T. K. Mau

University of California, Los Angeles, Los Angeles, CA, United States

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Publications (125)70.16 Total impact

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    ABSTRACT: Studies of radiofrequency (RF) coupling and heating to ignition of a conceptual tokamak fusion reactor by means of the fast magnetosonic wave at the first ion cyclotron harmonic (ω = 2ωci) arepresented. First, the current status of fast magnetosonic wave propagation and heating is briefly reviewed. Next, a spatially averaged, time-dependent start-up model is used to describe the role of RF heating for ignition of a tokamak reactor. The model shows that 240 MJ corresponding to an RF power level of 80 MW for a three-second period is quite sufficient to ignite a 700 MW(e) fusion power plant the size of NUWMAK. To couple the RF power to the torus, both external cavities coupled by apertures and poloidal coil systems are considered. Wave coupling from an external vacuum-filled cavity to fast magnetosonic waves via apertures provides matching through an impedance transformation from a co-axial feed to the plasma wave modes.
    Nuclear Fusion 01/2011; 19(9):1171. · 2.73 Impact Factor
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    ABSTRACT: Supplementary wave heating is examined for two-ion-species plasmas from the fundamental tritium cyclotron resonance (ω = ωCT) to the first deuteron harmonic (ω = 2ωCD). At the first harmonic, it is found that a sufficiently large parallel wavelength greatly reduces the region of confluence between the fast wave and ion Bernstein wave. Power absorption considerations neglecting mode conversion processes show that first-harmonic heating efficiency is relatively insensitive to the species concentrations and that a dominant ion heating can be expected. For minority heating at the fundamental deuteron resonance, it is shown that inclusion of the two-ion hybrid resonance absorption considerably lowers the Q from the case where only minority species cyclotron damping is considered.
    Nuclear Fusion 01/2011; 17(2):297. · 2.73 Impact Factor
  • S.C. Chiu, T.K. Mau
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    ABSTRACT: Analytical and numerical results of physical processes taking place around the second-harmonic resonance surface in ICRF heating are presented. It is shown that (1) symmetry of transmission coefficients follow from Onsager's reciprocity relation of the dielectric tensor, and (2) direct dissipation around the cyclotron harmonic layer is mostly due to the Bernstein branch and depends on k||, becoming low for low k||. The latter has the consequence that mode conversion and reflection are sensitively reduced by damping, but transmission is not.
    Nuclear Fusion 01/2011; 23(12):1613. · 2.73 Impact Factor
  • I.R. Shokair, R.W. Conn, T.K. Mau
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    ABSTRACT: The combined theory of ion cyclotron wave coupling, mode conversion and absorption is developed for bounded inhomogeneous plasmas in slab geometry. The plasma response to the RF waves is described by the linearized Vlasov equation, with an equilibrium distribution function and a poloidal magnetic field profile which are selfconsistent solutions of the equilibrium Vlasov and Maxwell's equations. A general form of the wave equation, valid to arbitrary order in Larmor radius and including first order gradient terms, is derived. Expressions are also derived for the plasma intrinsic impedance. The wave equation, keeping second order terms in Larmor radius, is solved numerically for a plasma bounded by two conducting walls using the INTOR tokamak design parameters. It was found that the effect of the one-dimensional poloidal field is small. Also, a rough estimate of the error due to truncation to second order in Larmor radius was made and was found to be very small except in the high temperature regions. The effects of plasma boundedness and reflections from the mode conversion region on the coupling of waves from the RF source as a function of the plasma parameters and antenna spectra are studied. It is found that, at certain values of the plasma and antenna parameters, eigenmodes are formed, resulting in an increase in the loading impedance. It is verified that the impedance is not symmetric with respect to the spectrum in the poloidal direction, with highly peaked resonances. As the plasma temperature is increased, the resonances broaden and eventually disappear at very high temperatures. This is attributed to increased single pass absorption and decreasing reflection from the mode conversion region.
