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New validation of global, nonlinear, ion-scale gyrokinetic simulations (GYRO) is carried out for L- and I-mode plasmas on Alcator C-Mod, utilizing heat fluxes, profile stiffness, and temperature fluctuations. Previous work at C-Mod found that ITG/TEM-scale GYRO simulations can match both electron and ion heat fluxes within error bars in I-mode [White PoP 2015], suggesting that multi-scale (cross-scale coupling) effects [Howard PoP 2016] may be less important in I-mode than in L-mode. New results presented here, however, show that global, nonlinear, ion-scale GYRO simulations are able to match the experimental ion heat flux, but underpredict electron heat flux (at most radii), electron temperature fluctuations, and perturbative thermal diffusivity in both L- and I-mode. Linear addition of electron heat flux from electron scale runs does not resolve this discrepancy. These results indicate that single-scale simulations do not sufficiently describe the I-mode core transport, and that multi-scale (coupled electron- and ion-scale) transport models are needed. A preliminary investigation with multi-scale TGLF, however, was unable to resolve the discrepancy between ion-scale GYRO and experimental electron heat fluxes and perturbative diffusivity, motivating further work with multi-scale GYRO simulations and a more comprehensive study with multi-scale TGLF. [ABSTRACT FROM AUTHOR]

Recent experiments on C-mod seeding nitrogen into ohmic plasmas with q95 = 3.4 found that the seeding greatly reduced long-wavelength (ITG-scale) turbulence. The long-wavelength turbulence that was reduced by the nitrogen seeding was localized to the region of r/α ≈ 0.85, where the turbulence is well above marginal stability (as evidenced by Qi/QGB ≫ 1). The nonlinear gyrokinetic code GYRO was used to simulate the expected turbulence in these plasmas, and the simulated turbulent density fluctuations and turbulent energy fluxes quantitatively agreed with the experimental measurements both before and after the nitrogen seeding. Unexpectedly, the intrinsic rotation of the plasma was also found to be affected by the nitrogen seeding, in a manner apparently unrelated to a change in the electron-ion collisionality that was proposed by other experiments. [ABSTRACT FROM AUTHOR]

The averaged toroidalflow of energetic minority ions during ICRF (ion cyclotron range of frequencies) heating is investigated in the Alcator C-Mod plasma by applying the GNET code, which can solve the drift kinetic equation with complicated orbits of accelerated energetic particles. It is found that a co-directional toroidalflow of the minority ions is generated in the region outside of the resonance location, and that the toroidal velocity reaches more than 40% of the central ion thermal velocity (Vtor ~ 300 km/s with PICRF ~ 2 MW). When we shift the resonance location to the outside of |r/a|~0.5|r/a|~0.5, the toroidalflow immediately inside of the resonance location is reduced to 0 or changes to the opposite direction, and the toroidal velocity shear is enhanced at r/a ~ 0.5. A radial diffusionequation for toroidalflow is solved by assuming a torque profile for the minority ion mean flow, and good agreements with experimental radial toroidalflow profiles are obtained. This suggests that the ICRF driven minority ion flow is related to the experimentally observed toroidal rotation during ICRF heating in the Alcator C-Mod plasma. [ABSTRACT FROM AUTHOR]

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 regime, the dilution primarily increased the critical gradient, while for plasmas in the SOC regime the dilution primarily decreased the stiffness. At r/a=0.8, where the experimental fluxes were above marginal stability, local GYRO predicted and experimental energy fluxes agreed, except for Qi in the SOC regime where GYRO under-predicted the experimental energy flux. At r/a=0.6, where the experimental fluxes were close to marginally stable, local GYRO predicted ITG modes to be strongly unstable and are responsible for both Qi and Qe (with Qi >Qe), as opposed to the experiment where Qi e. In contrast, global GYRO in this region predicted the ITG modes to be closer to marginal stability, and accurately predict the experimental Qi when the Ti profile is modified within experimental uncertainties. The fact that Qe is always less than Qi in the r/a=0.6 simulations with kδ?s≤1 indicates that high-k electron temperature gradient driven (ETG) modes must be included in future simulations and may be responsible for the electron energy transport in this case. [ABSTRACT FROM AUTHOR]

