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A synthetic gas puff imaging (GPI) diagnostic has been developed for the purpose of validating the three-dimensional gyrokinetic turbulence code XGC. The synthetic diagnostic is described and applied to XGC simulations of two Alcator C-Mod discharges. The turbulence characteristics deduced from the resulting simulated GPI frames, using analysis techniques similar to those applied to experimental data, are compared with locally derived characteristics extracted directly from the XGC output. The comparison of the two is shown to be potentially impacted by misalignment between the GPI view and the magnetic field, the dependence of the light emission on the electron density and temperature, and spatial and temporal variations in the neutral gas cloud induced by the turbulent plasma fluctuations. We conclude that quantitative and, in some cases even qualitative, validation of turbulence simulations need to account for these effects. While we cannot directly compare our results with experimental data due to the absence of high quality GPI data from the shots and times simulated by XGC, we do relate the overall characteristics obtained from the synthetic GPI analysis to published Alcator C-Mod GPI data.

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

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

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

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

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

High resolution measurements on the Alcator C-Mod tokamak [I. H. Hutchinson et al., Phys. Plasmas 1, 1551 (1994)] of the transport barrier in the “Enhanced Dα” (EDA) regime, which has increased particle transport without large edge localized modes, show steep density and temperature gradients over a region of 2–5 mm, with peak pressure gradients up to 12 MPa/m. Evolution of the pedestal at the L-H transition is consistent with a large, rapid drop in thermal conductivity across the barrier. A quasi-coherent fluctuation in density, potential, and Bpol, with f0∼50–150 kHz and kθ∼4 cm−1, always appears in the barrier during EDA, and drives a large particle flux. Conditions to access the steady-state EDA regime in deuterium include δ>0.35, q95>3.5, and L-mode target density ne>1.2×1020 m−3. A reduced q95 limit is found for hydrogen discharges.

Neutral particle densities and energy losses have been measured in the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994)]. Their effect on the formation and evolution of the edge barrier which accompanies the enhanced confinement regime are discussed. The neutrals can enter the edge dynamics through the particle, momentum, and energy balance. Neutral densities of up to 5×1016 m−3 have been measured in the edge barrier region. Neutrals enter the local dynamics around most of the periphery, not just at the X-point. High resolution measurements of the ionization profile have been obtained for the region near the separatrix. The profile shifts inside the separatrix as the plasma is making a transition from low-to high-mode confinement (H-mode) regimes, partly accounting for the dramatic rise in edge density. The measured neutral density is large enough to affect the bulk ion momentum by charge exchange, and thereby introduces a negative radial electric field at the edge. At the same time, significant edge heat flux, carried by the neutrals, contributes to the measured power loss. At very high edge densities, this loss mechanism could contribute to quenching H-modes.

The effect of divertor baffling on Alcator C-Mod [Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] discharges has been investigated using a novel divertor bypass. The bypass allows the in situ variation of the mechanical conductance from the divertor plenum to the main chamber between or even during discharges. The results indicate a strong dependence on baffling for the compression of deuterium and recycling impurity gases in the divertor plenum in a variety of C-Mod discharges, indicating that tight baffling would be desirable in a future tokamak reactor. There is no apparent effect of the divertor baffling for the Ohmic, L-mode and EDA (Enhanced Dα) H-mode discharges in this study on main chamber neutral particle populations, scrape-off-layer flows, nonrecycling impurity screening, global energy confinement or H-mode threshold powers.

In the present study recycling and transport in the edge plasma of Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] is modeled and analyzed with the multi-fluid code UEDGE [T. D. Rognlien et al., J. Nucl. Mater. 196–198, 347 (1992)]. Matching the experimental plasma density profiles in the scrape-off layer (SOL) requires a spatially dependent effective anomalous diffusion coefficient D⊥ growing rapidly towards the wall. The midplane pressure of neutral gas, Pmid, is a key parameter that reflects the magnitude of anomalous transport of plasma from the core. Recycling of plasma on the main chamber wall appears to be quite significant, especially in the case of high Pmid∼0.3 mTorr when the main wall provides ∼70% of recycling neutrals in the main chamber. In the upper SOL (well above the x point) draining of particles by the poloidal flow is weak and thus the particle balance is predominantly radial. For the radial heat transport it is found that energy flux carried by radial plasma conv...

