Transient Transport Experiments in the CDX-U Spherical Torus

Munsat, T.; Efthimion, P. C.; Jones, B.; Kaita, R.; Majeski, R.; Stutman, D.; Taylor, Chris

doi:10.2172/786537

Cited by 3 publications

(4 citation statements)

References 32 publications

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“…3 is consistent with diffusive thermal transport near the core. The thermal diffusivity is apparently much higher in the outer regions of the discharge [15]. Figure 3 The delay in the peak of the EBW T rad pulse is maximum at the magnetic axis following a cold pulse perturbation, as expected.…”

Section: Last Closedsupporting

confidence: 69%

Electron Bernstein wave research on CDX-U and NSTX

Taylor¹,

Efthimion²,

Hosea³

et al. 2001

AIP Conference Proceedings

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Mode-converted electron Bernstein waves (EBWs) potentially allow the measurement of local electron temperature (T e ) and the implementation of local heating and current drive in spherical torus (ST) devices, which are not directly accessible to low harmonic electron cyclotron waves. This paper reports on the measurement of X-mode radiation mode-converted from EBWs observed normal to the magnetic field on the midplane of the CDX-U and NSTX ST plasmas. The radiation temperature of the EBW emission was compared to T e measured by Thomson scattering and Langmuir probes. EBW mode conversion efficiencies of over 20% were measured on both CDX-U and NSTX. Sudden increases of mode conversion efficiency, of over a factor of three, were observed at H-mode transitions on NSTX, when the measured edge density profile steepened. The EBW mode conversion efficiency was found to depend on the density gradient at the mode conversion layer in the plasma scrape off, consistent with theoretical predictions. The EBW emission source was determined by a perturbation technique to be localized at the electron cyclotron resonance layer and was successfully used for radial transport studies. Recently, a new in-vessel antenna and Langmuir probe array were installed on CDX-U to better characterize and enhance the EBW mode conversion process. The probe incorporates a local adjustable limiter to control and maximize the mode conversion efficiency in front of the antenna by modifying the density profile in the plasma scrape off where fundamental EBW mode conversion occurs. Initial results show that the mode conversion efficiency can be increased to ~ 100% when the local limiter is inserted near the mode conversion layer. Plans for future EBW research, including EBW heating and current drive studies, are discussed.

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Section: Last Closedsupporting

confidence: 69%

Electron Bernstein wave research on CDX-U and NSTX

Taylor¹,

Efthimion²,

Hosea³

et al. 2001

AIP Conference Proceedings

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“…21 It is informative to compare the lithium PFC results on CDX-U with the predictions from ITER98P͑y,1͒ scaling. These are discharges with a liquid lithium tray limiter and/or lithium deposited on PFCs immediately preceding the shot.…”

Section: Plasma Performance With Lithium Pfcsmentioning

confidence: 99%

Low recycling and high power density handling physics in the Current Drive Experiment-Upgrade with lithium plasma-facing components

et al. 2007

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The Current Drive Experiment-Upgrade [T. Munsat, P. C. Efthimion, B. Jones, R. Kaita, R. Majeski, D. Stutman, and G. Taylor, Phys. Plasmas 9, 480 (2002)] spherical tokamak research program has focused on lithium as a large area plasma-facing component (PFC). The energy confinement times showed a sixfold or more improvement over discharges without lithium PFCs. This was an increase of up to a factor of 3 over ITER98P(y,1) scaling [ITER Physics Basis Editors, Nucl. Fusion 39, 2137 (1999)], and reflects the largest enhancement in confinement ever seen in Ohmic plasmas. Recycling coefficients of 0.3 or below were achieved, and they are the lowest to date in magnetically confined plasmas. The effectiveness of liquid lithium in redistributing heat loads at extremely high power densities was demonstrated with an electron beam, which was used to generate lithium coatings. When directed to a lithium reservoir, evaporation occurred only after the entire volume of lithium was raised to the evaporation temperature. The ability to dissipate a beam power density of about 60MW∕m2 could have significant consequences for PFCs in burning plasma devices.

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“…The peak electron temperature can be estimated to be ∼300 eV from the Spitzer resistivity, for moderately broadened electron temperature profiles, with a Z effective of 1. This compares with a measured value of 100-150 eV, from Thomson scattering, on earlier high recycling discharges [14]. Gas puffing was always terminated several milliseconds before the energy confinement time was analysed, so that the edge neutral density is dominated by recycling (still estimated at 50% in these discharges).…”

Section: Experimental Results From Cdx-u Lithium Operationmentioning

confidence: 48%

“…We note that results from a similar sized diverted spherical tokamak which employed high recycling carbon divertor targets, START, also typically exceeded neo-Alcator scaling, but by a factor not exceeding 2-4 [9]. Comparisons of CDX-U with simulation codes such as TSC [10], which was extensively used to model CDX-U as part of a benchmarking effort [11] for M3D [12] and NIMROD [13], have been successful at reproducing the sub-millisecond energy confinement times typically seen in high recycling discharges [14], but underpredict low recycling confinement by an order of magnitude or more. The failure of existing scaling laws and numerical codes to predict the confinement time in very low recycling discharges implies that existing approaches are inadequate to predict the performance for LTX.…”

Section: Experimental Results From Cdx-u Lithium Operationmentioning

confidence: 99%

Performance Projections For The Lithium Tokamak Experiment (LTX)

Majeski¹,

Berzak²,

Gray³

et al. 2009

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Use of a large-area liquid lithium limiter in the CDX-U tokamak produced the largest relative increase (an enhancement factor of 5-10) in Ohmic tokamak confinement ever observed. The confinement results from CDX-U do not agree with existing scaling laws, and cannot easily be projected to the new lithium tokamak experiment (LTX). Numerical simulations of CDX-U low recycling discharges have now been performed with the ASTRA-ESC code with a special reference transport model suitable for a diffusion-based confinement regime, incorporating boundary conditions for nonrecycling walls, with fuelling via edge gas puffing. This model has been successful at reproducing the experimental values of the energy confinement (4-6 ms), loop voltage (<0.5 V), and density for a typical CDX-U lithium discharge. The same transport model has also been used to project the performance of the LTX, in Ohmic operation, or with modest neutral beam injection (NBI). NBI in LTX, with a low recycling wall of liquid lithium, is predicted to result in core electron and ion temperatures of 1-2 keV, and energy confinement times in excess of 50 ms. Finally, the unique design features of LTX are summarized.

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Transient Transport Experiments in the CDX-U Spherical Torus

Cited by 3 publications

References 32 publications

Electron Bernstein wave research on CDX-U and NSTX

Electron Bernstein wave research on CDX-U and NSTX

Low recycling and high power density handling physics in the Current Drive Experiment-Upgrade with lithium plasma-facing components

Performance Projections For The Lithium Tokamak Experiment (LTX)

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