A database of toroidal momentum transport on five tokamaks, Alcator C-Mod, DIII-D, JET, NSTX and JT-60U, has been constructed under a wide range of conditions in order to understand the characteristics of toroidal momentum transport coefficients, namely the toroidal momentum diffusivity (χɸ) and the pinch velocity (Vpinch). Through an inter-machine comparison, the similarities and differences in the properties of χɸ and Vpinch among the machines have been clarified. Parametric dependences of these momentum transport coefficients have been investigated over a wide range of plasma parameters taking advantage of the different operation regimes in machines. The approach offers insights into the parametric dependences as follows. The toroidal momentum diffusivity (χɸ) generally increases with increasing heat diffusivity (χi). The correlation is observed over a wide range of χɸ, covering roughly two orders of magnitude, and within each of the machines over the whole radius. Through the inter-machine comparison, it is found that χɸ becomes larger in the outer region of the plasma. Also observed is a general trend for Vpinch in tokamaks; the inward pinch velocity (-Vpinch) increases with increasing χɸ. The results that are commonly observed in machines will support a toroidal rotation prediction in future devices. On the other hand, differences among machines have been observed. The toroidal momentum diffusivity, χɸ, is larger than or equal to χi in JET and JT-60U; on the other hand, χɸ is smaller than or equal to χi in NSTX, DIII-D and Alcator C-Mod. In DIII-D, the ratio -RVpinch/χɸ at r/a = 0.5-0.6 is about 2, which is small compared with that in other tokamaks (-RVpinch/χɸ ≈ 5). Based on these different observations, parametric dependences of χɸ/χi, RVpinch/χɸ and χɸ have been investigated in H-mode plasmas. Across the dataset from all machines, the ratio χɸ/χi tends to be larger in low . νe* at fixed Te/Ti and ρpol*. An increase in χɸ is observed with decreasing ne and/or increasing Te. The pinch number (-RVpinch/χɸ) is observed to increase with increasing R/Lne at both q95 = 5.5-7.2 and q95 = 3.7-4.5. Here νe*, ρpol*, Te, Ti, R/Lne and q95 are, respectively, the normalized effective electron collision frequency, the normalized ion poloidal Larmor radius, the electron and ion temperatures, the inverse ratio of density scale length, Lne, to the major radius, R, and the safety factor at the 95% flux surface. [ABSTRACT FROM AUTHOR]

This research describes advancements in the spectral analysis and error propagation techniques associated with x-ray imaging crystal spectroscopy (XICS) that have enabled this diagnostic to be used to accurately constrain particle, momentum, and heat transport studies in a tokamak for the first time. Doppler tomography techniques have been extended to include propagation of statistical uncertainty due to photon noise, the effect of non-uniform instrumental broadening as well as flux surface variations in impurity density. These methods have been deployed as a suite of modeling and analysis tools, written in interactive data language (IDL) and designed for general use on tokamaks. Its application to the Alcator C-Mod XICS is discussed, along with novel spectral and spatial calibration techniques. Example ion temperature and radial electric field profiles from recent I-mode plasmas are shown, and the impact of poloidally asymmetric impurity density and natural line broadening is discussed in the context of the planned ITER x-ray crystal spectrometer. [ABSTRACT FROM AUTHOR]

