Stationary density profiles in the Alcator C-mod tokamak

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

High field side launch of RF waves: A new approach to reactor actuators

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

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-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

Alcator C-Mod: research in support of ITER and steps beyond

This paper presents an overview of recent highlights from research on Alcator C-Mod. Significant progress has been made across all research areas over the last two years, with particular emphasis on divertor physics and power handling, plasma–material interaction studies, edge localized mode-suppressed pedestal dynamics, core transport and turbulence, and RF heating and current drive utilizing ion cyclotron and lower hybrid tools. Specific results of particular relevance to ITER include: inner wall SOL transport studies that have led, together with results from other experiments, to the change of the detailed shape of the inner wall in ITER; runaway electron studies showing that the critical electric field required for runaway generation is much higher than predicted from collisional theory; core tungsten impurity transport studies reveal that tungsten accumulation is naturally avoided in typical C-Mod conditions.United States. Department of Energy (DE-FC02-99ER54512-CMOD)United States. Department of Energy (DE-AC02-09CH11466)United States. Department of Energy (DE-FG02-96ER-54373)United States. Department of Energy (DE-FG02-94ER54235

LHCD on Alcator C-Mod : measurements with an upgraded MSE diagnostic and comparisons to simulation

Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2015.This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.Cataloged from student-submitted PDF version of thesis.Includes bibliographical references.Lower Hybrid Current Drive (LHCD) is a promising technique to sustain tokamak plasmas and provide control over the current profile--two important capabilities required for the development of tokamak fusion reactors. Upgraded measurement capabilities on the Alcator C-Mod Tokamak create a unique opportunity to study the plasma's toroidal electric current profile at magnetic fields, plasma densities, and magnetic geometries anticipated in future reactors in stationary discharges dominated by lhcd. The Motional Stark Effect (MSE) diagnostic uses polarized light to infer the plasma's internal current profile. The MSE diagnostic deployed on the Alcator C-Mod Tokamak previously experienced unacceptable calibration drift and sensitivity to partially-polarized background light that limited its ability to measure magnetic field pitch-angles. A comprehensive analytic study of the origin of polarization angle errors in MSE diagnostics and an experimental study using a robotic calibration system were conducted. Insight from this study guided the fabrication and installation of a first-of-a-kind in-situ calibration system for MSE diagnostics--a long sought capability-- and the development of thermal isolation schemes for the periscope. An experimental study of the effect of partially polarized background light identified this as a significant source of systematic error. Partial-polarization upon reflection was identified as the mechanism that leads to polarized light in a tokamak. Visible bremsstrahlung, divertor emission, and blackbody emission were identified as the dominant sources of light. A new technique, MSE multi-spectral line polarization (MSE-MSLP), was developed to measure the polarization on a single sight line in multiple wavelengths simultaneously using a high-throughput polarization polychromator. Wavelength-interpolation of the background light polarization utilizing this hardware decreases the error from background subtraction by a factor of 5-10 relative to time-interpolation, drastically improving the measurement quality while eliminating the need for neutral beam pulsing. The method also allows for simultaneous measurement of multiple polarized transitions within the Stark multiplet. The upgraded MSE diagnostic was used to measure the magnetic field pitch angle profile in plasmas with some or all of the plasma current driven by lhcd. Measurements were made across a range of single-parameter scans: lhcd power, loop voltage, plasma density, plasma current, and launched n// spectrum. The current profile is observed to broaden during lhcd, but consistently has significant on-axis current density, even in fully non-inductive plasmas. The current profile and hard x-ray (HXR) profiles are observed to be most sensitive to plasma current, with higher current yielding broader profiles. The current and HXR profiles as well as global current-drive efficiency are insensitive to changes in n// or loop voltage. Numerical simulations by the ray-tracing Fokker-Planck GENRAY/CQL3D code reproduce the total measured current in non-inductive conditions but fail to accurately predict the current and HXR profiles; the simulations consistently predict more current drive in the outer half of the plasma than is observed. This leads to a flattening of the HXR profile compared to the experimental profiles. These qualitative discrepancies persist across the range of plasma parameters scanned. Varying code inputs within their measurement uncertainties and adding experimentally-constrained levels of fast-electron diffusion do not reconcile profile discrepancies. Some qualitative profile trends in single parameter scans are reproduced by the simulations including broadening of profiles at higher current, and a weak dependence on the launched n//spectrum. However, HXR profile self-similarity across different densities and powers is not reproduced. These new comparisons between profile measurements and simulation suggest that the simulations are missing important physics in this operational regime.by Robert Thomas Mumgaard.Ph. D

