Recent experimental studies of edge and internal transport barriers in the DIII-D tokamak

Gohil, P.; Baylor, L. R.; Burrell, K.H.; Casper, T. A.; Doyle, E. J.; Greenfield, C. M.; Jernigan, T. C.; Kinsey, J. E.; Lasnier, C.J.; Moyer, R. A.; Murakami, M.; Rhodes, T. L.; Rudakov, D.L.; Staebler, G. M.; Wang, G; Watkins, J. G.; West, W.P.; Zeng, L.

doi:10.1088/0741-3335/45/5/307

Cited by 29 publications

(20 citation statements)

References 53 publications

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“…The finding that auxiliary RF heating, especially central electron heating, has a flattening effect on the density profile of impurities and may reverse the convection direction has been reported in various devices, such as JET, ASDEX-U, DIII-D, etc. [2][3][4][5][6] Although the underlying physics is not well understood, this effect can be used as an approach to controlling the impurity profile, which is urgent for upcoming ITER operations.…”

Section: Introductionmentioning

confidence: 99%

Investigation of impurity transport using laser blow-off technique in the HL-2A Ohmic and ECRH plasmas

et al. 2016

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Impurity transports in two neighboring discharges with and without electron cyclotron resonance heating (ECRH) are studied in the HL-2A tokamak by laser blow-off (LBO) technique. The progression of aluminium ions as the trace impurity is monitored by soft x-ray (SXR) and bolometer detector arrays with good temporal and spatial resolutions. Obvious difference in the time trace of the signal between the Ohmic and ECRH L-mode discharges is observed. Based on the numerical simulation with one-dimensional (1D) impurity transport code STRAHL, the radial profiles of impurity diffusion coefficient D and convective velocity V are obtained for each shot. The result shows that the diffusion coefficient D significantly increases throughout the plasma minor radius for the ECRH case with respect to the Ohmic case, and that the convection velocity V changes from negative (inward) for the Ohmic case to partially positive (outward) for the ECRH case. The result on HL-2A confirms the pump out effect of ECRH on impurity profile as reported on various other devices.

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Section: Introductionmentioning

confidence: 99%

Investigation of impurity transport using laser blow-off technique in the HL-2A Ohmic and ECRH plasmas

et al. 2016

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“…One evidence that is common to various experiments is that central rf heating appears to be effective in flattening also the impurity profiles. [12][13][14][15][16][17][18][19] In particular, in the ASDEX Upgrade ͑AUG͒ experiment 20 it has been observed on the one hand the importance of the rf power deposition radius, and on the other hand that dominant electron heating via electron cyclotron resonance heating ͑ECRH͒ decontaminates the plasma core more efficiently than dominant ion heating via ion cyclotron resonance heating ͑ICRH͒. 13 In DIII D tokamak ͑Ref.…”

Section: Introductionmentioning

confidence: 99%

“…21͒, the flattening effect of ECRH on the electron density has been found to be effective also on the intrinsic nickel. 14 In the Alcator C-Mod experiment ͑Ref. 22͒, where a robust method for producing plasmas with internal transport barriers ͑ITB͒ has been by off-axis ICRF heating, with only offaxis heating the plasma density continues to peak and impurities build up, leading to a radiative collapse.…”

Section: Introductionmentioning

confidence: 99%

Analysis of metallic impurity density profiles in low collisionality Joint European Torus H-mode and L-mode plasmas