Ion cyclotron range of frequency (ICRF) mode conversion has been shown to drive toroidal flow in JET D(3He) L-mode plasmas: Bt0 = 3.45T, ne0 ∼ 3×1019 m-3, Ip = 2.8 and 1.8 MA, PRF ≤ 3MW at 33MHz and -90° phasing. Central toroidal rotation in the counter-Ip direction, with ωϕ0 up to 10 krad s-1 (Vϕ0 ∼ 30 km s-1, central thermal Mach number Mth(0) ∼ 0.07 and Alfvén Mach number MA(0) ∼ 0.003) has been observed. The flow drive effect is sensitive to the 3He concentration and the largest rotation is observed in the range X[3He] = nHe3/ne ∼ 10-17%. The rotation profile is peaked near the magnetic axis, and the central rotation scales with the input RF power. The effective torque density profile from the RF power has been calculated and the total torque is estimated to be as high as 50% of the same power from neutral beam injection, and a factor of 5 larger than the direct momentum injection from the RF waves. RF physics modeling using the TORIC code shows that the interaction between the mode converted ion cyclotron wave and the 3He ions, and associated asymmetry in space and momentum, may be key for flow drive. [ABSTRACT FROM AUTHOR]

A numerical modeling of current profile modification by lower hybrid current drive (LHCD) using a fullwave/Fokker-Planck simulation code is presented. A MHD stable LHCD discharge on Alcator C-Mod was analyzed, and the current profile from full wave simulations was found to show better agreement with the experiment than a ray-tracing code. Comparison of full wave and ray-tracing simulation shows that, although ray-tracing can reproduce the stochastic wave spectrum broadening, the full wave calculation predicts even wider spectrum broadening, and the wave spectrum fills all of the kinematically allowed domain. This is the first demonstration of LHCD current profile modeling using a full wave simulation code in a multi-pass absorption regime, showing the clear impact of full wave effects on the LHCD driven current profile. [ABSTRACT FROM AUTHOR]

High-resolution charge-exchange recombination spectroscopic measurements of B5+ ions have enabled the first spatially resolved calculations of the radial electric field (Er) in the Alcator C-Mod pedestal region [E. S. Marmar, Fusion Sci. Technol. 51, 261 (2006)]. These observations offer new challenges for theory and simulation and provide for important comparisons with other devices. Qualitatively, the field structure observed on C-Mod is similar to that on other tokamaks. However, the narrow high-confinement mode (H-mode) Er well widths (5 mm) observed on C-Mod suggest a scaling with machine size, while the observed depths (up to 300 kV/m) are unprecedented. Due to the strong ion-electron thermal coupling in the C-Mod pedestal, it is possible to infer information about the main ion population in this region. The results indicate that in H-mode the main ion pressure gradient is the dominant contributor to the Er well and that the main ions have significant edge flow. C-Mod H-mode data show a clear correlation between deeper Er wells, higher confinement plasmas, and higher electron temperature pedestal heights. However, improved L-mode (I-mode) plasmas exhibit energy confinement equivalent to that observed in similar H-mode discharges, but with significantly shallower Er wells. I-mode plasmas are characterized by H-mode-like energy barriers, but with L-mode-like particle barriers. The decoupling of energy and particle barrier formation makes the I-mode an interesting regime for fusion research and provides for a low collisionality pedestal without edge localized modes. [ABSTRACT FROM AUTHOR]

