Matthew Reinke

Critical gradients and plasma flows in the edge plasma of Alcator C-Mod

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...

Scaling of the power exhaust channel in Alcator C-Mod

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 ...

Gas jet disruption mitigation studies on Alcator C-Mod and DIII-D

High-pressure noble gas jet injection is a mitigation technique which potentially satisfies the requirements of fast response time and reliability, without degrading subsequent discharges. Previously reported gas jet experiments on DIII-D showed good success at reducing deleterious disruption effects. In this paper, results of recent gas jet disruption mitigation experiments on Alcator C-Mod and DIII-D are reported. Jointly, these experiments have greatly improved the understanding of gas jet dynamics and the processes involved in mitigating disruption effects. In both machines, the sequence of events following gas injection is observed to be quite similar: the jet neutrals stop near the plasma edge, the edge temperature collapses and large MHD modes are quickly destabilized, mixing the hot plasma core with the edge impurity ions and radiating away the plasma thermal energy. High radiated power fractions are achieved, thus reducing the conducted heat loads to the chamber walls and divertor. A significant (2 × or more) reduction in halo current is also observed. Runaway electron generation is small or absent. These similar results in two quite different tokamaks are encouraging for the applicability of this disruption mitigation technique to ITER.

X-ray imaging crystal spectroscopy for use in plasma transport research

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.

Edge energy transport barrier and turbulence in the I-mode regime on Alcator C-Mod

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 ...

Gas jet disruption mitigation studies on Alcator C-Mod

Damaging effects of disruptions are a major concern for Alcator C-Mod, ITER and future tokamak reactors. High-pressure noble gas jet injection is a mitigation technique which potentially satisfies the operational requirements of fast response time and reliability, while still being benign to subsequent discharges. Disruption mitigation experiments using an optimized gas jet injection system are being carried out on Alcator C-Mod to study the physics of gas jet penetration into high pressure plasmas, as well as the ability of the gas jet impurities to convert plasma energy into radiation on timescales consistent with C-Mods fast quench times, and to reduce halo currents given C-Mods high-current density. The dependence of impurity penetration and effectiveness on noble gas species (He, Ne, Ar, Kr) is also being studied. It is found that the high-pressure neutral gas jet does not penetrate deeply into the C-Mod plasma, and yet prompt core thermal quenches are observed on all gas jet shots. 3D MHD modelling of the disruption physics with NIMROD shows that edge cooling of the plasma triggers fast growing tearing modes which rapidly produce a stochastic region in the core of the plasma and loss of thermal energy. This may explain the apparent effectiveness of the gas jet in C-Mod despite its limited penetration. The higher-Z gases (Ne, Ar, Kr) also proved effective at reducing halo currents and decreasing thermal deposition to the divertor surfaces. In addition, noble gas jet injection proved to be benign for plasma operation with C-Mods metal (Mo) wall, actually improving the reliability of the startup in the following discharge.

Vacuum ultraviolet impurity spectroscopy on the Alcator C-Mod tokamak.

Vacuum ultraviolet spectroscopy is used on the Alcator C-Mod tokamak to study the physics of impurity transport and provide feedback on impurity levels to assist experimental operations. Sputtering from C-Mods all metal (Mo+W) plasma facing components and ion cyclotron range of frequency antenna and vessel structures (sources for Ti, Fe, Cu, and Ni), the use of boronization for plasma surface conditioning and Ar, Ne, or N(2) gas seeding combine to provide a wealth of spectroscopic data from low-Z to high-Z. Recently, a laser blow-off impurity injector has been added, employing CaF(2) to study core and edge impurity transport. One of the primary tools used to monitor the impurities is a 2.2 m Rowland circle spectrometer utilizing a Reticon array fiber coupled to a microchannel plate. With a 600 lines/mm grating the 80<λ<1050 Å range can be scanned, although only 40-100 Å can be observed for a single discharge. Recently, a flat-field grating spectrometer was installed which utilizes a varied line spacing grating to image the spectrum to a soft x-ray sensitive Princeton Instruments charge-coupled device camera. Using a 2400 lines/mm grating, the 10<λ<70 Å range can be scanned with 5-6 nm observed for a single discharge. A variety of results from recent experiments are shown that highlight the capability to track a wide range of impurities.