Detailed measurements of the magnitude and location of volumetric recombination occurring in the detached divertor of Alcator C-Mod tokamak [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] are presented. The drop in divertor plate ion current during detachment is due to two mechanisms: (1) volumetric recombination in the divertor plasma; and (2) reductions in the divertor ion source. Depending on plasma conditions, each of these can be the primary mechanism for the observed ion current reduction in detachment. The ion source during detachment is inferred and its magnitude is consistent with the measured divertor power flow. A scaling of the density in the divertor recombining region for L- (low confinement) mode plasmas is found, ne,r∝ne0.8⋅PSOL2/7. A model based on pressure variation along a flux surface during detachment is consistent with the main features of this scaling.

Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] has operated a High-confinement-mode (H-mode) plasma together with a dissipative divertor and low core Zeff. The initially attached plasma is characterized by steady-state enhancement factor, HITER89P [P. N. Yushmanov et al., Nucl. Fusion 30, 1999 (1990)], of 1.9, central Zeff of 1.1, and a radiative fraction of ∼50%. Feedback control of a nitrogen gas puff is used to increase radiative losses in both the core/edge and divertor plasmas in almost equal amounts. Simultaneously, the core plasma maintains HITER89P of 1.6 and Zeff of 1.4 in this nearly 100% radiative state. The power and particle flux to the divertor plates have been reduced to very low levels while the core plasma is relatively unchanged by the dissipative nature of the divertor.

Regimes of high-confinement mode have been studied in the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994)]. Plasmas with no edge localized modes (ELM-free) have been compared in detail to a new regime, enhanced Dα (EDA). EDA discharges have only slightly lower energy confinement than comparable ELM-free ones, but show markedly reduced impurity confinement. Thus EDA discharges do not accumulate impurities and typically have a lower fraction of radiated power. The edge gradients in EDA seem to be relaxed by a continuous process rather than an intermittent one as is the case for standard ELMy discharges and thus do not present the first wall with large periodic heat loads. This process is probably related to fluctuations seen in the plasma edge. EDA plasmas are more likely at low plasma current (q>3.7), for moderate plasma shaping, (triangularity ∼0.35–0.55), and for high neutral pressures. As observed in soft x-ray emission, the pedestal width is found to scale with the same parameters that determine the EDA/ELM-free boundary.

Analysis of the experimental data from tokamaks and linear divertor simulators leads to the conclusion that plasma recombination is a crucial element of plasma detachment. Different mechanisms of plasma recombination relevant to the experimental conditions of the tokamak scrape-off layer (SOL) and divertor simulators are considered. The physics of Molecular Activated Recombination (MAR) involving vibrationally excited molecular hydrogen are discussed. Although conventional Electron–Ion Recombination (EIR) alone can strongly alter the plasma parameters, MAR impact can be substantial for both tokamak SOL plasma and divertor simulators. Investigation of the effects of EIR on the plasma flow in divertor simulators shows that due to the balances of (a) energy transport and electron cooling, and (b) the plasma flow and recombination, that EIR extinguishes the simulator plasma at an electron temperature about 0.15 eV.

A series of experiments has been carried out to determine the relative screening of the low recycling gaseous impurities, nitrogen and methane, in Alcator C-Mod tokamak [Phys. Plasmas 1, 1511 (1994)] discharges. The impurity density in the core plasma is directly proportional to the rate of impurity injection. Screening was found to vary weakly with plasma density but to be markedly dependent on the poloidal position of injection. Detachment of the plasma results in significantly less screening both in Ohmic and in high confinement (H-mode) discharges, particularly when the gas is injected into the divertor or at the inner vessel wall. Evidence of impurity transport from the scrape-off layer into the divertor using bolometry and visible spectroscopy is presented and modeling of the transport using a two-dimensional Monte Carlo code is described.

Highly localized direct electron heating [full width at half-maximum (FWHM)≅0.2a] via mode converted ion Bernstein waves has been observed in the Alcator C-Mod Tokamak [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)]. Electron heating at or near the plasma center (r/a⩾0.3) has been observed in H(3He) discharges at B0=(6.0–6.5) T and ne(0)≅1.8×1020 m−3. [Here, the minority ion species is indicated parenthetically.] Off-axis heating (r/a⩾0.5) has also been observed in D(3He) plasmas at B0=7.9 T. The concentration of 3He in these experiments was in the range of n3He/ne≅(0.2–0.3) and the locations of the mode conversion layer and electron heating peak could be controlled by changing the 3He concentration or toroidal magnetic field (B0). The electron heating profiles were deduced using a rf modulation technique. Detailed comparisons with one-dimensional and toroidal full-wave models in the ion cyclotron range of frequencies have been carried out. One-dimensional full-wave code predictions were found to ...