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In the absence of any auxiliary torque input, the DIII-D plasma consists of nonzero toroidal angular momentum, in other words, it rotates. This effect is commonly observed in tokamaks, being referred to as intrinsic rotation. Measurements of intrinsic rotation profiles have been made in DIII-D [J. Luxon, Nucl. Fusion 42, 614 (2002)] H-mode discharges, with both Ohmic heating (OH) and electron cyclotron heating (ECH) in which there is no auxiliary torque. Recently, the H-mode data set has been extended with the newly configured DIII-D simultaneous co- and counter-directed neutral beam injection (NBI) capability resulting in control of the local torque deposition, where co and counter refer to the direction relative to the toroidal plasma current. Understanding intrinsic rotation is important for projection toward burning plasma performance where any NBI torque will be relatively small. The toroidal velocity is recognizably important regarding issues of stability and confinement. In DIII-D ECH H-modes the rotation profile is hollow, co-directed at large minor radius and depressed, or actually counter-directed, nearer the magnetic axis. This profile varies with the ECH power deposition profile to some extent. In contrast, OH H-modes have a relatively flat co-directed rotation profile. There is a scaling of the DIII-D intrinsic toroidal velocity with W/Ip, as seen in intrinsic rotation in Alcator C-Mod [J. Rice, Nucl. Fusion 39, 1175 (1999)], where W is the total plasma thermal energy and Ip is the magnitude of the toroidal plasma current. This common scaling resulted in a dimensionless similarity experiment between DIII-D and Alcator C-Mod on intrinsic rotation, obtaining a single spatial point match in the toroidal velocity normalized to the ion thermal velocity. The balanced NBI capability in DIII-D is a useful tool to push scaling studies to higher values of the plasma normalized energy, notwithstanding the details of torque deposition for co-NBI versus counter-NBI. There are theories which address intrinsic rotation, both extensions of neoclassical theory and related to turbulent transport. At this time, the comparisons with theory are qualitative. [ABSTRACT FROM AUTHOR]

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 (2.6–7.86 T) and plasma current (0.4–1.7 MA). 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 reduction in thermal transport prior to the transition in particle confinement. Pedestal temperatures in this case are also higher than in the favorable configuration. Both high-field and reversed-field results suggest that parameters at the L-H transition are influencing the evolution and parameters of the H-mode pedestal. [ABSTRACT FROM AUTHOR]

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 (EN) mode, is investigated by new high frequency (132 and 140 GHz) reflectometer channels operating in the ordinary (O) mode. This additional hardware has been installed in an effort to study localized turbulence associated with internal transport barriers (ITBs). The EN mode is a suitable candidate for this study, since an ITB exists transiently as the outer density decreases much faster than the core density in this mode. The fact that the density decays from the outside inward allows us to study fluctuations progressing towards the plasma core. Our results mark the first localized observation of the QC mode at medium density: 2.2×1020 m-3 (132 GHz). Correlating the reflectometry measurements with other fluctuating quantities provides some insight regarding the causality of the EN-mode development. [ABSTRACT FROM AUTHOR]

A set of external coils (A-coils) capable of producing nonaxisymmetric, predominantly n=1, fields with different toroidal phase and a range of poloidal mode m spectra has been used to determine the threshold amplitude for mode locking over a range of plasma parameters in Alcator C-Mod [I. H. Hutchinson, R. Boivin, F. Bombarda, P. Bonoli, S. Fairfax, C. Fiore, J. Goetz, S. Golovato, R. Granetz, M. Greenwald et al., Phys. Plasmas 1, 1511 (1994)]. The threshold perturbations and parametric scalings, expressed in terms of (B21/BT), are similar to those observed on larger, lower field devices. The threshold is roughly linear in density, with typical magnitudes of order 10-4. This result implies that locked modes should not be significantly more problematic for the International Thermonuclear Experimental Reactor [I. P. B. Editors, Nucl. Fusion 39, 2286 (1999)] than for existing devices. Coordinated nondimensional identity experiments on the Joint European Torus [Fusion Technol. 11, 13 (1987)], DIII-D [Fusion Technol. 8, 441 (1985)], and C-Mod, with matching applied mode spectra, have been carried out to determine more definitively the field and size scalings. Locked modes on C-Mod are observed to result in braking of core toroidal rotation, modification of sawtooth activity, and significant reduction in energy and particle confinement, frequently leading to disruptions. Intrinsic error fields inferred from the threshold studies are found to be consistent in amplitude and phase with a comprehensive model of the sources of field errors based on “as-built” coil and bus-work details and coil imperfections inferred from measurements using in situ magnetic diagnostics on dedicated test pulses. Use of the A-coils to largely cancel the 2/1 component of the intrinsic nonaxisymmetric field has led to expansion of the accessible operating space in C-Mod, including operation up to 2 MA plasma current at 8 T. [ABSTRACT FROM AUTHOR]

