with lower hybrid current drive, the instability appears to be destabilized by the non-resonant suprathermal electron pressure—rather than by wave-particle resonance, rotates toroidally with the plasma and grows independently of the (1, 1) sawtooth crash driven by the thermal plasma pressure.

Recent progress on lower hybrid current drive (LHCD) experiment and simulation towards steady-state (t ⩾ 3–5 × τr, where τr is the current relaxation time) regimes on Alcator C-Mod is presented. Highly non-inductive reversed shear plasmas are obtained with spontaneous generation of internal transport barriers at the density close to what is expected on ITER steady-state scenarios. Progress has been made to better understand and mitigate the unexpected degradation of LHCD efficiency at high densities, which poses an issue both for extending the non-inductive plasmas to advanced tokamak regimes (with a high bootstrap current fraction, fBS) on C-Mod and for predicting the performance of LHCD on future devices such as ITER. Several physics mechanisms that potentially contribute anomalous losses of LHCD power have been studied extensively. Numerical modelling of collisional absorption in cold scrape-off layer plasmas has been integrated into a ray-tracing code. The LHEAF full-wave code clarifies that full-wave effects move the power deposition profile closer to the separatrix than a calculation based on the WKB approximation, leading to a lower efficiency. Non-linear wave interactions were studied experimentally by LH wave spectral measurements using Langmuir probes, suggesting parametric decay instabilities to explain the remaining difference between experiments and simulations. An experimental demonstration of recovery of good LHCD efficiency at high densities is also reported. Improving the single pass absorption is proposed as a key to recover LHCD efficiency. A wave physics design of additional launcher (LH3) has been developed in order to demonstrate an improved LHCD performance by realizing high (∼80%) single pass absorption, in which the synergistic interaction in the velocity space with the existing launcher is maximized.

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.

The design, construction and initial results of a new lower hybrid current drive (LHCD) launcher on Alcator C-Mod (Hutchinson et al 1994 Phys. Plasmas 1 1511) are presented. The new LHCD launcher (LH2) is based on a novel splitter concept which evenly distributes the microwave power in four ways in the poloidal direction. This design allows for simplification of the feeding structure while keeping the flexibility to vary the peak launched toroidal index of refraction, Ntoroidal, from ?3.8 to 3.8. An integrated model predicts good plasma coupling over a wide range of edge densities, while poloidal variations of the edge density are found to affect the evenness of power splitting in the poloidal direction. The measured transmission loss is about 30% lower than the previous launcher, and a clean Ntoroidal spectrum has been confirmed. Power handling capability exceeding an empirical weak conditioning limit and reliable operation up to 1.1?MW net LHCD power have been achieved. A survey of antenna?plasma coupling shows the existence of a millimetric vacuum gap in front of the launcher. Fully non-inductive, reversed shear plasma operation has been demonstrated and sustained for multiple current diffusion times. The current drive efficiency, ?LH ? neR0Ip/PLH, of these plasmas is (0.2?0.25) ? 1020?m?2A?W?1, which is in agreement with the expected efficiency on the International Thermonuclear Experimental Reactor (ITER).