    Nuclear Fusion 01/2011; 28(8):1393. · 2.73 Impact Factor
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    The Review of scientific instruments 12/2009; 80(12):129901. · 1.52 Impact Factor
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    ABSTRACT: An integrated study of compact stellaralor power plants, ARIES-CS, has been conducted to explore attractive compact stellarator configurations and to define key research and development (RD) areas. The large size and mass predicted by earlier stellarator power plant studies had led to cost projections much higher than those of the advanced tokamak power plant. As such, the first major goal of the ARIES-CS research was to investigate if stellarator power plants can be made to be comparable in size to advanced tokamak variants while maintaining desirable stellarator properties. As stellarator fusion core components would have complex shapes and geometry, the second major goal of the ARIES-CS study was to understand and quantify, as much as possible, the impact of the complex shape and geometry of fusion core components. This paper focuses on the directions we pursued to optimize the compact stellarator as a fusion power plant, summarizes the major findings from the study, highlights the key design aspects and constraints associated with a compact stellarator, and identifies the major issues to help guide future RD. fusion power plant, compact stellarator.
    Fusion Science and Technology. 01/2008; 54(3):655-672.
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    ABSTRACT: Novel stellarator configurations have been devel-oped for ARIES-CS. These configurations are optimized to provide good plasma confinement and flux surface integrity at high beta. Modular coils have been designed for them in which the space needed for the breeding blanket and radiation shielding was specifically targeted such that reactors generating GW electrical powers would require only moderate major radii (10 m). These con-figurations are quasi-axially symmetric in the magnetic field topology and have small numbers of field periods (3) and low aspect ratios (6). The baseline design chosen for detailed systems and power plant studies has three field periods, aspect ratio 4.5, and major radius 7.5 m operating at b ; 6.5% to yield 1 GW of electric power. The shaping of the plasma accounts for 75% of the rotational transform. The effective helical ripples are very small (0.6% everywhere), and the energy loss of alpha particles is calculated to be 5% when operating in high-density regimes. An interesting feature in this configuration is that instead of minimizing all residues in the magnetic spectrum, we preferentially retained a small amount of the nonaxisymmetric mirror field. The pres-ence of this mirror and its associated helical field alters the ripple distribution, resulting in the reduced ripple-trapped loss of alpha particles despite the long connec-tion length in a tokamak-like field structure. Additionally, we discuss two other potentially attractive classes of con-figurations, both quasi-axisymmetric: one with only two field periods, very low aspect ratios (;2.5), and less complex coils, and the other with the plasma shaping designed to produce low-shear rotational transform so as to ensure the robustness and integrity of flux surfaces when operating at high b. KEYWORDS: power plant studies, quasi-axisymmetric stel-larators, configuration optimization Note: Some figures in this paper are in color only in the electronic version.
    Fusion Science and Technology. 01/2008; 54(3).
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    ABSTRACT: Divertor heat load distributions due to thermal and alpha particles have been assessed in an NCSX-based compact stellarator reactor. A divertor plate system is envisaged, with 4 plates per field period and covering 7% of the plasma surface area. The field-line tracing technique is employed; for thermal flux, the conventional approach is used, while for alphas, their characteristic exit pattern from the plasma and subsequent gyro- orbits are approximated. For the ARIES-CS reference design point (R=7.75 m, A=4.5, B=5.7 T, beta=6.4% and P<sub>net</sub>=1000 MW), combined peak heat loads in the 5 -18 MW/m<sup>2</sup> range on the plates have been obtained, assuming a 75% radiation fraction both in the core and at the edge, and a 5% alpha loss fraction. The alpha heat flux could be a dominant determining factor. Further optimization study is warranted to lower all peak heat loads to satisfy the accepted limit of les10 MW/m<sup>2</sup>.
