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4 results on '"null D.W. Johnson"'

1. Physics Considerations in the Design of NCSX

Author

null G.H. Neilson, null M.C. Zarnstorff, null L.P. Ku, null E.A. Lazarus, null P.K. Mioduszewski, null M. Fenstermacher, null E. Fredrickson, null G.Y. Fu, null A. Grossman, null P.J. Heitzenroeder, null R.H. Hatcher, null S.P. Hirshman, null S.R. Hudson, null D.W. Johnson, null H.W. Kugel, null J.F. Lyon, null R. Majeski, null D.R. Mikkelsen, null D.A. Monticello, null B.E. Nelson, null N. Pomphrey, null W.T. Reiersen, null A.H. Reiman, null P.H. Rutherford, null J.A. Schmidt, null D.A. Spong, and null D.J. Strickler

Subjects
Physics, business.industry, Nuclear engineering, Divertor, Ripple, Electrical engineering, National Compact Stellarator Experiment, Plasma, Collisionality, law.invention, law, Beta (plasma physics), Orbit (dynamics), business, Stellarator
Abstract

Compact stellarators have the potential to make steady-state, disruption-free magnetic fusion systems with beta approximately 5% and relatively low aspect ratio (R/ < 4.5) compared to most drift-optimized stellarators. Magnetic quasi-symmetry can be used to reduce orbit losses. The National Compact Stellarator Experiment (NCSX) is designed to test compact stellarator physics in a high-beta quasi-axisymmetric configuration and to determine the conditions for high-beta disruption-free operation. It is designed around a reference plasma with low ripple, good magnetic surfaces, and stability to the important ideal instabilities at beta approximately 4%. The device size, available heating power, and pulse lengths provide access to a high-beta target plasma state. The NCSX has magnetic flexibility to explore a wide range of equilibrium conditions and has operational flexibility to achieve a wide range of beta and collisionality values. The design provides space to accommodate plasma-facing components for divertor operation and ports for an extensive array of diagnostics.

Published
2002

2. Results of NSTX Heating Experiments

Author

null D. Mueller, null M. Ono, null M.G. Bell, null R.E. Bell, null M. Bitter, null C. Bourdelle, null D.S. Darrow, null P.C. Efthimion, null E.D. Fredrickson, null D.A Gates, null R.J. Goldston, null L.R. Grisham, null R.J. Hawryluk, null K.W. Hill, null J.C. Hosea, null S.C. Jardin, null H. Ji, null S.M. Kaye, null R. Kaita, null H.W. Kugel, null D.W. Johnson, null B.P. LeBlanc, null R. Majeski, null E. Mazzucato, null S.S. Medley, null J.E. Menard, null H.K. Park, null S.F. Paul, null C.K. Phillips, null M.H. Redi, null A.L. Rosenberg, null C.H. Skinner, null V.A. Soukhanovskii, null B. Stratton, null E.J Synakowski, null G. Taylor, null J.R. Wilson, null S.J. Zweben, null Y-K.M. Peng, null R. Barry, null T. Bigelow, null C.E. Bush, null M. Carter, null R. Maingi, null M. Menon, null P.M. Ryan, null D.W. Swain, null J. Wilgen, and null 37 additional authors

Subjects
Physics, Toroid, Physics::Plasma Physics, Beta (plasma physics), Torus, Plasma, Magnetohydrodynamics, Atomic physics, Scaling, Neutral beam injection, Bootstrap current
Abstract

The National Spherical Torus Experiment (NSTX) at Princeton is designed to assess the potential of the low-aspect-ratio spherical torus concept for magnetic plasma confinement. The plasma has been heated by up to 5 MW of neutral beam injection, NBI, at an injection energy of 90 keV and up to 6 MW of high harmonic fast wave, HHFW, at 30 MHz. NSTX has achieved beta T of 32%. A variety of MHD phenomena have been observed to limit eta. NSTX has now begun addressing E scaling, eta limits and current drive issues. During the NBI heating experiments, a broad Ti profile with Ti up to 2 keV, Ti > Te and a large toroidal rotation. Transport analysis suggests that the impurity ions have diffusivities approaching neoclassical. For L-Mode plasmas, E is up to two times the ITER-89P L-Mode scaling and exceeds the ITER-98pby2 H-Mode scaling in some cases. Transitions to H-Mode have been observed which result in an approximate doubling of tE. after the transition in some conditions. During HH FW heating, Te > Ti and Te up to 3.5 keV were observed. Current drive has been studied using coaxial helicity injection (CHI), which has produced 390 kA of toroidal current and HHFW, which has produced H-modes with significant bootstrap current fraction at low Ip, high q and high{sub etap}.