Enhanced confinement modes up to a toroidal field of BT=8 T have been studied with up to 3.5 MW of radiofrequency (rf) heating power in the ion cyclotron range of frequencies (ICRF) at 80 MHz. H-mode is observed when the edge temperature exceeds a threshold value. The high confinement mode (H-mode) with higher confinement enhancement factors (H) and longer duration became possible after boronization by reducing the radiated power from the main plasma. A quasi-steady state with high confinement (H=2.0), high normalized beta (βN=1.5), low radiated power fraction (Pradmain/Ploss=0.3), and low effective charge (Zeff=1.5) has been obtained in Enhanced Dα H-mode. This type of H-mode has enhanced levels of continuous Dα emission and very little or no edge localized mode (ELM) activity, and reduced core particle confinement time relative to ELM-free H-mode. The pellet enhanced performance (PEP) mode is obtained by combining core fueling with pellet injection and core heating. A highly peaked pressure profile with a central value of 8 atmospheres was observed. The steep pressure gradient drives off-axis bootstrap current, resulting in a shear reversed safety factor (q) profile. Suppression of sawteeth appears to be important in maintaining the highly peaked pressure profile. Lithium pellets were found to be more effective than deuterium pellets in raising q0.

The regime of high particle and energy confinement known as the H mode [Phys. Rev. Lett. 49, 1408 (1982)] has been extended to a unique range of operation for divertor tokamaks up to toroidal fields of nearly 8 T, line‐averaged electron densities of 3×1020 m−3, and surface power densities of nearly 0.6 MW/m2 in the compact high‐field tokamak Alcator C Mod [Phys. Plasmas 1, 1511 (1994)]. H modes are achieved in Alcator C Mod with Ion Cyclotron Resonant Frequency (ICRF) heating and with Ohmic heating alone without boronization of the all molybdenum tiled first wall. Large increases in charge exchange flux are observed during the H mode over the entire range of energies from 2 to 10 keV. There appears to be an upper limit to the midplane neutral pressure, of about 0.08 Pa above which no H modes have been observed. The plasmas with the best energy confinement have the lowest midplane neutral pressures, below 0.01 Pa. There is an edge electron temperature threshold such that Te≥280 eV ±40 eV for sustaining the...

The divertor of the Alcator C‐Mod tokamak [Phys. Plasmas 1, 1511 (1994)] routinely radiates a large fraction of the power entering the scrape‐off layer. This dissipative divertor operation occurs whether the divertor is detached or not, and large volumetric radiative emissivities, up to 60 MW m−3 in ion cyclotron range of frequency (ICRF) heated discharges, have been measured using bolometer arrays. An analysis of both Ohmic and ICRF‐heated discharges has demonstrated some of the relative merits of detached divertor operation versus high‐recycling divertor operation. An advantage of detached divertor operation is that the power flux to the divertor plates is decreased even further than its already low value. Some disadvantages are that volumetric losses outside the separatrix in the divertor region are decreased, the neutral compression ratio is decreased, and the penetration efficiency of impurities increases.

The electron density in the edge, divertor and X‐point regions of Alcator C‐Mod [Proceedings of the IEEE 13th Symposium on Fusion Engineering (Institute of Electrical and Electronic Engineers, New York, 1990), Vol. 1, p. 13] has been measured using the Stark broadening of high‐n (n=8 through 11) transitions of the Balmer series of deuterium. These measurements have been made during typical single null, diverted operation and during gas‐puffing through capillaries located within the first wall and divertor. Electron densities up to 1.6×1021 m−3 have been measured in the X‐point region and slightly lower densities (5−7×1020 m−3) have been measured in the divertor and edge regions. These results are factors of 2 to 5 larger than the density measurements from the CO2 laser interferometer system in the main chamber (3×1020 m−3) and comparable with probe measurements in the divertor. The ratios of the widths of adjacent lines within the Balmer series have been compared with ratios calculated for the Stark broad...

In the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] a substantial improvement in fusion performance has been realized by combining the enhanced confinement due to tritium fueling with the enhanced confinement due to extensive conditioning of the limiter with lithium. This combination has resulted in not only significantly higher global energy confinement times than have previously been obtained in high current supershots, but also in the highest central ratio of thermonuclear fusion output power to input power observed to date.