20 years of research on the Alcator C-Mod tokamak

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-modes 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.

Overview of the Alcator C-MOD Research Program

Alcator C-MOD has compared plasma performance with plasma-facing components (PFCs) coated with boron to all-metal PFCs to assess projections of energy confinement from current experiments to next-generation burning tokamak plasmas. Low-Z coatings reduce metallic impurity influx and diminish radiative losses leading to higher H-mode pedestal pressure that improves global energy confinement through profile stiffness. RF sheath rectification along flux tubes that intersect the RF antenna is found to be a major cause of localized boron erosion and impurity generation. Initial lower hybrid current drive (LHCD) experiments (PLH < 900?kW) in preparation for future advanced-tokamak studies have demonstrated fully non-inductive current drive at Ip ~ 1.0?MA with good efficiency, Idrive = 0.4 PLH/neoR (MA, MW, 1020?m?3,m). The potential to mitigate disruptions in ITER through massive gas-jet impurity puffing has been extended to significantly higher plasma pressures and shorter disruption times. The fraction of total plasma energy radiated increases with the Z of the impurity gas, reaching 90% for krypton. A positive major-radius scaling of the error field threshold for locked modes (Bth/B ? R0.68?0.19) is inferred from its measured variation with BT that implies a favourable threshold value for ITER. A phase contrast imaging diagnostic has been used to study the structure of Alfv?n cascades and turbulent density fluctuations in plasmas with an internal transport barrier. Understanding the mechanisms responsible for regulating the H-mode pedestal height is also crucial for projecting performance in ITER. Modelling of H-mode edge fuelling indicates high self-screening to neutrals in the pedestal and scrape-off layer (SOL), and reproduces experimental density pedestal response to changes in neutral source, including a weak variation of pedestal height and constant width. Pressure gradients in the near SOL of Ohmic L-mode plasmas are observed to scale consistently as , and show a significant dependence on X-point topology. Fast camera images of intermittent turbulent structures at the plasma edge show they travel coherently through the SOL with a broad radial velocity distribution having a peak at about 1% of the ion sound speed, in qualitative agreement with theoretical models. Fast D? diagnostics during gas puff imaging show a complex behaviour of discrete ELMs, starting with an n ? 10 precursor oscillation followed by a rapid primary ejection as the pedestal crashes and then multiple, slower secondary ejections.

Pedestal structure and stability in H-mode and I-mode: a comparative study on Alcator C-Mod

New experimental data from the Alcator C-Mod tokamak are used to benchmark predictive modelling of the edge pedestal in various high-confinement regimes, contributing to greater confidence in projection of pedestal height and width in ITER and reactors. ELMy H-modes operate near stability limits for ideal peeling–ballooning modes, as shown by calculations with the ELITE code. Experimental pedestal width in ELMy H-mode scales as the square root of βpol at the pedestal top, i.e. the dependence expected from theory if kinetic ballooning modes (KBMs) were responsible for limiting the pedestal width. A search for KBMs in experiment has revealed a short-wavelength electromagnetic fluctuation in the pedestal that is a candidate driver for inter-edge localized mode (ELM) pedestal regulation. A predictive pedestal model (EPED) has been tested on an extended set of ELMy H-modes from C-Mod, reproducing pedestal height and width reasonably well across the data set, and extending the tested range of EPED to the highest absolute pressures available on any existing tokamak and to within a factor of three of the pedestal pressure targeted for ITER. In addition, C-Mod offers access to two regimes, enhanced D-alpha (EDA) H-mode and I-mode, that have high pedestals, but in which large ELM activity is naturally suppressed and, instead, particle and impurity transport are regulated continuously. Pedestals of EDA H-mode and I-mode discharges are found to be ideal magnetohydrodynamic (MHD) stable with ELITE, consistent with the general absence of ELM activity. Invocation of alternative physics mechanisms may be required to make EPED-like predictions of pedestals in these kinds of intrinsically ELM-suppressed regimes, which would be very beneficial to operation in burning plasma devices.