Published
2002

3. Stability and Confinement Properties of Auxiliary Heated NSTX Discharges

Author

null J.E. Menard, null R.E.Bell, null C. Bourdelle, null D.S. Darrow, null E.D. Fredrickson, null D.A. Gates, null L.R. Grisham, null S.M. Kaye, null B.P. LeBlanc, null R. Maingi, null S.S. Medley, null D. Mueller, null F. Paoletti, null S.A. Sabbagh, null D. Stutman, null D.W. Swain, null J.R. Wilson, null M.G. Bell, null J.M. Bialek, null C.E.Bush, null J.C. Hosea, null D.W. Johnson, null R. Kaita, null H.W. Kugel, null R.J. Maqueda, null M. Ono, null Y-K.M. Peng, null C.H. Skinner, null V.A. Soukhanovskii, null E.J. Synakowski, null G. Taylor, null G.A. Wurden, and null and S.J. Zweben

Subjects
Electron density, Chemistry, Thomson scattering, Electron temperature, Radius, Plasma, Atomic physics, Spherical tokamak, Energy source, Dimensionless quantity
Abstract

The National Spherical Torus Experiment (NSTX) is a spherical tokamak with nominal plasma major radius R(subscript ''0'') = 0.85 m, minor radius a = 0.66 m, and aspect ratio A > 1.28. Typical discharge parameters are plasma current I (subscript ''p'') = 0.7-1.4 MA, toroidal magnetic field B(subscript ''t0'') = 0.25-0.45 Tesla at major radius R(subscript ''0''), elongation = 1.7-2.2, triangularity 0.3-0.5, line-average electron density = 2-5 x 10(superscript ''19'') m(superscript ''-3''), electron temperature T(subscript ''e'')(0) = 0.5-1.5 keV, and ion temperature T(subscript ''i'')(0) = 0.5-2 keV. The NSTX auxiliary heating systems can routinely deliver 4.5 MW of 80-keV deuterium neutral beams and 3 MW of 30-MHz high-harmonic fast-wave power. Kinetic profile diagnostics presently include a 10-channel, 30-Hz multipulse Thomson scattering system (MPTS), a 17-channel charge-exchange recombination spectroscopy (CHERS) system, a 48-chord ultra-soft X-ray (USXR) array, and a 15-chord bolometry array. Initial experiments utilizing auxiliary heating on NSTX have focused on MHD stability limits, confinement trends, studying H-mode characteristics, and performing initial power balance calculations.

Published
2001

4. Observations Concerning the Injection of a Lithium Aerosol into the Edge of TFTR Discharges

Author

null D.K. Mansfield, null D.W. Johnson, null B. Grek, null H. Kugel, null M.G. Bell, and null et al

Subjects
Core electron, Chemistry, Electric field, Limiter, chemistry.chemical_element, Neutron, Lithium, Plasma, Atomic physics, Tokamak Fusion Test Reactor, Aerosol
Abstract

A new method of actively modifying the plasma-wall interaction was tested on the Tokamak Fusion Test Reactor. A laser was used to introduce a directed lithium aerosol into the discharge scrape-off layer. The lithium introduced in this fashion ablated and migrated preferentially to the limiter contact points. This allowed the plasma-wall interaction to be influenced in situ and in real time by external means. Significant improvement in energy confinement and fusion neutron production rate as well as a reduction in the plasma Zeff have been documented in a neutral-beam-heated plasma. The introduction of a metallic aerosol into the plasma edge increased the internal inductance of the plasma column and also resulted in prompt heating of core electrons in Ohmic plasmas. Preliminary evidence also suggests that the introduction of an aerosol leads to both edge poloidal velocity shear and edge electric field shear.

Published
2000

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