    Fusion Engineering, 2007. SOFE 2007. 2007 IEEE 22nd Symposium on; 07/2007
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    ABSTRACT: The National Spherical Torus Experiment (NSTX) has made considerable progress in advancing the scientific understanding of high performance long-pulse plasmas needed for future spherical torus (ST) devices and ITER. Plasma durations up to 1.6 s (five current redistribution times) have been achieved at plasma currents of 0.7 MA with non-inductive current fractions above 65% while simultaneously achieving βT and βN values of 17% and 5.7 (%m T MA−1), respectively. A newly available motional Stark effect diagnostic has enabled validation of current-drive sources and improved the understanding of NSTX 'hybrid'-like scenarios. In MHD research, ex-vessel radial field coils have been utilized to infer and correct intrinsic EFs, provide rotation control and actively stabilize the n = 1 resistive wall mode at ITER-relevant low plasma rotation values. In transport and turbulence research, the low aspect ratio and a wide range of achievable β in the NSTX provide unique data for confinement scaling studies, and a new microwave scattering diagnostic is being used to investigate turbulent density fluctuations with wavenumbers extending from ion to electron gyro-scales. In energetic particle research, cyclic neutron rate drops have been associated with the destabilization of multiple large toroidal Alfven eigenmodes (TAEs) analogous to the 'sea-of-TAE' modes predicted for ITER, and three-wave coupling processes have been observed for the first time. In boundary physics research, advanced shape control has enabled studies of the role of magnetic balance in H-mode access and edge localized mode stability. Peak divertor heat flux has been reduced by a factor of 5 using an H-mode-compatible radiative divertor, and lithium conditioning has demonstrated particle pumping and results in improved thermal confinement. Finally, non-solenoidal plasma start-up experiments have achieved plasma currents of 160 kA on closed magnetic flux surfaces utilizing coaxial helicity injection.
    Nuclear Fusion 01/2007; 47(10):S645-S657. · 2.73 Impact Factor
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    ABSTRACT: Alpha loss is an important issue in stellarators, where the inherent non-axisymmetry of the magnetic geometry causes an appreciable magnetic field ripple along the flux surfaces inside the plasma. This results in a non-negligible fraction of alpha-particles being promptly lost from the plasma and hitting the plasma facing components (PFC) at energies close to their born value of 3.5 MeV. The PFC armor must not only accommodate the heat load from the alpha-particle flux but it must also accommodate the He implantation resulting from the high-energy alpha fluxes while providing the required lifetime. An effort is underway as part of the ARIES-CS study to characterize the alpha-particle loss for a compact stellarator and to explore engineering solutions to accommodate the impact of these alphas on the PFC’s. This paper summarizes the initial results from this effort.
    Journal of Nuclear Materials. 01/2007;
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    ABSTRACT: The National Spherical Torus Experiment (NSTX) has made considerable progress in advancing the scientific understanding of high performance long-pulse plasmas needed for future spherical torus (ST) devices and ITER. Plasma durations up to 1.6 s (five current redistribution times) have been achieved at plasma currents of 0.7 MA with non-inductive current fractions above 65% while simultaneously achieving betaT and betaN values of 17% and 5.7 (%m T MA-1), respectively. A newly available motional Stark effect diagnostic has enabled validation of current-drive sources and improved the understanding of NSTX 'hybrid'-like scenarios. In MHD research, ex-vessel radial field coils have been utilized to infer and correct intrinsic EFs, provide rotation control and actively stabilize the n = 1 resistive wall mode at ITER-relevant low plasma rotation values. In transport and turbulence research, the low aspect ratio and a wide range of achievable beta in the NSTX provide unique data for confinement scaling studies, and a new microwave scattering diagnostic is being used to investigate turbulent density fluctuations with wavenumbers extending from ion to electron gyro-scales. In energetic particle research, cyclic neutron rate drops have been associated with the destabilization of multiple large toroidal Alfven eigenmodes (TAEs) analogous to the 'sea-of-TAE' modes predicted for ITER, and three-wave coupling processes have been observed for the first time. In boundary physics research, advanced shape control has enabled studies of the role of magnetic balance in H-mode access and edge localized mode stability. Peak divertor heat flux has been reduced by a factor of 5 using an H-mode-compatible radiative divertor, and lithium conditioning has demonstrated particle pumping and results in improved thermal confinement. Finally, non-solenoidal plasma start-up experiments have achieved plasma currents of 160 kA on closed magnetic flux surfaces utilizing coaxial helicity injection.