Detailed measurements and transport analysis of divertor conditions in Alcator C‐Mod [Phys. Plasmas 1, 1511 (1994)] are presented for a range of line‐averaged densities, 0.7

After many years of fusion research, the conditions needed for a D–T fusion reactor have been approached on the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. For the first time the unique phenomena present in a D–T plasma are now being studied in a laboratory plasma.The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral‐beam heating, in a supershot discharge and 6.7 MW in a high‐βp discharge following a current rampdown. The fusion power density in a core of the plasma is ≊2.8 MW m−3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) [Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 239] at 1500 MW total fusion power. The energy confinement time, τE, is observed to increase in D–T, relative to D plasmas, by 20% and the ni(0) Ti(0) τE product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter‐H‐mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high‐βp discharges. Ion cyclotron range of frequencies (ICRF) heating of a D–T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP [Nucl. Fusion 34, 1247 (1994)] simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is well described by the first‐orbit loss mechanism. No loss due to alpha‐particle‐driven instabilities has yet been observed. D–T experiments on TFTR will continue to explore the assumptions of the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.

Two-dimensional scrape-off layer turbulence (SOLT) code simulations are compared with an L-mode discharge on the Alcator C-Mod tokamak [Greenwald et al., Phys. Plasmas 21, 110501 (2014)]. Density and temperature profiles for the simulations were obtained by smoothly fitting Thomson scattering and mirror Langmuir probe (MLP) data from the shot. Simulations differing in turbulence intensity were obtained by varying a dissipation parameter. Mean flow profiles and density fluctuation amplitudes are consistent with those measured by MLP in the experiment and with a Fourier space diagnostic designed to measure poloidal phase velocity. Blob velocities in the simulations were determined from the correlation function for density fluctuations, as in the analysis of gas-puff-imaging (GPI) blobs in the experiment. In the simulations, it was found that larger blobs moved poloidally with the E × B flow velocity, vE, in the near-SOL, while smaller fluctuations moved with the group velocity of the dominant linear (interc...

The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a FluorinertTM system, modification of the vacuum system to handle tritium, preparation, and testing of the neutral beam system for tritium operation and a final deuterium–deuterium (D–D) run to simulate expected deuterium–tritium (D–T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D–T experiments using D–D have been performed. The physics objectives of D–T operation are production of ≊10 MW of fusion power, evaluation of confinement, and heating in deuterium–tritium plasmas, evaluation of α‐particle heating of electrons, and collective effects driven by alpha particles and testing of diag...

Main ion dilution has been predicted by gyrokinetic simulations to have a significant effect on ion thermal transport in C-Mod ohmic plasmas. This effect was verified experimentally with a specific set of experiments on C-Mod in which ohmic deuterium plasmas across the linear ohmic confinement (LOC) through the saturated ohmic confinement (SOC) regimes were diluted by seeding with nitrogen gas (Z = 7) injection. The seeding was observed to increase the normalized ion temperature gradients (ITGs) by up to 30% without a corresponding increase in the gyrobohm normalized ion energy flux, indicating a change in either the stiffness or the critical ion temperature gradient associated with ITG turbulence. The seeding also reversed the direction of the intrinsic toroidal rotation in plasmas slightly above the normal intrinsic rotation reversal critical density. GYRO simulations of the seeded and unseeded plasmas show that the seeding affected both the critical gradient and the stiffness. For plasmas in the LOC re...

The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of critical physical models into new parameter ranges and into new regimes. Using only high-power radio frequency (RF) waves for heating and current drive with innovative launching structures, C-Mod operates routinely at reactor level power densities and achieves plasma pressures higher than any other toroidal confinement device. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components—approaches subsequently adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and the Enhanced Dα H-mode regimes, which have high performance without large edge localized modes and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and demonstrated that self-generated flow shear can be strong enough in some cases to significantly modify transport. C-Mod made the first quantitative link between the pedestal temperature and the H-mode's performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. RF research highlights include direct experimental observation of ion cyclotron range of frequency (ICRF) mode-conversion, ICRF flow drive, demonstration of lower-hybrid current drive at ITER-like densities and fields and, using a set of novel diagnostics, extensive validation of advanced RF codes. Disruption studies on C-Mod provided the first observation of non-axisymmetric halo currents and non-axisymmetric radiation in mitigated disruptions. A summary of important achievements and discoveries are included., United States. Dept. of Energy (Cooperative Agreement DE-FC02-99ER54512), United States. Dept. of Energy (Cooperative Agreement DE-FG03-94ER-54241), United States. Dept. of Energy (Cooperative Agreement DE-AC02-78ET- 51013), United States. Dept. of Energy (Cooperative Agreement DE-AC02-09CH11466), United States. Dept. of Energy (Cooperative Agreement DE-FG02-95ER54309), United States. Dept. of Energy (Cooperative Agreement DE-AC02-05CH11231), United States. Dept. of Energy (Cooperative Agreement DE-AC52-07NA27344), United States. Dept. of Energy (Cooperative Agreement DE-FG02- 97ER54392), United States. Dept. of Energy (Cooperative Agreement DE-SC00-02060)

