S. Wukitch

Princeton University, Princeton, New Jersey, United States

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Publications (84)81.93 Total impact

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    ABSTRACT: Radio frequency (RF) power in the ion cyclotron range of frequencies (ICRF) is one of the primary auxiliary heating techniques for Experimental Advanced Superconducting Tokamak (EAST). The ICRF system for EAST has been developed to support long-pulse high-β advanced tokamak fusion physics experiments. The ICRF system is capable of delivering 12 MW 1000-s RF power to the plasma through two antennas. The phasing between current straps of the antennas can be adjusted to optimize the RF power spectrum. The main technical features of the ICRF system are described. Each of the 8 ICRF transmitters has been successfully tested to 1.5 MW for a wide range of frequency (25–70 MHz) on a dummy load. Part of the ICRF system was in operation during the EAST 2012 spring experimental campaign and a maximum power of 800 kW (at 27 MHz) lasting for 30 s has been coupled for long pulse H mode operation.
    Fusion Engineering and Design. 01/2014;
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    ABSTRACT: We describe results from imaging observations of atomic line and continuum emission in the 550.6 nm region on Alcator C-Mod. Both the 550.6 nm neutral molybdenum emission and the adjacent 549 nm continuum emission are imaged separately to isolate line emission. A few complications of using imaging to infer erosion in this wavelength region are discussed including subtraction of continuum emission and determination of an appropriate S/XB coefficient. Diagnostics of surface erosion and thermography using these emissions are briefly reviewed, and used to study phenomenology during ohmic operation, ion cyclotron range of frequencies heating (ICRH), and lower hybrid current drive (LHCD). In addition to broadening of Mo I emission regions in the outer divertor and main limiter during ICRH compared to ohmic operation, mid-plane localized heating of the main limiter associated with fast-ion impact is observed which exceeds the divertor heat flux. During LHCD operation, several localized regions of increased brightness associated with hot spots are interpreted as heating due to localized density peaking, which re-iterates the importance of imaging continuum emission for subtraction. These sources of surface heating exacerbate plasma-material interactions at the device wall and may require additional mitigation if they cannot be avoided in future machines.
    Plasma Physics and Controlled Fusion 12/2013; 55(12):5010-. · 2.37 Impact Factor
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    ABSTRACT: Radio frequency (RF) sheath rectification is a leading mechanism suspected of causing anomalously high erosion of plasma facing materials in RF-heated plasmas on Alcator C-Mod. An extensive experimental survey of the plasma potential (ΦP) in RF-heated discharges on C-Mod reveals that significant ΦP enhancement (>100 V) is found on outboard limiter surfaces, both mapped and not mapped to active RF antennas. Surfaces that magnetically map to active RF antennas show ΦP enhancement that is, in part, consistent with the recently proposed slow wave rectification mechanism. Surfaces that do not map to active RF antennas also experience significant ΦP enhancement, which strongly correlates with the local fast wave intensity. In this case, fast wave rectification is a leading candidate mechanism responsible for the observed enhancement.
    Journal of Nuclear Materials 07/2013; · 2.02 Impact Factor
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    ABSTRACT: Lower hybrid current drive (LHCD) is an attractive option for non-inductive tokamak operation due to its high current drive efficiency and ability to drive current off axis. The parameters of the Alcator C-Mod LHCD system (f0 = 4.6 GHz, B 5.5 T, ) are similar to the proposed LHCD system on ITER. This paper will describe improvements in LHCD technology on C-Mod designed to increase single-pass absorption at high , extend pulse length (to >3 s), and increase power delivered to the plasma (to ~2 MW). Modelling of lower hybrid (LH) wave propagation indicates that the observed loss of LHCD efficiency at higher can be mitigated by enhancing the single pass power absorption through use of an off mid-plane launcher. The four rows of the launcher are located above the mid-plane (with Ip and B both clockwise viewing from the top down) in order to exploit the poloidal upshift of n|| as rays propagate from the antenna into the plasma. The transmitter protection system (TPS) was redesigned to model the coolant temperature in real time and shut off the klystron beam voltage if the coolant is close to boiling. The TPS upgrade has been installed and operated on C-Mod for pulses up to 4.5 s into dummy loads and 1.0 s into the plasma. A new movable local LH launcher protection limiter was designed to reduce reflection coefficients across a wide range of launcher positions. Finally, a high power waveguide double-stub tuner is under development to provide feedback controlled load matching to reduce power reflected from the antenna under poor coupling conditions.