    Nuclear Fusion - NUCL FUSION. 01/2007; 47(10).
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    ABSTRACT: The National Spherical Torus Experiment (NSTX) has explored the effects of shaping on plasma performance as determined by many diverse topics including the stability of global magnetohydrodynamic (MHD) modes (e.g., ideal external kinks and resistive wall modes), edge localized modes (ELMs), bootstrap current drive, divertor flux expansion, and heat transport. Improved shaping capability has been crucial to achieving βt ∼ 40%. Precise plasma shape control has been achieved on NSTX using real-time equilibrium reconstruction. NSTX has simultaneously achieved elongation κ ∼ 2.8 and triangularity δ ∼ 0.8. Ideal MHD theory predicts increased stability at high values of shaping factor S ≡ q95Ip/(aBt), which has been observed at large values of the S ∼ 37[MA/(m∙T)] on NSTX. The behavior of ELMs is observed to depend on plasma shape. A description of the ELM regimes attained as shape is varied will be presented. Increased shaping is predicted to increase the bootstrap fraction at fixed Ip. The achievement of strong shaping has enabled operation with 1 s pulses with Ip = 1 MA, and for 1.6 s for Ip = 700 kA. Analysis of the noninductive current fraction as well as empirical analysis of the achievable plasma pulse length as elongation is varied will be presented. Data are presented showing a reduction in peak divertor heat load due to increasing in flux expansion.
    Physics of Plasmas 05/2006; 13(5):056122-056122-7. · 2.38 Impact Factor
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    ABSTRACT: The National Spherical Torus Experiment Ono et al., Nucl. Fusion, 44, 452 2004 is targeting long pulse high performance, noninductive sustained operations at low aspect ratio, and the demonstration of nonsolenoidal startup and current rampup. The modeling of these plasmas provides a framework for experimental planning and identifies the tools to access these regimes. Simulations based on neutral beam injection NBI-heated plasmas are made to understand the impact of various modifications and identify the requirements for 1 high elongation and triangularity, 2 density control to optimize the current drive, 3 plasma rotation and/or feedback stabilization to operate above the no-wall limit, and 4 electron Bernstein waves EBW for off-axis heating/current drive H/CD. Integrated scenarios are constructed to provide the transport evolution and H/CD source modeling, supported by rf and stability analyses. Important factors include the energy confinement, Z eff , early heating/H mode, broadening of the NBI-driven current profile, and maintaining q0 and q min 1.0. Simulations show that noninductive sustained plasmas can be reached at I P = 800 kA, B T = 0.5 T, 2.5, N 5, 15%, f NI = 92%, and q0 1.0 with NBI H/CD, density control, and similar global energy confinement to experiments. The noninductive sustained high plasmas can be reached at I P = 1.0 MA, B T = 0.35 T, 2.5, N 9, 43%, f NI = 100%, and q0 1.5 with NBI H/CD and 3.0 MW of EBW H/CD, density control, and 25% higher global energy confinement than experiments. A scenario for nonsolenoidal plasma current rampup is developed using high harmonic fast wave H/CD in the early low I P and low T e phase, followed by NBI H/CD to continue the current ramp, reaching a maximum of 480 kA after 3.4 s. © 2006 American Institute of Physics.