et al. 2006

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This paper describes the behavior of nickel in low confinement (L-mode) and high confinement (H-mode) Joint European Torus (JET) discharges [P. J. Lomas, Plasma Phys. Control. Fusion 31, 1481 (1989)] characterized by the application of radio-frequency (rf) power heating and featuring ITER (International Thermonuclear Experimental Reactor) relevant collisionality. The impurity transport is analyzed on the basis of perturbative experiments (laser blow off injection) and is compared with electron heat and deuterium transport. In the JET plasmas analyzed here, ion cyclotron resonance heating (ICRH) is applied either in mode conversion (MC) to heat the electrons or in minority heating (MH) to heat the ions. The two heating schemes have systematically different effects on nickel transport, yielding flat or slightly hollow nickel density profiles in the case of ICRH in MC and peaked nickel density profiles in the case of rf applied in MH. Accordingly, both diffusion coefficients and pinch velocities of nickel are found to be systematically different. Linear gyrokinetic calculations by means of the code GS2 [M. Kotschenreuther, G. Rewoldt, and W.M. Tang, Comput. Phys. Commun. 88, 128 (1995)] provide a possible explanation of such different behavior by exploring the effects produced by the different microinstabilities present in these plasmas. In particular, trapped electron modes driven by the stronger electron temperature gradients measured in the MC cases, although subdominant, produce a contribution to the impurity pinch directed outwards that is qualitatively in agreement with the pinch reversal found in the experiment. Particle and heat diffusivities appear to be decoupled in MH shots, with χe and DD≫DNi, and are instead quite similar in the MC ones. In the latter case, nickel transport appears to be driven by the same turbulence that drives the electron heat transport and is sensitive to the value of the electron temperature gradient length. These findings give ground to the idea that in ITER it should be possible to find conditions in which the risk of accumulation of metals such as nickel can be contained.

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“…The leading candidates for ITB triggering and also governing the physics of the ITBs are the turbulence suppression by the ω E×B shearing rate [9][10][11][12][13][14][15][16], by negative or small magnetic shear [14,[17][18][19], by αstabilization [19,[22][23][24], by a large density gradient [25,26] and by the rational values of the q-profile [27,28]. In this work, we have used three different transport models, the mixed Bohm/GyroBohm [14,29], the Weiland [30,31] and the 'retuned' GLF23 [4,11,32], to carry out all the transport simulations.…”

Section: Introductionmentioning

confidence: 99%

Fully predictive time-dependent transport simulations of ITB plasmas in JET, JT-60U and DIII-D

Tala¹,

Imbeaux²,

Parail³

et al. 2006

Nucl. Fusion

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For the first time, the predictive capabilities of the mixed Bohm/GyroBohm, Weiland and ‘retuned’ GLF23 transport models are investigated with ITB discharges from the ITPA ITB database with fully predictive, time-dependent transport simulations. A range of plasma conditions is examined for JET, JT-60U and DIII-D discharges with internal transport barriers (ITBs). The simulations show that the Bohm/GyroBohm model is able to follow the time evolution of the discharge from the preheating phase without an ITB through the ITB onset phase until the high performance phase with fair accuracy in most cases in JET and JT-60U. This indicates the importance of the interplay between the magnetic shear and ωE×B flow shear in ITB formation since these are the mechanisms that govern the ITB physics in the model. In order to achieve good agreement in DIII-D, the α-stabilization had to be included in the model, emphasizing the role played by the α-stabilization in the physics of the ITBs. The Weiland and GLF23 transport models show limited agreement between the model predictions and experimental time evolution of the ITB and kinetic plasma profiles. The Weiland model does not form a clear ITB in any of the three tokamaks despite varying plasma profiles, such as the q-profile. On the other hand, the average temperatures and density are often in fair agreement with experimental values. The GLF23 model often predicts an ITB, but its radial location is often too far inside the plasma, or shrinks as the simulations proceed in time. Consequently, the central temperatures at the end of the simulations during the high performance phase are usually underestimated. It is worth noting that GLF23 features in general better predictions of the Te and Ti profiles outside the ITB than the other models studied. Achieving the quantitative capability to predict the multi-channel ITB dynamics with fully predictive, time-dependent transport simulations has turned out to be extremely challenging.

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Recent experimental studies of edge and internal transport barriers in the DIII-D tokamak

Cited by 29 publications

References 53 publications

Investigation of impurity transport using laser blow-off technique in the HL-2A Ohmic and ECRH plasmas

Investigation of impurity transport using laser blow-off technique in the HL-2A Ohmic and ECRH plasmas

Analysis of metallic impurity density profiles in low collisionality Joint European Torus H-mode and L-mode plasmas

Fully predictive time-dependent transport simulations of ITB plasmas in JET, JT-60U and DIII-D

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