A novel “Shoelace” antenna has been used to inductively excite a short-wavelength edge fluctuation in a tokamak boundary layer for the first time. The principal design parameters, k[subscript ⊥] = 1.5 ± 0.1 cm[superscript −1] and 45 < f < 300 kHz, match the Quasi-Coherent Mode (QCM, k[subscript ⊥] ∼ 1.5 cm[superscript −1], f ∼ 50−150 kHz) in Alcator C-Mod, responsible for exhausting impurities in the steady-state, ELM-free Enhanced D[subscript α] H-mode. In H-mode, whether or not there is a QCM, the antenna drives coherent, field-aligned perturbations in density, [˜ over n][subscript e], and field, [˜ over B][subscript θ], which are guided by field lines, propagate in the electron diamagnetic drift direction, and exhibit a weakly damped (γ/ω[subscript 0] ∼ 5%−10%) resonance near the natural QCM frequency. This result is significant, offering the possibility that externally driven modes may be used to enhance particle transport. In L-mode, the antenna drives only a non-resonant [˜ over B][subscript θ] response. The facts that the driven mode has the same wave number and propagation direction as the QCM, and is resonant at the QCM frequency, suggest the antenna may couple to this mode, which we have shown elsewhere to be predominantly drift-mode-like [B. LaBombard et al., Phys. Plasmas 21, 056108 (2014)]., United States. Dept. of Energy (Cooperative Agreement DE-FC02-99ER54512)

This paper describes 2D imaging measurements of plasma turbulence made in the scrape-off layer of the Alcator C-Mod tokamak simultaneously at two different poloidal locations, one near the outer midplane and the other near the divertor X-point region. These images were made with radial and poloidal resolution using two gas puff imaging diagnostics not directly connected along a B field line. The turbulence correlation structure has a significantly different tilt angle with respect to the local flux surfaces for the midplane and X-regions, and a slightly different ellipticity and size. The time-averaged turbulence velocities can be different in the midplane and X-regions, even within the same flux surface in the same shot. The structures are partially consistent with a magnetic flux tube mapping model, and the velocities are compared with various models for turbulence flow.

Measurements of poloidal variation, nz/⟨nz⟩, in high-Z impurity density have been made using photodiode arrays sensitive to vacuum ultraviolet and soft x-ray emission in Alcator C-Mod plasmas. In/out asymmetries in the range of −0.2 0, of a flux surface is found to be well described by a combination of centrifugal, poloidal electric field, and ion-impurity friction effects. Up/down asymmetries, −0.05 0 corresponding to accumulation opposite the ion ∇B drift direction. Measurements of the up/down asymmetry of molybdenum are found to disagree with predictions from recent neoclassical theory in the trace limit, nzZ2/ni≪1. Non-trace levels of impurities are expected to modify the main-ion poloidal flow and thus change friction-driven impurity density asymmetries and impurity poloidal rotation, vθ,z. Artificially modifying main...

Ion cyclotron range of frequency (ICRF) heating is expected to provide auxiliary heating for ITER and future fusion reactors where high Z metallic plasma facing components (PFCs) are being considered. Impurity contamination linked to ICRF antenna operation remains a major challenge particularly for devices with high Z metallic PFCs. Here, we report on an experimental investigation to test whether a field aligned (FA) antenna can reduce impurity contamination and impurity sources. We compare the modification of the scrape of layer (SOL) plasma potential of the FA antenna to a conventional, toroidally aligned (TA) antenna, in order to explore the underlying physics governing impurity contamination linked to ICRF heating. The FA antenna is a 4-strap ICRF antenna where the current straps and antenna enclosure sides are perpendicular to the total magnetic field while the Faraday screen rods are parallel to the total magnetic field. In principle, alignment with respect to the total magnetic field minimizes integrated E|| (electric field along a magnetic field line) via symmetry. A finite element method RF antenna model coupled to a cold plasma model verifies that the integrated E|| should be reduced for all antenna phases. Monopole phasing in particular is expected to have the lowest integrated E||. Consistent with expectations, we observed that the impurity contamination and impurity source at the FA antenna are reduced compared to the TA antenna. In both L and H-mode discharges, the radiated power is 20%–30% lower for a FA-antenna heated discharge than a discharge heated with the TA-antennas. However, inconsistent with expectations, we observe RF induced plasma potentials (via gas-puff imaging and emissive probes to be nearly identical for FA and TA antennas when operated in dipole phasing). Moreover, the highest levels of RF-induced plasma potentials are observed using monopole phasing with the FA antenna. Thus, while impurity contamination and sources are indeed reduced with the FA antenna configuration, the mechanism determining the SOL plasma potential in the presence of ICRF and its impact on impurity contamination and sources remains to be understood.