The time evolution of toroidal rotation velocity profiles has been measured in Alcator C-Mod [Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] plasmas using a tangentially viewing x-ray spectrometer array. The strong co-current toroidal rotation in enhanced Dα(EDA) high confinement mode (H-mode) plasmas is observed to propagate in from the edge on a time scale similar to the energy confinement time. The ensuing steady state rotation velocity profiles in both Ohmic and ion cyclotron range of frequencies (ICRF) heated EDA H modes, which are generated in the absence of any external momentum input, are found to be relatively flat. These profiles may be simulated by a simple diffusion model with the boundary condition of an edge rotation, which appears during the H-mode period. The observed profiles are well matched by the simulations using a momentum diffusivity of ∼ 0.1 m²/s, which is much larger than the calculated neo-classical value, and the momentum transport may he regarded as anomalous. The Alcator C-Mod rotation observations have been compared in detail with the calculations of neo-classical and sub-neo-classical theory, to the predictions from modeling of ICRF wave induced energetic ion orbit shifts, and to estimates from turbulence driven mechanisms. The magnitude and scalings of the observed rotation results are in accord with neo-classical and suh-neo-classical calculations, hut the measured momentum diffusivity is higher than the predictions by a large factor. The prediction of rotation reversal with a high magnetic field side resonance location for ICRF wave induced ion orbit shifts has not been observed in the experiments. While the turbulence driven rotation calculations are mostly qualitative, they represent some of the observed features. [ABSTRACT FROM AUTHOR]

Mode conversion (MC) of long wavelength fast electromagnetic magnetosonic waves (fast wave, or FW) into shorter wavelength electrostatic (ion-Bernstein, or IBW) or slow electromagnetic (ion cyclotron, or ICW) waves is of great interest in laboratory, magnetic fusion and space physics experiments. Such processes are particularly important in multi-ion species plasmas. In this paper we report recent results from high power ion cyclotron range of frequencies (ICRF) heating experiments in the Alcator C-Mod tokamak. Mode converted waves near the ³He-H hybrid layer have been detected by means of phase contrast imaging in H(³He,D) plasmas [E. Nelson-Melby et al., Phys. Rev. Lett. 90, 155004 (2003)]. The measured wave k spectrum and spatial location are in agreement with theoretical predictions [F. W. Perkins, Nucl. Fusion 17, 1197 (1977)], which showed that in a sheared magnetic field, mode-conversion of FW into ICW may dominate over IBW for appropriate ion species (i.e., D-T, or equivalently, H- ³He). Recent modeling with full wave codes, as well as solving the hot plasma dispersion equation in the presence of sheared magnetic fields, verifies the interpretation of such a mode conversion process. Thus, the geometry of the magnetic field, as well as the particular ion species mix, influences the physics of ICRF mode conversion. In this paper, we also report recent results on the study of mode conversion electron heating (MCEH) in D(H) plasmas [Y. Lin et al., Plasmas Phys. Controlled Fusion 45, 1013 (2003)]. By comparing the experimentally measured MCEH profile with modeling, the study shows that the MC ICW may make a significant contribution to the direct electron heating when the D-H hybrid layer is off axis on the high field side. Preliminary results of mode conversion poloidal plasma flow drive experiments in D(³He) are also reported. [ABSTRACT FROM AUTHOR]