    Physics of Plasmas 05/2006; 13(5). · 2.38 Impact Factor
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    ABSTRACT: In order to reduce recirculating power fraction to acceptable levels, the spherical torus concept relies on the simultaneous achievement of high toroidal β and high bootstrap fraction in steady state. In the last year, as a result of plasma control system improvements, the achievable plasma elongation on NSTX has been raised from κ ~ 2.1 to κ ~ 2.6—approximately a 25% increase. This increase in elongation has led to a substantial increase in the toroidal β for long pulse discharges. The increase in β is associated with an increase in plasma current at nearly fixed poloidal β, which enables higher βt with nearly constant bootstrap fraction. As a result, for the first time in a spherical torus, a discharge with a plasma current of 1 MA has been sustained for 1 s (0.8 s current flat-top). Data are presented from NSTX correlating the increase in performance with increased plasma shaping capability. In addition to improved shaping, H-modes induced during the current ramp phase of the plasma discharge have been used to reduce flux consumption and to delay the onset of MHD instabilities. Based on these results, a modelled integrated scenario, which has 100% non-inductive current drive with very high toroidal β, will also be discussed. The NSTX poloidal field coils are currently being modified to produce the plasma shape which is required for this scenario, which requires high triangularity (δ ~ 0.8) at elevated elongation (κ ~ 2.5). The other main requirement of steady state on NSTX is the ability to drive a fraction of the total plasma current with RF waves. The results of high harmonic fast wave heating and current drive studies as well as electron Bernstein wave emission studies will be presented.
    Nuclear Fusion 01/2006; 46(3):S22. · 2.73 Impact Factor
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    ABSTRACT: An advanced tokamak plasma configuration is developed based on equilibrium, ideal MHD stability, bootstrap current analysis, vertical stability and control, and poloidal field coil analysis. The plasma boundaries used in the analysis are forced to coincide with the 99% flux surface from the free-boundary equilibrium. Using an accurate bootstrap current model and external current drive profiles from ray tracing calculations in combination with optimized pressure profiles, βN values above 7.0 have been obtained. The minimum current drive requirement is found to lie at a lower βN of 6.0. The external kink mode is stabilized by a tungsten shell located at 0.33 times the minor radius and a feedback system. Plasma shape optimization has led to an elongation of 2.2 and triangularity of 0.9 at the separatrix. Vertical stability could be achieved by a combination of tungsten shells located at 0.33 times the minor radius and feedback control coils located behind the shield. The poloidal field coils were optimized in location and current, providing a maximum coil current of 8.6 MA. These developments have led to a simultaneous reduction in the power plant major radius and toroidal field from those found in a previous study [S.C. Jardin, C.E. Kessel, C.G. Bathke, D.A. Ehst, T.K. Mau, F. Najmabadi, T.W. Petrie, the ARIES Team, Physics basis for a reversed shear tokamak power plant, Fusion Eng. Design 38 (1997) 27].
    Fusion Engineering and Design. 01/2006;
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    ABSTRACT: The advanced tokamak is considered as the basis for a fusion power plant. The ARIES-AT design has an aspect ratio of A ≡ R/a = 4.0, an elongation and triangularity of κ = 2.20, δ = 0.90 (evaluated at the separatrix surface), a toroidal beta of β = 9.1% (normalized to the vacuum toroidal field at the plasma center), which corresponds to a normalized beta of β N ≡ 100 × β/(I P (MA)/a(m)B(T)) = 5.4. These beta values are chosen to be 10% below the ideal MHD stability limit. The bootstrap-current fraction is f BS ≡ I BS /I P = 0.91. This leads to a design with total plasma current I P = 12.8 MA, and toroidal field of 11.1 T (at the coil edge) and 5.8 T (at the plasma center). The major and minor radii are 5.2 and 1.3 m. The effects of H-mode edge gradients and the stability of this configuration to non-ideal modes is analyzed. The current drive system consists of ICRF/FW for on-axis current drive and a Lower Hybrid system for off-axis. Transport projections are presented using the drift-wave based GLF23 model. The approach to power and particle exhaust using both plasma core and scrape-off-layer radiation is presented.