Engineering upgrades to the MSE diagnostic on Alcator C-Mod

Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2015.This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.Cataloged from student-submitted PDF version of thesis.Includes bibliographical references (pages 135-136).The Motional Stark Effect (mse) diagnostic deployed on the alcator c-mod tokamak previously experienced unacceptable calibration drift and sensitivity to partially-polarized background light that limited its ability to measure magnetic field pitch-angles. The need to identify the cause of the polarization drift and to characterize sources of error motivated the development of a robotic calibration system consisting of a motorized three axis positioning system and a light source capable of generating arbitrary polarization states. The system produces linear polarization angles with accuracy of 15mm2 sr for four simultaneous polarization measurements on the same viewing sightline. It sequentially passes light among four different < 0.9nm FWHM temperature-tuned filters to custom avalanche photodiode detectors operating at NA = 0.6. The polychromator is imaging, non-vignetting, high transmission, utilizes mostly off-the shelf optics, is easily aligned, remote-controlled and can be replicated at relatively low cost. This enables wavelength-interpolation of the mse background light and simultaneous detection of different parts of the Stark manifold.by Robert Thomas Mumgaard.S.M

A small, novel, remote in-vessel inspection system for the Alcator C-Mod tokamak

A small robust system has been constructed for in-situ visual inspection of the Alcator C-Mod tokamak. The system consists of a small, light, wide-angle high definition camera and LED package housed in a nacelle on the end of thin, rigid, 3.5 m long support pole. The nacelle has two actuated degrees of freedom allowing the camera to observe nearly 4π steradians. The support pole has a specific slight curve that allows it to pass to either side of the center column of the tokamak to observe the entirety of the vessel interior, while still fitting through the small aspect ratio Alcator C-Mod vacuum port structure. The support pole and camera can enter the vessel through any horizontal vacuum port with an inner diameter greater than 4 cm, thus a dedicated port is not required. The inspection is typically undertaken during maintenance periods when the vessel is filled with a noble gas near atmospheric pressure thus minimizing the influx of water vapor and the concomitant loss of wall conditioning. The system is operated manually, producing photos and video which are reviewed in near real-time. Nearly the entire vessel, including the plasma facing components, can be carefully inspected in 3–5 h. The system provides improved characterization of the interior components and surfaces of the tokamak with a modest engineering and operational effort. Information gathered from the system has identified damage to plasma facing components that were interfering with tokamak operation, as well as damage to mechanical components which were redesigned during the remainder of the campaign, thereby enhancing program planning.United States. Department of Energy (Award DE-FC02-99ER54512

LHCD during current ramp experiments on Alcator C-Mod

The lower hybrid current drive (LHCD) system on Alcator C-Mod is capable of sustaining fully non-inductive discharges for multiple current relaxation times (τ cr ∼ 200 ms) at line averaged densities in the range of 5x10 19 m -3 . Some of these non-inductive discharges develop unstable MHD modes that can greatly reduce current drive performance, particularly in discharges with plasma current of 0.5 MA or less [1,2]. Avoiding these unstable MHD modes motivated an experiment to test if the stable current profile shape of a higher current non-inductive discharge could be achieved in a lower current discharge. Starting from a discharge at 0.8 MA, the plasma current was ramped down to 0.5 MA over 200 ms. The surface voltage of the plasma swings negative during the ramp, with the loop voltage reversal impacting the edge fast electron measurements immediately. Little change can be seen during the I p ramp in the core fast electron measurements, indicating that the loop voltage reversal does not penetrate fully to the magnetic axis on the timescale of the current ramp. The resulting discharge did not exhibit deleterious MHD instabilities, however the existence of this one discharge does not necessarily represent a robust solution to the problem.United States. Department of Energy. Office of Fusion Energy Sciences (Award DE-FC02-99ER54512)United States. Department of Energy. Office of Fusion Energy Sciences (Award DE-AC02-76CH03073

Core impurity transport in Alcator C-Mod L-, I- and H-mode plasmas

Core impurity transport has been investigated for a variety of confinement regimes in Alcator C-Mod plasmas from x-ray emission following injection of medium and high Z materials. In ohmic L-mode discharges, impurity transport is anomalous (D[subscript eff] ≫ D[subscript nc]) and changes very little across the LOC/SOC boundary. In ion cyclotron range of frequencies (ICRF) heated L-mode plasmas, the core impurity confinement time decreases with increasing ICRF input power (and subsequent increasing electron temperature) and increases with plasma current. Nearly identical impurity confinement characteristics are observed in I-mode plasmas. In enhanced D[subscript α] H-mode discharges the core impurity confinement times are much longer. There is a strong connection between core impurity confinement time and the edge density gradient across all confinement regimes studied here. Deduced central impurity density profiles in stationary plasmas are generally flat, in spite of large amplitude sawtooth oscillations, and there is little evidence of impurity convection inside of r/a = 0.3 when averaged over sawteeth.United States. Department of Energy (Contract DE-FC02-99ER54512)United States. Dept. of Energy. Fusion Energy Postdoctoral Research Program (Oak Ridge Institute for Science and Education