    Nuclear Fusion 06/2013; 53(7):073012. · 2.73 Impact Factor
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    ABSTRACT: Long-lived (1, 1) 'snake' modes were discovered nearly three decades ago, but basic questions regarding their formation, stability, and superb particle confinement—shown by surviving tens to hundreds of sawtooth cycles—have remained unanswered. High-resolution spectroscopic imaging diagnostics permit studies of heavy-impurity-ion snakes with unprecedented temporal and spatial resolution, making it possible to positively identify the SXR signals with specific ion charge states and to infer, for the first time, the perturbed impurity density, Zeff, and resistivity at the centre of these long-lived helical modes. The results show a new scenario for the formation of heavy-impurity-ion snakes, which can begin as a broad 1/1 kink asymmetry of the central impurity-ion density, that grows and undergoes a seamless transition to a large crescent-shaped helical island-like structure inside q < 1, with a regularly sawtoothing core. This type of formation departs strongly from the nonlinear island model based on a modified Rutherford equation proposed originally to describe the pellet-induced snakes and expanded further to account for the impurity effects (e.g. and ). These new high-resolution observations show details of their evolution and the accompanying sawtooth oscillations that suggest important differences between the density and temperature dynamics, ruling out a purely pressure-driven process. Instead, many features arise naturally from nonlinear interactions in a 3D MHD model that separately evolves the plasma density and temperature.
    Nuclear Fusion 04/2013; 53(4):043019. · 2.73 Impact Factor
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    ABSTRACT: 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.
    Physical Review Letters 02/2013; 110(6):065006. · 7.73 Impact Factor
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    ABSTRACT: Deuterium high-confinement (H-mode) plasmas, lasting up to 3.45 s, have been generated in the EAST by ion cyclotron range of frequency (ICRF) heating. H-mode access was achieved by coating the molybdenum-tiled first wall with lithium to reduce the hydrogen recycling from the wall. H-mode plasmas with plasma currents between 0.4 and 0.6 MA and axial toroidal magnetic fields between 1.85 and 1.95 T were generated by 27 MHz ICRF heating of deuterium plasma with hydrogen minority. The ICRF input power required to access the H-mode was 1.6–1.8 MW. The line-averaged density was in the range (1.83–2.3) × 1019 m−3. 200–500 Hz type-III edge localized mode activity was observed during the H-mode phase. The H-mode confinement factor, H98IPB(y, 2), was ~0.7.
    Nuclear Fusion 01/2013; 53(2). · 2.73 Impact Factor
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    ABSTRACT: This paper describes the operation of a double stub fast ferrite tuner (FFT) that we have designed for the Alcator C-Mod 4.6GHz Lower Hybrid Current Drive (LHCD) system. This FFT is unique because it uses a single electromagnet coil and permanent magnet on each tuning stub. The ferrite is located on the center of the broad face of the waveguide. The FFT is required to withstand over 200kW of power (20kW/cm2) at high VSWR (>5) for 1-3 second pulses spaced 10 minutes apart. Breakdown measurements and fabrication considerations will be discussed. Also, simulation of thermal conditions will be shown. The FFT will be computer controlled and must react to matching a load in a few hundred microseconds. This puts a severe requirement on power supply response time and its variation. In addition, the calculation time of the controlling software algorithms must be considered as well as the diffusion time of the controlling magnetic field through the waveguide wall. We will discuss these requirements and what we have done to meet them.