Recent studies of internal transport and double transport barrier regimes in the Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] have explored the limits for forming, maintaining, and controlling these plasmas. The C-Mod provides a unique platform for studying such discharges: the ions and electrons are tightly coupled by collisions and the plasma has no internal particle or momentum sources. The double-barrier mode comprised of an edge barrier with an internal transport barrier (ITB) can be induced at will using off-axis ion cyclotron range of frequency (ICRF) injection on either the low or high field side of the plasma with either of the available ICRF frequencies (70 or 80 MHz). When an enhanced Dα high confinement mode (EDA H-mode) is accessed in Ohmic plasmas, the double barrier ITB forms spontaneously if the H-mode is sustained for ∼2 energy confinement times. The ITBs formed in both Ohmic and ICRF heated plasmas are quite similar regardless of the trigger method. They are characterized by strong central peaking of the electron density, and a reduction of the core particle and energy transport. The control of impurity influx and heating of the core plasma in the presence of the ITB have been achieved with the addition of central ICRF power in both the Ohmic H-mode and ICRF induced ITBs. The radial location of the particle transport barrier is dependent on the toroidal magnetic field but not on the location of the ICRF resonance. A narrow region of decreased electron thermal transport, as determined by sawtooth heat pulse analysis, is found in these plasmas as well. Transport analysis indicates that a reduction of the particle diffusivity in the barrier region allows the neoclassical pinch to drive the density and impurity accumulation in the plasma center. [ABSTRACT FROM AUTHOR]

Nonlinear gyrokinetic simulations of trapped electron mode (TEM) turbulence, within an internal particle transport barrier, are performed and compared with experimental data. The results provide a mechanism for transport barrier control with on-axis radio frequency heating, as demonstrated in Alcator C-Mod experiments [S. J. Wukitch et al., Phys. Plasmas 9, 2149 (2002)]. Off-axis heating produces an internal particle and energy transport barrier after the transition to enhanced D ahigh confinement mode. The barrier foot reaches the half-radius, with a peak density 2.5 times the edge density. While the density profile peaks, the temperature profile remains relatively unaffected. The peaking and concomitant impurity accumulation are controlled by applying modest central heating power late in the discharge. Gyrokinetic turbulence simulations of the barrier formation phase, using the GS2 code [W. Dorland et al., Phys. Rev. Lett. 85, 5579 (2000)] show that toroidal ion temperature gradient driven modes are suppressed inside the barrier foot, but continue to dominate in the outer half-radius. As the density gradient steepens further, trapped electron modes are driven unstable. The onset of TEM turbulence produces an outflow that strongly increases with the density gradient, upon exceeding a new nonlinear critical density gradient, which significantly exceeds the linear critical density gradient. The TEM turbulent outflow ultimately balances the inward Ware pinch, leading to steady state. Moreover, the simulated turbulent particle diffusivity matches that inferred from particle balance using measured density profile data and the calculated Ware pinch. This turbulent diffusivity exhibits a strong unfavorable temperature dependence that allows control with central heating. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

The radial profiles of titanium K[sub α] spectra are measured with photon counting x-ray charge coupled device cameras for plasmas with a neoclassical internal transport barrier (N-ITB) in the Compact Helical System Heliotron/Torsatron [K. Matsuoka et al., Plasma Phys. Controlled Fusion 42, 1145 (2000)]. The K[sub α] lines are dominated by the resonance line of He-like titanium in the plasma core where a thermal transport barrier results in a high electron temperature of ∼3 keV. The impurity diffusion coefficients are evaluated from the ionization levels of titanium. The diffusion coefficient inside the neoclassical ITB is <0.03 m²/s, which is one order of magnitude lower than that of the plasma without neoclassical ITB. [ABSTRACT FROM AUTHOR]

The formation of internal transport barriers (ITBs) has been observed in the core region of Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] under a variety of conditions. The improvement in core confinement following pellet injection (pellet enhanced performance or PEP mode) has been well documented on Alcator C-Mod in the past. Recently three new ITB phenomena have been observed which require no externally applied particle or momentum input. Short lived ITBs form spontaneously following the high confinement to low confinement mode transition and are characterized by a large increase in the global neutron production (enhanced neutron or EN modes). Experiments with ion cyclotron range of frequencies power injection to the plasma off-axis on the high field side results in the central density rising abruptly and becoming peaked. The ITB formed at this time lasts for ten energy confinement times. The central toroidal rotation velocity decreases and changes sign as the density rises. Similar spontaneous ITBs have been observed in ohmically heated H-mode plasmas. All of these ITB events have strongly peaked density profiles with a minimum in the density scale length occurring near r/a = 0.5 and have improved confinement parameters in the core region of the plasma. [ABSTRACT FROM AUTHOR]