Reduced-model scrape-off layer turbulence (SOLT) simulations of an enhanced D-alpha (EDA) H-mode shot observed in the Alcator C-Mod tokamak were conducted to compare with observed variations in the scrape-off-layer (SOL) width of the parallel heat flux profile. In particular, the role of the competition between sheath- and conduction-limited parallel heat fluxes in determining that width was studied for the turbulent SOL plasma that emerged from the simulations. The SOL width decreases with increasing input power and with increasing separatrix temperature in both the experiment and the simulation, consistent with the strong temperature dependence of the parallel heat flux in balance with the perpendicular transport by turbulence and blobs. The particularly strong temperature dependence observed in the case analyzed is attributed to the fact that these simulations produce SOL plasmas which are in the conduction-limited regime for the parallel heat flux. A persistent quasi-coherent (QC) mode dominates the SOLT simulations and bears considerable resemblance to the QC mode observed in C-Mod EDA operation. The SOLT QC mode consists of nonlinearly saturated wave-fronts located just inside the separatrix that are convected poloidally by the mean flow, continuously transporting particles and energy and intermittently emitting blobs into the SOL.

A coordinated effort to measure divertor heat flux characteristics in fully attached, similarly shaped H-mode plasmas on C-Mod, DIII-D, and NSTX was carried out in 2010 in order to construct a predictive scaling relation applicable to next step devices including ITER, FNSF, and DEMO. Few published scaling laws are available and those that have been published were obtained under widely varying conditions and divertor geometries, leading to conflicting predictions for this critically important quantity. This study was designed to overcome these deficiencies. Analysis of the combined data set reveals that the primary dependence of the parallel heat flux width is robustly inverse with Ip, which all three tokamaks independently demonstrate. An improved Thomson scattering system on DIII-D has yielded very accurate scrape off layer (SOL) profile measurements from which tests of parallel transport models have been made. It is found that a flux-limited model agrees best with the data at all collisionalities, while...

Experiments in Alcator C-Mod in (Enhanced D-alpha) EDA H-modes with extrinsic impurity seeding (N2, Ne, and Ar) have demonstrated a direct correlation between plasma energy confinement and edge power flow, achieving values of H98 ≥ 1 for edge power flows only marginally exceeding the scaled power for access to H-mode confinement in these conditions. For lower Z impurity seeding (N2 and Ne), plasmas with high energy confinement are obtained with a radiative power fraction of 85% or larger and a reduction of the peak heat flux at the divertor by more than a factor of 5 compared to similar attached conditions. The H-mode plasmas thus achieved in Alcator C-Mod meet or exceed the requirements both in terms of divertor heat flux handling and energy confinement for ITER QDT = 10 operation and with an edge power flow only marginally above the H-mode threshold power (by 1.0–1.4) as expected in ITER.

We report extended studies of the I-mode regime [Whyte et al., Nucl. Fusion 50, 105005 (2010)] obtained in the Alcator C-Mod tokamak [Marmar et al., Fusion Sci. Technol. 51(3), 3261 (2007)]. This regime, usually accessed with unfavorable ion B × ∇B drift, features an edge thermal transport barrier without a strong particle transport barrier. Steady I-modes have now been obtained with favorable B × ∇B drift, by using specific plasma shapes, as well as with unfavorable drift over a wider range of shapes and plasma parameters. With favorable drift, power thresholds are close to the standard scaling for L–H transitions, while with unfavorable drift they are ∼ 1.5–3 times higher, increasing with Ip. Global energy confinement in both drift configurations is comparable to H-mode scalings, while density profiles and impurity confinement are close to those in L-mode. Transport analysis of the edge region shows a decrease in edge χeff, by typically a factor of 3, between L- and I-mode. The decrease correlates with ...