    Fusion Engineering and Design. 01/2006; 80:25-62.
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    ABSTRACT: The ARIES-AT study was initiated to assess the potential of high-performance tokamak plasmas together with advanced technology in a fusion power plant and to identifying physics and technology areas with the highest leverage for achieving attractive and competitive fusion power in order to guide fusion R&D. The 1000-MWe ARIES-AT design has a major radius of 5.2 m, a minor radius of 1.3 m, a toroidal β of 9.2% (βN = 5.4) and an on-axis field of 5.6 T. The plasma current is 13 MA and the current-drive power is 35 MW. The ARIES-AT design uses the same physics basis as ARIES-RS, a reversed-shear plasma. A distinct difference between ARIES-RS and ARIES-AT plasmas is the higher plasma elongation of ARIES-AT (κx = 2.2) which is the result of a “thinner” blanket leading to a large increase in plasma β to 9.2% (compared to 5% for ARIES-RS) with only a slightly higher βN. ARIES-AT blanket is a simple, low-pressure design consisting of SiC composite boxes with a SiC insert for flow distribution that does not carry any structural load. The breeding coolant (Pb–17Li) enters the fusion core from the bottom, and cools the first wall while traveling in the poloidal direction to the top of the blanket module. The coolant then returns through the blanket channel at a low speed and is superheated to ∼1100 °C. As most of the fusion power is deposited directly into the breeding coolant, this method leads to a high coolant outlet temperature while keeping the temperature of the SiC structure as well as interface between SiC structure and Pb–17Li to about 1000 °C. This blanket is well matched to an advanced Brayton power cycle, leading to an overall thermal efficiency of ∼59%. The very low afterheat in SiC composites results in exceptional safety and waste disposal characteristics. All of the fusion core components qualify for shallow land burial under U.S. regulations (furthermore, ∼90% of components qualify as Class-A waste, the lowest level). The ARIES-AT study shows that the combination of advanced tokamak modes and advanced technology leads to an attractive fusion power plant with excellent safety and environmental characteristics and with a cost of electricity (4.7 ¢/kWh), which is competitive with those projected for other sources of energy.
    Fusion Engineering and Design. 01/2006;
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    ABSTRACT: The major objective of the National Spherical Torus Experiment (NSTX) is to understand basic toroidal confinement physics at low aspect ratio and high βT in order to advance the spherical torus (ST) concept. In order to do this, NSTX utilizes up to 7.5 MW of neutral beam injection, up to 6 MW of high harmonic fast waves (HHFWs), and it operates with plasma currents up to 1.5 MA and elongations of up to 2.6 at a toroidal field up to 0.45 T. New facility, and diagnostic and modelling capabilities developed over the past two years have enabled the NSTX research team to make significant progress towards establishing this physics basis for future ST devices. Improvements in plasma control have led to more routine operation at high elongation and high βT (up to ~40%) lasting for many energy confinement times. βT can be limited by either internal or external modes. The installation of an active error field (EF) correction coil pair has expanded the operating regime at low density and has allowed for initial resonant EF amplification experiments. The determination of the confinement and transport properties of NSTX plasmas has benefitted greatly from the implementation of higher spatial resolution kinetic diagnostics. The parametric variation of confinement is similar to that at conventional aspect ratio but with values enhanced relative to those determined from conventional aspect ratio scalings and with a BT dependence. The transport is highly dependent on details of both the flow and magnetic shear. Core turbulence was measured for the first time in an ST through correlation reflectometry. Non-inductive start-up has been explored using PF-only and transient co-axial helicity injection techniques, resulting in up to 140 kA of toroidal current generated by the latter technique. Calculated bootstrap and beam-driven currents have sustained up to 60% of the flat-top plasma current in NBI discharges. Studies of HHFW absorption have indicated parametric decay of the wave and associated edge thermal ion heating. Energetic particle modes, most notably toroidal Alfvén eigenmodes and fishbone-like modes result in fast particle losses, and these instabilities may affect fast ion confinement on devices such as ITER. Finally, a variety of techniques has been developed for fuelling and power and particle control.