Investigates the effect of divertor baffling on Alcator C-Mod discharges and main plasma using a novel divertor bypass. Compression of deuterium and recycling impurity gases in the divertor plenum; Control of plasma density; Impact of increased helium pressure on rate of helium ash removal from the system.

High-speed lithium pellets have been injected into Alcator C tokamak plasmas in order to measure the internal magnetic field, and thus current density profiles. In the pellet ablation cloud, intense visible line radiation from the Li[sup +] ion (λ ≈ 5485 Å, ls2s ³S -- ls2p ³P) is polarized due to the Zeeman effect, and measurement of the polarization angle yields the direction of the total local magnetic field. A "snap shot" of the q profile is obtained as the pellet penetrates from the edge into the center of the discharge, in a time of about 300 µs. The spatial resolution of the measurement is about 1 cm. At a toroidal field of B[sub T] = 10 T, the emission in the unshifted π component of the Zeeman triplet is more than 80% polarized, and q profiles have been obtained. The pellets are perturbative (〉Δn[sub e]〈/〉n[sub e]〈 ≈ 1), but the total pellet penetration time is at least a factor of 1000 smaller than the classical skin time. It can thus be anticipated that the current density profile should not be perturbed significantly during the time of the measurement. With some relatively straightforward modifications and refinements, precision approaching 10% for the measurement of q profiles should be achievable. The technique appears viable, using Li, as long as the toroidal field is ...4 T. [ABSTRACT FROM AUTHOR]

The TFTR horizontal x-ray crystal spectrometer has been improved by installation of γ and neutron shielding and is now operating in the TFTR Test Cell up to the presently maximum neutron production rates of 2×1016 neutrons/s. Further optimization by using a modified detector may assure operation of the spectrometer during the DT phase when TFTR is covered by an igloo. The instrument is set up for simultaneous observation of the entire satellite spectrum of Ni xxvii, and it provides data on the time evolution of the central ion temperature and the central plasma rotation velocity. Ion temperatures of 20–30 keV have been measured under ‘‘supershot’’ conditions. The highest rotation velocity observed with unidirectional injection was 1.1×106 m/s. [ABSTRACT FROM AUTHOR]

We present a high throughput (f/3) visible (3500–7000 Å) Doppler spectrometer for toroidal rotation velocity measurements of the Alcator C-Mod tokamak plasma. The spectrometer has a temporal response of 1 ms and a rotation velocity sensitivity of ∼105 cm/s. This diagnostic will have a tangential view and map out the plasma rotation at several locations along the outer half of the minor radius (r/a>0.5). The plasma rotation will be determined from the Doppler shifted wavelengths of Dα and magnetic and electric dipole transitions of highly ionized impurities in the plasma. The fast time resolution and high spectral resolving power are possible due to a 6 in. diam circular transmission grating that is capable of λ/Δλ∼15 500 at 5769 Å in conjunction with a 50 μm slit. [ABSTRACT FROM AUTHOR]

New observations of the formation and dynamics of long-lived impurity-induced helical "snake" modes in tokamak plasmas have recently been carried out on Alcator C-Mod. The snakes form as an asymmetry in the impurity ion density that undergoes a seamless transition from a small helically displaced density to a large crescent-shaped helical structure inside q<1, with a regularly sawtoothing core. The observations show that the conditions for the formation and persistence of a snake cannot be explained by plasma pressure alone. Instead, many features arise naturally from nonlinear interactions in a 3D MHD model that separately evolves the plasma density and temperature. [ABSTRACT FROM AUTHOR]