Negative magnetic shear modes of operation in the Alcator C-Mod tokamak near the beta limit

Bonoli, P. T.; Porkoláb, M.; Ramos, J. J.; Nevins, W. M.; Kessel, C.

doi:10.1088/0741-3335/39/2/001

Cited by 25 publications

(31 citation statements)

References 21 publications

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“…9, using the TRANSP code with LSC [17], predicts 30-40 kA of LHCD, localized off axis with a peak at r/a=0.65. Extrapolation to the planned upgrade power of 2.7 MW predicts ~150 kA in comparable discharges, in line with advanced scenario modeling [18]. Further reduction of density would clearly be desirable.…”

Section: To 14 S (A) Notable Changes Include a Decrease In Electronsupporting

confidence: 57%

“…Simulation of these discharges continues in an effort to understand the evolution more clearly. In other related experiments [18] delay of q 0 = 1 onset has been demonstrated at I p = 0.8 MA (q 95 =4.2), in the range required for hybrid scenario experiments that will help assess the properties of this regime on ITER. Lower hybrid heating and current drive experiments on the Alcator C-Mod tokamak have been extended to detailed measurements of the driven current and comparisons to models as well as application of LHCD to plasma scenarios that will lead to eventual advance tokamak conditions for several current relaxation times.…”

Section: Use Of Lhcd During the Current Rampmentioning

confidence: 84%

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Lower Hybrid Heating and Current Drive on the Alcator C-Mod Tokamak

2009

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Abstract.On the Alcator C-Mod tokamak, lower hybrid current drive (LHCD) is being used to modify the current profile with the aim of obtaining advanced tokamak (AT) performance in plasmas with parameters similar to those that would be required on ITER. To date, power levels in excess of 1 MW at a frequency of 4.6 GHz have been coupled into a variety of plasmas. Experiments have established that LHCD on C-Mod behaves globally as predicted by theory. Bulk current drive efficiencies, n 20 I lh R/P lh ~ 0.25, inferred from magnetics and MSE are in line with theory. Quantitative comparisons between local measurements, MSE, ECE and hard x-ray bremsstrahlung, and theory/simulation using the GENRAY, TORIC-LH CQL3D and TSC-LSC codes have been performed. These comparisons have demonstrated the off-axis localization of the current drive, its magnitude and location dependence on the launched n || spectrum, and the use of LHCD during the current ramp to save volt-seconds and delay the peaking of the current profile. Broadening of the x-ray emission profile during ICRF heating indicates that the current drive location can be controlled by the electron temperature, as expected. In addition, an alteration in the plasma toroidal rotation profile during LHCD has been observed with a significant rotation in the counter current direction. Notably, the rotation is accompanied by peaking of the density and temperature profiles on a current diffusion time scale inside of the half radius where the LH absorption is taking place. PACS: 52.50.Sw, 52.55.Wq IntroductionThe Alcator C-Mod program has as one of its goals the aim of producing advanced tokamak (AT) discharges. In order to provide the current profile control required for obtaining reduced or reverse shear a lower hybrid (LH) rf power system [1] has been installed on the tokamak. One of the advantages of exploring the applicability of LH current drive on Alcator C-Mod is the equivalence of many of the important parameters, such as plasma density, magnetic field, and rf frequency to those required for a reactor, and in particular ITER. In addition, C-Mod uses heating sources that provide low torque to the plasma, as would be true for a reactor, and the time constants involved allow for relaxation of the relevant profiles. Initial experiments [2,3] have demonstrated that the system behaves as expected with bulk current drive efficiencies η = n 20 I lh R/P lh ~ 0.25 in line with theory and expectations from previous experiments. Recent work, reported in this paper, concentrates on detailed measurements of the spatial location of the driven current, the role of fast electron diffusion in spreading the current and other effects of applying LH power to the plasma. In addition, to extrapolate the application of LHCD for use in future devices a program of model development and validation is underway. The plan of this paper is as follows. The local measurements of fast electrons and CD derived from hard x-ray measurements and MSE along with comparisons to modeling will be