Parametric dependences of the heat flux footprint on the outer divertor target plate are explored in EDA H-mode and ohmic L-mode plasmas over a wide range of parameters with attached plasma conditions. Heat flux profile shapes are found to be independent of toroidal field strength, independent of power flow along magnetic field lines and insensitive to x-point topology (single-null versus double-null). The magnitudes and widths closely follow that of the “upstream” pressure profile, which are correlated to plasma thermal energy content and plasma current. Heat flux decay lengths near the strike-point in H- and L-mode plasmas scale approximately with the inverse of plasma current, with a diminished dependence at high collisionality in L-mode. Consistent with previous studies, pressure gradients in the boundary scale with plasma current squared, holding the magnetohydrodynamic ballooning parameter approximately invariant at fixed collisionality—strong evidence that critical-gradient transport physics plays a key role in setting the power exhaust channel.

The steep gradient edge region and scrape-off-layer (SOL) on the low-field-side of Alcator C-Mod [I. H. Hutchinson, R. Boivin, F. Bombarda et al., Phys. Plasmas 1, 1511 (1994)] tokamak plasmas are studied using gas-puff-imaging diagnostics. In L-mode plasmas, the region extending ∼2 cm inside the magnetic separatrix has fluctuations showing a broad, turbulent spectrum, propagating in the electron diamagnetic drift direction, whereas features in the open field line region propagate in the ion diamagnetic drift direction. This structure is robust against toroidal field strength, poloidal null-point geometry, plasma current, and plasma density. Global parameter dependence of spectral and spatial structure of the turbulence inside the separatrix is explored and characterized, and both the intensity and spectral distributions are found to depend strongly on the plasma density normalized to the tokamak density limit. In H-mode discharges the fluctuations at and inside the magnetic separatrix show fundamentally ...

This paper describes quantitative comparisons between turbulence measured in the scrape-off layer (SOL) of Alcator C-Mod [S. Scott, A. Bader, M. Bakhtiari et al., Nucl. Fusion 47, S598 (2007)] and three dimensional computations using electromagnetic gyrofluid equations in a two-dimensional tokamak geometry. These comparisons were made for the outer midplane SOL for a set of inner-wall limited, near-circular Ohmic plasmas. The B field and plasma density were varied to assess gyroradius and collisionality scaling. The poloidal and radial correlation lengths in the experiment and computation agreed to within a factor of 2 and did not vary significantly with either B or density. The radial and poloidal propagation speeds and the frequency spectra and poloidal k-spectra also agreed fairly well. However, the autocorrelation times and relative Da fluctuation levels were higher in the experiment by more than a factor of 2. Possible causes for these disagreements are discussed. 2009 American Institute of Physics.

Recent experiments have led to a fundamental shift in our view of edge transport physics; transport near the last-closed flux surface may be more appropriately described in terms of a critical gradient phenomenon rather than a diffusive and/or convective paradigm. Edge pressure gradients, normalized by the square of the poloidal magnetic field strength, appear invariant in plasmas with the same normalized collisionality, despite vastly different currents and magnetic fields—a behavior that connects with first-principles electromagnetic plasma turbulence simulations. Near-sonic scrape-off layer (SOL) flows impose a cocurrent rotation boundary condition on the confined plasma when B×∇B points toward the active x-point, suggesting a link to the concomitant reduction in input power needed to attain high-confinement modes. Indeed, low-confinement mode plasmas are found to attain higher edge pressure gradients in this configuration, independent of the direction of B, evidence that SOL flows may affect transport and “critical gradient” values in the edge plasma.

This paper reports on studies of the edge transport barrier and transition threshold of the high confinement (H) mode of operation on the Alcator C-Mod tokamak [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)], over a wide range of toroidal field (2.6–7.86T) and plasma current (0.4–1.7MA). The H-mode power threshold and edge temperature at the transition increase with field. Barrier widths, pressure limits, and confinement are nearly independent of field at constant current, but the operational space at high B shifts toward higher temperature and lower density and collisionality. Experiments with reversed field and current show that scrape-off-layer flows in the high-field side depend primarily on configuration. In configurations with the B×∇B drift away from the active X-point, these flows lead to more countercurrent core rotation, which apparently contributes to higher H-mode thresholds. In the unfavorable case, edge temperature thresholds are higher, and slow evolution of profiles indicates a red...