    Fusion Engineering (SOFE), 2013 IEEE 25th Symposium on; 01/2013
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    ABSTRACT: Recent ICRF heating experiments in HT-7 and EAST tokomaks devices are presented. Experimental results in the HT-7 device showed that the ions can be directly heated by the Mode Converted Ion Bernstein Waves (MC IBW). Numerical simulation and experimental evidence confirmed that the interaction between the MC IBW and 7Li ions at the first ion cyclotron harmonic resonance of 7Li is the main mechanism for RF power deposition. In the EAST tokomaks, density pump-out was observed during L-mode discharges at a high electron density of 4.0×1019 m-3. Increases of both the electron temperature by 1.0 keV and the stored energy by 30 kJ have been obtained. Efficient ions and electrons heating were observed with the H cyclotron resonance layer at plasma center. The electrons are predominantly heated by collisions with high energy minority ions, not directly heated by the MC waves. In 2012 ICRF campaign at EAST, H mode target by ICRF and 3He minority heating have been investigated. The ICRF-generated H-mode plasma firstly was successfully achieved. Plasma density and stored energy are greatly increased during the H mode phase. The central electron temperature has an increase of 300eV, which is opposite to the H-mode plasma target by LHCD alone. In the case of 3He minority heating, the first efficient 3He ions heating were obtained in the deuterium target plasma. In combination with ICRF and LHW, the ICRF heating efficiency is much higher than that of ICRF only since the LHW preheats background plasma to enhance the absorption of ICRF. However, combined operation of ICRF and LH often results in a degradation of the LH wave coupling due to RF sheaths induced local density modification by ICRF antenna. When a reciprocating Langmuir probe is magnetically connected to a powered ICRH antenna, large RF-induced potentials up to 90V are observed near the location of RF limiters.
    24th IAEA Fusion Energy Conference, San Diego, USA; 10/2012
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    ABSTRACT: A swept-frequency X-mode reflectometer has been built on Alcator C-Mod to measure the scrape-off layer (SOL) density profiles adjacent to the lower hybrid launcher. The reflectometer system operates between 100 and 146 GHz at sweep rates from 10 μs to 1 ms and covers a density range of ∼10(16)-10(20) m(-3) at B(0) = 5-5.4 T. This paper discusses the analysis of reflectometer density profiles and presents first experimental results of SOL density profile modifications due to the application of lower hybrid range-of-frequencies power to L-mode discharges. Comparison between density profiles measured by the X-mode reflectometer and scanning Langmuir probes is also shown.
    The Review of scientific instruments 10/2012; 83(10):10E309. · 1.52 Impact Factor
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    ABSTRACT: In the Alcator C-Mod tokamak, strong, steady-state variations of molybdenum density within a flux surface are routinely observed in plasmas using hydrogen minority ion cyclotron resonant heating. In/out asymmetries, up to a factor of 2, occur with either inboard or outboard accumulation depending on the major radius of the minority resonance layer. These poloidal variations can be attributed to the impurity's high charge and large mass in the neoclassical parallel force balance. The large mass enhances the centrifugal force, causing outboard accumulation while the high charge enhances ion-impurity friction and makes impurities sensitive to small poloidal variations in the plasma potential. Quantitative comparisons between existing parallel high-Z impurity transport theories and experimental results for r/a < 0.7 show good agreement when the resonance layer is on the high-field side of the tokamak but disagree substantially for low-field side heating. Ion-impurity friction is insufficient to explain the experimental results, and the accumulation of impurity density on the inboard side of flux surface is shown to be driven by a poloidal potential variation due to magnetic trapping of non-thermal, cyclotron heated minority ions. Parallel impurity transport theory is extended to account for cyclotron effects and shown to agree with experimentally measured impurity density asymmetries.
