Double transport barrier plasmas in Alcator C-Mod

Rice, J. E.; Bonoli, P. T.; Marmar, E.; Wukitch, S.J.; Boivin, R. L.; Fiore, C. L.; Granetz, R.; Greenwald, M.; Hubbard, A.; Hughes, J. W.; Hutchinson, I. H.; Irby, J.; Lin, Y.; Mossessian, D.; Porkoláb, M.; Schilling, G.; Snipes, J. A.; Wolfe, S.

doi:10.1088/0029-5515/42/5/303

Cited by 96 publications

(151 citation statements)

References 33 publications

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“…This technique has been demonstrated in TEXTOR [370], Alcator C-Mod for discharges with ITBs [151] and studied in detail in ASDEX Upgrade with ICRH and ECRH [371]. The reduction of the central concentration of W can be explained by the changes caused by the central heating in the plasma particle transport [371,372].…”

Section: Behaviour Of High-z Plasma-facing Components In Tokamak Expementioning

confidence: 99%

Chapter 4: Power and particle control

Loarte¹,

et al. 2007

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Abstract.Progress, since the ITER Physics Basis publication, in understanding the processes that will determine the properties of the plasma edge and its interaction with material elements in ITER is described. Experimental areas where significant progress has taken place are : energy transport in the SOL in particular of the anomalous transport scaling, particle transport in the SOL that plays a major role in the interaction of diverted plasmas with the main chamber material elements, ELM energy deposition on material elements and the transport mechanism for the ELM energy from the main plasma to the plasma facing components, the physics of plasma detachment and neutral dynamics including the edge density profile structure and the control of plasma particle content and He removal, the erosion of low and high Z materials in fusion devices, their transport to the core plasma and their migration at the plasma edge including the formation of mixed materials, the processes determining the size and location of the retention of tritium in fusion devices and methods to remove it and the processes determining the efficiency of the various fuelling methods as well as their development towards the ITER requirements. This experimental progress has been accompanied by the development of modelling tools for the physical processes at the edge plasma and plasma-materials interaction and the further validation of these models by comparing their predictions with the new experimental results. Progress in the modelling development and validation has been mostly concentrated in the following areas : refinement of the predictions for ITER with plasma edge modelling codes by inclusion of detailed geometrical features of the divertor and the introduction of physical effects, which can play a 2 major role in determining the divertor parameters at the divertor for ITER conditions such as hydrogen radiation transport and neutral-neutral collisions, modelling of the ion orbits at the plasma edge, which can play a role in determining power deposition at the divertor target, models for plasma-materials and plasma dynamics interaction during ELMs and disruptions, models for the transport of impurities at the plasma edge to describe the core contamination by impurities and the migration of eroded materials at the edge plasma and its associated tritium retention and models for the turbulent processes that determine the anomalous transport of energy and particles across the SOL. The implications for the expected performance of the reference regimes in ITER, the operation of the ITER device and the lifetime of the plasma facing materials are discussed. Introduction.This chapter outlines the significant progress achieved since the ITER Physics Basis in understanding basic scrape-off layer (SOL) and divertor processes in a tokamak. The interaction of plasma with first-wall surfaces will have considerable impact on the performance of fusion plasmas, the lifetime of plasma facing components, and the retention of tritium in next step Burning Plasma E...

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Section: Behaviour Of High-z Plasma-facing Components In Tokamak Expementioning

confidence: 99%

Chapter 4: Power and particle control

Loarte¹,

et al. 2007

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“…It was demonstrated previously that this continued particle accumulation in the plasma core could be halted by the application of a small amount of central ICRF heating while preserving the ITB profile. 12,13 The effect of incrementally adding central ICRF power to an established ITB plasma was explored using Ohmic EDA H-mode plasmas. A reproducible target ITB was formed after setting up an EDA H-mode in an Ohmic plasma and then increasing amounts of ICRF power were added over a series of discharges.…”

Section: Control Of the Core Transport And Particle Accumulationmentioning

confidence: 99%

“…10,11,12,13 While several different criteria are often used to define an ITB, the most general definition maintains that an ITB exists in the plasma when there is a clear change in the plasma transport at an internal boundary in some quantity such as energy, particles, or momentum 14 . Steady state ITBs (lasting at least 10 energy confinement times) arise during EDA H-mode plasmas when either the H-mode is produced in an Ohmic plasma or when ICRF heating is applied well off axis, typically with the resonance location being at or greater than r/a=0.5 on either the low or high field side of the plasma.…”