    Nuclear Fusion 10/2005; 45(10):S168. · 2.73 Impact Factor
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    ABSTRACT: Divertor heat load evaluation studies are being carried out for an example compact stellarator reactor configuration modeled after the National Compact Stellarator Experiment (NCSX). Using the field line tracing technique, an initial divertor plate geometry has been obtained, with one plate per field period, which covers 15% of the plasma surface area. A plate heat load peaking factor of 14 is acceptable with reasonable core and SOL radiation fractions in order to satisfy the engineering constraint of 10 MW/m peak heat load. Assessment of heat flux due to ejected energetic alpha particles is equally important for their contribution to the heat load. A particle gyro-orbit code follows these particles in the scrape-off layer (SOL) until they hit the divertor plates and the first wall, while the distribution of lost particles on the plasma surface can be calculated with a Monte Carlo guiding center code with collisions
    Fusion Engineering 2005, Twenty-First IEEE/NPS Symposium on; 10/2005
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    ABSTRACT: Integrated scenario simulations are done for NSTX that address four primary objectives for developing advanced spherical torus (ST) configurations: high β and high βN inductive discharges to study all aspects of ST physics in the high β regime; non-inductively sustained discharges for flattop times greater than the skin time to study the various current drive techniques; non-inductively sustained discharges at high β for flattop times much greater than a skin time which provides the integrated advanced ST target for NSTX and non-solenoidal startup and plasma current rampup. The simulations done here use the tokamak simulation code and are based on a discharge 109070. TRANSP analysis of the discharge provided the thermal diffusivities for electrons and ions, the neutral beam deposition profile and other characteristics. CURRAY is used to calculate the high harmonic fast wave (HHFW) heating depositions and current drive. GENRAY/CQL3D is used to establish the heating and CD deposition profiles for electron Bernstein waves (EBW). Analysis of the ideal MHD stability is done with JSOLVER, BALMSC and PEST2. The simulations indicate that the integrated advanced ST plasma is reachable, obtaining stable plasmas with βT ≈ 40% at βN's of 7.7–9, IP = 1.0 MA and BT = 0.35 T. The plasma is 100% non-inductive and has a flattop of four skin times. The resulting global energy confinement corresponds to a multiplier of H98(y),2 = 1.5. The simulations have demonstrated the importance of HHFW heating and CD, EBW off-axis CD, strong plasma shaping, density control and early heating/H-mode transition for producing and optimizing these plasma configurations.
    Nuclear Fusion 07/2005; 45(8):814. · 2.73 Impact Factor

Publication Stats

330 Citations
70.16 Total Impact Points

Institutions

  • 1982–2011
    • University of California, Los Angeles
      • • Department of Mechanical and Aerospace Engineering
      • • School of Engineering and Applied Science
      Los Angeles, CA, United States
  • 1978–2011
    • University of Wisconsin, Madison
      • • Department of Nuclear Engineering
      • • Department of Electrical and Computer Engineering
      Madison, MS, United States
  • 1981–2009
    • University of California, San Diego
      • • Department of Mechanical and Aerospace Engineering (MAE)
      • • Marine Physical Laboratory (MPL)
      • • Center for Energy Research (CER)
      San Diego, California, United States
  • 2003
    • CSU Mentor
      • Department of Electrical & Computer Engineering
      Long Beach, California, United States
  • 2002
    • General Atomics
      San Diego, California, United States
    • University of the Pacific (California - USA)
      Stockton, California, United States
  • 1998–2002
    • Princeton University
      • Princeton Plasma Physics Laboratory
      Princeton, NJ, United States
  • 1995
    • Laboratory of Plasma Physics
      Paliseau, Île-de-France, France