    Plasma Physics and Controlled Fusion 03/2012; 54(4):045004. · 2.37 Impact Factor
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    ABSTRACT: Plasma heating using fast waves was successfully performed on the Experimental Advanced Superconducting Tokamak (EAST) in the H minority regime in deuterium plasmas at 27 MHz and Bo = 2.0 T. With 1.0 MW of ion cyclotron range of frequency (ICRF) power injected at a line-averaged electron density of 4.0 × 1019 m−3, the electron temperature increased from 1.0 keV to above 2.0 keV and the loop voltage dropped. An increase in the stored energy by 30 kJ was obtained. The first H-mode plasma of 6.4 s was achieved with a combination of lower hybrid wave and ICRF heating. Density pump-out was observed during L-mode discharges at a high electron density of 4.0 × 1019 m−3. In these discharges, re-attachment of the plasma was observed when ICRF power was applied.
    Nuclear Fusion 03/2012; 52(3):032002. · 2.73 Impact Factor
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    ABSTRACT: Radio frequency (RF) rectification of the plasma sheath is being actively studied on C-Mod as a likely mechanism that leads to prohibitively high molybdenum levels in the plasma core of ion cyclotron RF (ICRF) heated discharges. We installed emissive, ion sensitive, Langmuir, and 3-D B-dot probes to quantify the plasma potentials ({Phi}{sub P}) in ICRF and lower hybrid (LH) heated discharges. Two probe sets were mounted on fixed limiter surfaces and one set of probes was mounted on a reciprocating (along the major radius) probe. Initial results showed that RF rectification is strongly dependent on the local plasma density and not on the local RF fields. The RF sheaths had a threshold-like appearance at the local density of {approx}10{sup 16} m-{sup 3}. Radial probe scans revealed that the RF sheaths peaked in the vicinity of the ICRF limiter surface, agreeing with a recent theory. The highest {Phi}{sub P}'s were observed on magnetic field lines directly mapped to the active ICRF antenna. Measurements in LH heated plasmas showed a strong {Phi}{sub P} dependence on the parallel index of refraction n{sub ||} of the launched LH waves: {Phi}{sub P} is greater at lower n{sub ||}. Little dependence was observed on the local plasma density.
    AIP Conference Proceedings. 12/2011; 1406(1).
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    ABSTRACT: A swept-frequency X-mode reflectometer has being installed on Alcator C-Mod to measure the SOL density profiles at three poloidal locations adjacent to the new Lower Hybrid Launcher. First results on density profile modifications at the LH launcher due to ICRH or LHCD non-linear effects will be presented. Experimental measurements indicate that the application of LH power creates a density depletion near the LH launcher, which is consistent with the influence of a ponderomotive force. At high ne, LH power increases the density in the far SOL. Application of low ICRF power decreases the density in front of the LH launcher, which may be consistent with ICRF sheath induced convective cells. Preliminary results, however, indicate that field line mapping and increasing ICRF power do not modify the density profile significantly.
    12/2011;
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    ABSTRACT: We have modulated on-axis ICRF minority heating to trigger fluctuations and core electron transport in Alcator C-Mod Internal Transport Barriers (ITB's). Temperature swings of 50% produced strong bursts of density fluctuations, measured by phase contrast imaging (PCI), while edge fluctuations from reflectometry, Mirnov coils, and gas puff imaging (GPI) simultaneously diminished. The PCI fluctuations are in phase with sawteeth, further evidence that they originate within the ITB foot. Linear gyrokinetic analysis with GS2 shows TEMs are driven unstable in the ITB by the on-axis heating, as in Refs. [1,2]. Nonlinear gyrokinetic simulations of turbulence in the ITB are compared with fluctuation data using a synthetic diagnostic [1]. Strong ITB's were produced with high quality ion and electron profile data. [4pt] [1] D. R. Ernst et al., 20th IAEA Fusion Energy Conference (2006), Chengdu, China, paper IAEA-CN-149/TH/1-3. [2] D. R. Ernst et al., Phys. Plasmas 11 (2004) 2637.