Section: Introductionmentioning

confidence: 99%

“…In addition, a narrow region of decreased electron heat transport as determined from analysis of the heat pulse following a sawtooth crash is observed at or near the location of the break in the density profile 12 . ITBs in Alcator C-Mod are often accompanied by a reduction and sometimes reversal of the central rotation velocity determined by Doppler shifts of the argon impurity line radiation 10,13,15 . It should be noted that most of the discharges discussed here are sawtoothing throughout the ITB presence indicating that the q profile increases monotonically and that the magnetic shear is positive in the ITB region.…”

Section: Introductionmentioning

confidence: 99%

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Control of internal transport barriers on Alcator C-Mod

et al. 2004

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Recent studies of internal transport and double transport barrier regimes in Alcator C-Mod [I. H. Hutchinson, et al., Phys. Plasmas 1, 1511] have explored the limits for forming, maintaining, and controlling these plasmas. C-Mod provides a unique platform for studying such discharges: the ions and electrons are tightly coupled by collisions and the plasma has no internal particle or momentum sources. The double-barrier mode comprised of an edge barrier with an internal transport barrier (ITB) can be induced at will using off-axis ion cyclotron range of frequency (ICRF) injection on either the low or high field side of the plasma with either of the available ICRF frequencies (70 or 80 MHz). When enhanced D α high confinement mode (EDA H-mode) is accessed in Ohmic plasmas, the double barrier ITB forms spontaneously if the Hmode is sustained for ~2 energy confinement times. The ITBs formed in both Ohmic and ICRF heated plasmas are quite similar regardless of the trigger method. They are characterized by strong central peaking of the electron density, and reduction of the core particle and energy transport. Control of impurity influx and heating of the core plasma in the presence of the ITB have been achieved with the addition of central ICRF power in both Ohmic H-mode and ICRF induced ITBs. The radial location of the particle transport is dependent on the toroidal magnetic field but not on the location of the ICRF resonance. A narrow region of decreased electron thermal transport, as determined by sawtooth heat pulse analysis, is found in these plasmas as well. Transport analysis indicates that reduction of the particle diffusivity in the barrier region allows the neoclassical pinch to drive the density and impurity accumulation in the plasma center. Examination of the gyrokinetic stability at the trigger time for the ITB suggests that the density and temperature profiles are inherently stable to ion temperature gradient (ITG) and trapped electron (TEM) modes in the core inside of the ITB location.

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“…It is usually explained by neoclassical inward drifts, which can be dominant in the absence of large turbulent transport or macroscopic instabilities as sawteeth [24]. However, central (electron) heating strongly increases the anomalous impurity transport [24] and suppresses the accumulation as already shown in [25] and confirmed at other devices [26][27][28]. At t = 2.5 s the NBI is switched off and at the same time the ion cyclotron resonance heating (ICRH) is switched on.…”

Section: Ten Years Of W Programme In Asdex Upgrade -Challenges and Comentioning

confidence: 85%

Ten years of W programme in ASDEX Upgrade—challenges and conclusions

Neu¹,

Bobkov²,

Dux³

et al. 2009

Phys. Scr.

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Abstract. Since 1999 ASDEX Upgrade increased its tungsten plasma facing components and finally reached a full W coverage in 2007. Most of the initial goals of the investigations were successfully achieved. A highlight of the investigations were multiple start-ups and operation without any boronisation demonstrating that performance and confinement similar to boronised operation with carbon PFCs can be reached in high power, high density discharges. This also allowed the investigation of the hydrogen retention without disturbing effects from the low-Z coating. A strong reduction of hydrogen retention was found in gas balance measurements as well as in post mortem analyses. On the other hand, an almost complete suppression of low-Z divertor radiation was achieved after boronisation, providing valuable information on the control requirements of radiative cooling by artificially introduced impurities. Among the challenges remains the strong increase of the W source and W concentration resulting from ICRH. At the same time it helped to identify the underlying physics and may lead to solutions superior to the presently used ones.

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Double transport barrier plasmas in Alcator C-Mod

Cited by 96 publications

References 33 publications

Chapter 4: Power and particle control

Chapter 4: Power and particle control

Control of internal transport barriers on Alcator C-Mod

Ten years of W programme in ASDEX Upgrade—challenges and conclusions

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