Resistive wall stabilization of high-beta plasmas in DIII–D

Strait, E. J.; Bialek, J.; Bogatu, N.; Chance, M. S.; Chu, M. S.; Edgell, D. H.; Garofalo, A. M.; Jackson, G.L.; Jensen, T. H.; Johnson, L. C.; Kim, J.S.; Haye, R.J. La; Navratil, G. A.; Okabayashi, M.; Reimerdes, H.; Scoville, J. T.; Turnbull, A. D.; Walker, M.L.

doi:10.1088/0029-5515/43/6/306

Cited by 93 publications

(125 citation statements)

References 23 publications

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“…These results represent a factor of 3-4 increase in duration of sustained beta above the n=1 no-wall limit relative to performance prior (shown in red) to improvements in plasma shaping [42] and error field correction and mode control, and a factor of 2 increase in duration above the no-wall limit relative to plasmas including improved shaping but without error field correction and mode control [41,42]. Thus, control of the resonant field amplification and resistive wall modes is a very important tool for performance optimization in NSTX as previously observed in high-beta conventional aspect ratio tokamaks [46,47].…”

Section: Low-β Locked-mode Threshold Scalingsmentioning

confidence: 88%

Progress in Understanding Error-field Physics in NSTX Spherical Torus Plasmas

Ménard¹

2010

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Abstract. The low aspect ratio, low magnetic field, and wide range of plasma beta of NSTX plasmas provide new insight into the origins and effects of magnetic field errors. An extensive array of magnetic sensors has been used to analyze error fields, to measure error field amplification, and to detect resistive wall modes in real time. The measured normalized error-field threshold for the onset of locked modes shows a linear scaling with plasma density, a weak to inverse dependence on toroidal field, and a positive scaling with magnetic shear. These results extrapolate to a favorable error field threshold for ITER. For these low-beta locked-mode plasmas, perturbed equilibrium calculations find that the plasma response must be included to explain the empirically determined optimal correction of NSTX error fields. In high-beta NSTX plasmas exceeding the n=1 no-wall stability limit where the RWM is stabilized by plasma rotation, active suppression of n=1 amplified error fields and the correction of recently discovered intrinsic n=3 error fields have led to sustained high rotation and record durations free of low-frequency core MHD activity. For sustained rotational stabilization of the n=1 RWM, both the rotation threshold and magnitude of the amplification are important. At fixed normalized dissipation, kinetic damping models predict rotation thresholds for RWM stabilization to scale nearly linearly with particle orbit frequency. Studies for NSTX find that orbit frequencies computed in general geometry can deviate significantly from those computed in the high aspect ratio and circular plasma cross-section limit, and these differences can strongly influence the predicted RWM stability. The measured and predicted RWM stability is found to be very sensitive to the E × B rotation profile near the plasma edge, and the measured critical rotation for the RWM is approximately a factor of two higher than predicted by the MARS-F code using the semi-kinetic damping model.

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Section: Low-β Locked-mode Threshold Scalingsmentioning

confidence: 88%

Progress in Understanding Error-field Physics in NSTX Spherical Torus Plasmas

Ménard¹

2010

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“…We have also found that fixed and free boundary modes for n = 2 and 3 are stable in DCON. This is not surprising since the β N for this dual equilibrium is 4.6, a value yielding stable plasmas in some finite aspect ratio machines [14].…”

mentioning

confidence: 83%

Dual equilibrium in a finite aspect ratio tokamak

Gourdain

Leboeuf²

2008

Physics Letters A

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“…As in the DIII-D C-Coil [11], the proposed RWM Coil can be operated in a duplex fashion to both correct "dc" error field and to feedback stabilize "ac" RWMs.…”

Section: −1mentioning

confidence: 99%

Physics Basis for the Advanced Tokamak Fusion Power Plant ARIES-AT

Jardin¹,

Kessel²,

Mau³

et al. 2003

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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.

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Resistive wall stabilization of high-beta plasmas in DIII–D

Cited by 93 publications

References 23 publications

Progress in Understanding Error-field Physics in NSTX Spherical Torus Plasmas

Progress in Understanding Error-field Physics in NSTX Spherical Torus Plasmas

Dual equilibrium in a finite aspect ratio tokamak

Physics Basis for the Advanced Tokamak Fusion Power Plant ARIES-AT

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