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Section: To 14 S (A) Notable Changes Include a Decrease In Electronsupporting

confidence: 57%

Section: Use Of Lhcd During the Current Rampmentioning

confidence: 84%

Lower Hybrid Heating and Current Drive on the Alcator C-Mod Tokamak

2009

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“…A number of authors have calculated that the stability limit can be increased by a broadening of the pressure profile (Turnbull 1996(Turnbull , 1997Phillips 1996;Bonoli 1997;Manickam 1994). The impact of the pressure profile on the ideal stability limit for a specific q profile and for a D-shape equilibria (K = 1.8, 6 = 1.7) is shown in Fig.…”

Section: Ts Taylormentioning

confidence: 99%

Physics of advanced tokamaks

Taylor¹

1997

Plasma Phys. Control. Fusion

208

153

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Significant reductions in the size and cost of a fusion power plant core can be realized if simultaneous improvements in the energy replacement time, τ E , and the plasma pressure or beta, β T = 2 µ 0 P /B 2 can be achieved in steady-state conditions with high selfdriven, bootstrap current fraction. Significant recent progress has been made in experimentally achieving these high performance regimes and in developing a theoretical understanding of the underlying physics. Three operational scenarios have demonstrated potential for steadystate high performance, the radiative improved mode, the high internal inductance or high i scenario, and the negative central magnetic shear, NCS (or reversed shear) scenario. In a large number of tokamaks, reduced ion thermal transport to near neoclassical values, and reduced particle transport have been observed in the region of negative or very low magnetic shear: the transport reduction is consistent with stabilization of microturbulence by sheared E × B flow. There is strong temporal and spatial correlation between the increased sheared E × B flow, the reduction in the measured turbulence, and the reduction in transport. The DIII-D tokamak, the JET tokamak and the JT-60U tokamak have all observed significant increases in plasma performance in the NCS operational regime. Strong plasma shaping and broad pressure profiles, provided by the H-mode edge, allow high beta operation, consistent with theoretical predictions; and normalized beta values up to β T /(I /aB) ≡ β N ∼ 4.5% m T MA −1 simultaneously with confinement enhancement over L-mode scaling, H = τ/τ ITER-89P ∼ 4, have been achieved in the DIII-D tokamak. In the JT-60U tokamak, deuterium discharges with negative central magnetic shear have reached equivalent breakeven conditions, Q DT (equiv) = 1.

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“…Although previous experimental [12][13][14][15][16][17] and theoretical 7,9,13,18,19 work has shown that ␤ limits increase with pressure profile width, much of the focus has been on equilibria with internal transport barriers and characteristically high pressure peaking. In this previous work, the dependence of ␤ limits on q min , 10,11,14 and the potential for high f BS at high q min ͑Ref.…”

Section: Introductionmentioning

confidence: 99%

Optimization of DIII-D advanced tokamak discharges with respect to the β limit

et al. 2005

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Results are presented from comparisons of modeling and experiment in studies to assess the best choices of safety factor q profile, pressure profile, and discharge shape for high β, steady-state, noninductive advanced tokamak operation in the DIII-D device [J. L. Luxon, Nucl. Fusion 42, 614 (2002)]. These studies are motivated by the need for high qminβN to maximize the self-driven bootstrap current while maintaining high toroidal β to increase fusion gain. Modeling shows that increases in the normalized beta βN stable to ideal, low toroidal mode number (n=1,2), instabilities can be obtained through broadening of the pressure profile and use of a symmetric double-null divertor shape. Experimental results are in agreement with this prediction. The general trend is for qminβN to increase with the minimum q value (qmin) although βN decreases as qmin increases. By broadening the pressure profile, βN≈4 is obtained with qmin≈2. Modeling of equilibria with near 100% bootstrap current indicates that operation with βN≈5 should be possible with a sufficiently broad pressure profile.

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Negative magnetic shear modes of operation in the Alcator C-Mod tokamak near the beta limit

Cited by 25 publications

References 21 publications

Lower Hybrid Heating and Current Drive on the Alcator C-Mod Tokamak

Lower Hybrid Heating and Current Drive on the Alcator C-Mod Tokamak

Physics of advanced tokamaks

Optimization of DIII-D advanced tokamak discharges with respect to the β limit

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