In this paper we compare the structure and motion of edge turbulence observed in L-mode vs. H-mode plasmas in the National Spherical Torus Experiment (NSTX) [M. Ono, M. G. Bell, R. E. Bell et al., Plasma Phys. Controlled Fusion 45, A335 (2003)]. The radial and poloidal correlation lengths are not significantly different between the L-mode and the H-mode in the cases examined. The poloidal velocity fluctuations are lower and the radial profiles of the poloidal turbulence velocity are somewhat flatter in the H-mode compared with the L-mode plasmas. These results are compared with similar measurements Alcator C-Mod [E. Marmar, B. Bai, R. L. Boivin et al., Nucl. Fusion 43, 1610 (2003)], and with theoretical models.

Factor of ∼2 higher power thresholds for low- to high-confinement mode transitions (L-H) with unfavorable x-point topologies in Alcator C-Mod [Phys. Plasmas 1, 1511 (1994)] are linked to flow boundary conditions imposed by the scrape-off layer (SOL). Ballooning-like transport drives flow along magnetic field lines from low- to high-field regions with toroidal direction dependent on upper/lower x-point balance; the toroidal rotation of the confined plasma responds, exhibiting a strong counter-current rotation when B×∇B points away from the x point. Increased auxiliary heating power (rf, no momentum input) leads to an L-H transition at approximately twice the edge electron pressure gradient when B×∇B points away. As gradients rise prior to the transition, toroidal rotation ramps toward the co-current direction; the H mode is seen when the counter-current rotation imposed by the SOL flow becomes compensated. Remarkably, L-H thresholds in lower-limited discharges are identical to lower x-point discharges; SOL...

The recently upgraded phase-contrast imaging (PCI) diagnostic is used to characterize the transition from the low (L) to the enhanced Dα (EDA) high (H) confinement mode in Alcator C-Mod [I. H. Hutchinson, R. Boivin, F. Bombarda et al., Phys. Plasmas 1, 1511 (1994)] plasmas. PCI yields information on line integrated density fluctuations along vertical chords. The number of channels has been increased from 12 to 32 and the sampling rate from 1 MHz to 10 MHz. This expansion of diagnostic capabilities is used to study broadband turbulence in L and EDA H mode and to analyze the quasicoherent (QC) mode associated with EDA H mode. Changes in broadband turbulence at the transition from L to EDA H mode can be interpreted as an effect of the Doppler rotation of the bulk plasma. Additional fluctuation measurements of Dα light and the poloidal magnetic field show features correlated with PCI in two different frequency ranges at the transition. The backtransition from EDA H to L mode, the so-called enhanced neutron (E...

Spectral line shifts have been observed in the Dα transition of neutral deuterium in the Alcator C-mod [I. H. Hutchinson, Proceedings of the IEEE 13th Symposium on Fusion Engineering (IEEE, Piscataway, 1990), Vol. 1, p. 13] tokamak and interpreted as Doppler shifts due to a neutral flow of approximately 5×103 m/s. The emission originates from plasma at the inner wall, as indicated by the Zeeman patterns of the transition. The direction of the flow changes when the plasma transitions from a limited to a diverted configuration. In all cases, the flows are directed toward the point of contact with the wall or divertor plate. Narrow spectral profiles have been observed for the Dα transition near the midplane and within the divertor. Temperatures not higher than 0.86 eV are indicated for the divertor measurements. A difference in plasma conditions for separate locations along a single line of sight has been observed.

A region of steep plasma gradients, the so-called "narrow-feature," has been found in the near scrape-off layer (SOL) of inner-wall limited (IWL) discharges. Dedicated IWL discharges were carried out in Alcator C-Mod [Marmar et al., Nucl. Fusion 55, 104020 (2015)] to study this phenomenon, allowing detailed observations of the plasma profiles and fluctuations. Langmuir probe (LP) measurements show a clear two-decay length ne and T-e profile structure at the outer midplane. The Gas-Puff Imaging (GPI) diagnostic shows large turbulent fluctuations across the last closed flux-surface, hence supporting the hypothesis that turbulent phenomena play a role in setting the profile steepness. Flux-driven non-linear turbulence simulations of two C-Mod discharges have been carried out, allowing a three-way comparison between LP, GPI, and simulation data. Observations and simulations correlate the steep gradient region characterizing the narrow feature with sheared poloidal flows and a deviation of the plasma potential from its floating value. The E x B shear rate exceeds the linear ballooning growth rate, indicating that the narrow feature could result from the effects of sheared flows, although causality could not be established. The fluctuation level in the narrow feature remains of order unity across the entire SOL, indicating that the transport reduction in the near-SOL cannot result from a simple quench rule. Published by AIP Publishing.