    11/2011;
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    ABSTRACT: Demonstrating discharges on Alcator C-Mod with ITER characteristics is important to study plasma behavior during various phases and validate modeling used to project to ITER. Concentration has been on the rampup and rampdown phases. The flattop phase must meet, as close as possible, a number of parameters simultaneously; q95, elongation, n/nGr, beta-N, and H98. Experiments were performed to meet these parameters, lowering the toroidal field to 2.7 T and using 2^nd harmonic hydrogen minority heating. The lower field allowed more reliable access to these parameters. These discharges meet the ITER parameters closely, with the n/nGr value reaching 0.72 approaching the ITER value of 0.85, and were sustained for 0.5 to 1 s. EDA H-modes were obtained, showing the quasi-coherent mode at about 100 kHz, with some intermittent ELMy behavior. In addition, MHD modes are observed in the 10-25 kHz range with toroidal mode numbers n=2,3, which appear to be correlated with increasing betaN. Work supported by DE-AC02-09CH11466 and DE-FC02-99ER54512.
    11/2011;
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    ABSTRACT: The design, construction and initial results of a new lower hybrid current drive (LHCD) launcher on Alcator C-Mod (Hutchinson et al 1994 Phys. Plasmas 1 1511) are presented. The new LHCD launcher (LH2) is based on a novel splitter concept which evenly distributes the microwave power in four ways in the poloidal direction. This design allows for simplification of the feeding structure while keeping the flexibility to vary the peak launched toroidal index of refraction, Ntoroidal, from −3.8 to 3.8. An integrated model predicts good plasma coupling over a wide range of edge densities, while poloidal variations of the edge density are found to affect the evenness of power splitting in the poloidal direction. The measured transmission loss is about 30% lower than the previous launcher, and a clean Ntoroidal spectrum has been confirmed. Power handling capability exceeding an empirical weak conditioning limit and reliable operation up to 1.1 MW net LHCD power have been achieved. A survey of antenna–plasma coupling shows the existence of a millimetric vacuum gap in front of the launcher. Fully non-inductive, reversed shear plasma operation has been demonstrated and sustained for multiple current diffusion times. The current drive efficiency, ηLH ≡ neR0Ip/PLH, of these plasmas is (0.2–0.25) × 1020 m−2A W−1, which is in agreement with the expected efficiency on the International Thermonuclear Experimental Reactor (ITER).
    Nuclear Fusion 09/2011; 51(10):103024. · 2.73 Impact Factor
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    ABSTRACT: 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. (C) 2011 American Institute of Physics. [doi: 10.1063/1.3566059]
    Physics of Plasmas 05/2011; 18(5). · 2.38 Impact Factor
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    ABSTRACT: 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.
    Physics of Plasmas 04/2011; 18(5):056105-056105-13. · 2.38 Impact Factor
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    ABSTRACT: The physics that sets the width of the power exhaust channel in a tokamak scrape-off layer and its scaling with engineering parameters is of fundamental importance for reactor design, yet it remains to be understood. An extensive array of divertor heat flux diagnostics was recently commissioned in Alcator C-Mod with the aim of improving our understanding. Initial results are reported from EDA H-mode discharges in which plasma current, input power, toroidal field and magnetic topology were varied. The integral width of the outer divertor heat flux footprint is found to lie in the range of 3–5mm mapped to the mid-plane. Widths are insensitive to single versus double-null topology and the magnitude of toroidal field. Pedestal physics appears to largely determine these widths; a dependence of width on plasma thermal energy is noted, yielding a reduction in width as plasma current is increased for the best EDA H-modes.
    Journal of Nuclear Materials 01/2011; 415(1). · 2.02 Impact Factor

Publication Stats

321 Citations
81.93 Total Impact Points

Institutions

  • 2004–2013
    • Princeton University
      • Princeton Plasma Physics Laboratory
      Princeton, New Jersey, United States
  • 1998–2011
    • Massachusetts Institute of Technology
      • Plasma Science and Fusion Center (PSFC)
      Cambridge, Massachusetts, United States
  • 2004–2007
    • University of Texas at Austin
      • Fusion Research Center
      Austin, TX, United States