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Yoon, S. W.*; Ahn, J.-W.*; Jeon, Y. M.*; Suzuki, Takahiro; Hahn, S. H.*; Ko, W. H.*; Lee, K. D.*; Chung, J. I.*; Nam, Y. U.*; Kim, J.*; et al.
Nuclear Fusion, 51(11), p.113009_1 - 113009_9, 2011/11
Times Cited Count：33 Percentile：79.01(Physics, Fluids & Plasmas)Typical ELMy H-mode discharges have been achieved on the KSTAR tokamak with the combined auxiliary heating of NBI and ECRH. The minimum external heating power required is about 1.1 MW at a line-averaged density higher than 1.4
10
m
and a toroidal field of 2 T. There is a clear indication of the increase of the L-H threshold power at densities lower than 
. The initial analysis of energy confinement time (
) predicted that was higher than the prediction of multi-machine scaling laws by a factor 1.4-1.6. However, when the contribution of fast ion confinement to the total energy was taken into account, better agreed with the scaling results. A clear increase of electron and ion temperature in the pedestal was observed in the H-mode phase but the core ion temperature did not change significantly. On the other hand, the toroidal rotation also increased over all radii in the H-mode phase. The measured ELM frequency was from 30-50 Hz and the drop of total energy appeared to be less than 5%. Between large ELM spikes, small/grassy ELMs were also identified when mixed heating of NBI and ECRH was applied.
Kurihara, Kenichi; Lister, J. B.*; Humphreys, D. A.*; Ferron, J. R.*; Treutterer, W.*; Sartori, F.*; Felton, R.*; Br
mond, S.*; Moreau, P.*; JET-EFDA Contributors*
Fusion Engineering and Design, 83(7-9), p.959 - 970, 2008/12
Times Cited Count：25 Percentile：81.47(Nuclear Science & Technology)The existing large and medium-size tokamaks are expected to explore more advanced operation scenarios toward the ITER and a future power reactor. To specify one or more solutions to keep a steady-state plasma with high performance, and to avoid plasma instabilities almost completely, a plasma control system for ITER should have two important aspects: Technical inheritance of the currently-working functions, and flexible or adaptive structure. First, we make review on the system configuration and essential functions employed in each plasma control system from the viewpoint of hardware as well as software. Second, we survey ITER control system requirements for the current CODAC design. Third, flexible structure in the plasma control system should be discussed. Finally, on the basis of the above discussion, we would like to envisage a future plasma control system for ITER and a fusion power plant.
Hender, T. C.*; Wesley, J. C.*; Bialek, J.*; Bondeson, A.*; Boozer, A. H.*; Buttery, R. J.*; Garofalo, A.*; Goodman, T. P.*; Granetz, R. S.*; Gribov, Y.*; et al.
Nuclear Fusion, 47(6), p.S128 - S202, 2007/06
Times Cited Count：916 Percentile：100(Physics, Fluids & Plasmas)no abstracts in English
Gribov, Y.*; Humphreys, D. A.*; Kajiwara, Ken*; Lazarus, E. A.*; Lister, J. B.*; Ozeki, Takahisa; Portone, A.*; Shimada, Michiya*; Sips, A. C. C.*; Wesley, J. C.*
Nuclear Fusion, 47(6), p.S385 - S403, 2007/06
Times Cited Count：133 Percentile：97.45(Physics, Fluids & Plasmas)This chapter describes the progress achieved in these areas in the tokamak experiments since the ITER Physics Basis (1999 Nucl. Fusion 39 2577) was written and the results of assessment of ITER to provide the plasma initiation and basic control. This chapter considers only plasma initiation and plasma basic control. The experiments on plasma initiation performed in DIII-D and JT-60U, as well as the theoretical studies performed for ITER, have demonstrated that, ITER can produce plasma initiation in a low toroidal electric field of 0.3V/m, if it is assisted by about 2MW of ECRF heating. The plasma basic control is described, which includes control of the plasma current, position and shape - the plasma magnetic control, as well as control of other plasma global parameters or their profiles - the plasma performance control.
Prater, R.*; La Haye, R. J.*; Luce, T. C.*; Petty, C. C.*; Strait, E. J.*; Ferron, J. R.*; Humphreys, D. A.*; Isayama, Akihiko; Lohr, J.*; Nagasaki, Kazunobu*; et al.
Nuclear Fusion, 47(5), p.371 - 377, 2007/05
Times Cited Count：58 Percentile：87.07(Physics, Fluids & Plasmas)The 
/
neoclassical tearing mode (NTM) has been observed to strongly degrade confinement and frequently lead to a disruption in high 
discharges in DIII-D if allowed to grow to large size. Stabilization of grown NTMs by application of highly localized electron cyclotron current drive (ECCD) at the island location has led to operation at increased plasma pressure, up to the no-wall kink limit. After the NTM is stabilized by the ECCD, the correct location for the current drive is maintained using information from real-time equilibrium reconstructions which include measurements from the motional Stark effect diagnostic. This same process is used alternatively to prevent the mode from ever growing, leading to performance at the pressure limit in high performance hybrid discharges with above 4%. Modeling using the modified Rutherford equation shows that the required power is in close agreement with the experimental threshold for prevention of the 2/1 NTM.
Prater, R.*; La Haye, R. J.*; Luce, T. C.*; Petty, C. C.*; Strait, E. J.*; Ferron, J. R.*; Humphreys, D. A.*; Isayama, Akihiko; Lohr, J.*; Nagasaki, Kazunobu*; et al.
Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 8 Pages, 2007/03
Onset of the m/n=2/1 neoclassical tearing mode (NTM) has been prevented in high-performance DIII-D discharges using localized electron cyclotron current drive (ECCD). Active tracking of the 
=2 surface location, using real-time equilibrium reconstructions with motional Stark effect data, allows the current drive to be maintained at the rational surface even in the absence of a detectable mode. With the application of this technique in DIII-D hybrid discharges, the 2/1 mode is avoided and good energy confinement is maintained for more than 1 second with at the estimated n=1 no-wall stability limit for ideal kink modes (
approximately equals 3.9 % and normalized beta 
approximately equals 3.2). The results can be understood through modeling using the modified Rutherford equation.
