Effects of heating power on divertor in-out asymmetry and scrape-off layer flow in reversed field on Experimental Advanced Superconducting Tokamak

Liu, S. C.; Guo, H.Y.; Wang, Li; Wang, H. Q.; Gan, Kaifu; Xia, Tie; Xu, G. S.; Xu, X. Q.; Liu, Z. X.; Chen, L.; Niu, Y.; Zhang, W.; Chen, R.; Shao, L. M.; Ding, Siye; Hu, G.H.; Liu, Y. L.; Zhao, Nan; Li, Y. L.; Gong, Xianzu; Gao, Xiang

doi:10.1063/1.4904205

Cited by 5 publications

(4 citation statements)

References 59 publications

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“…Similar results for the SOL flow reversal have been observed in the JT-60U tokamak by Mach probe measurements for two ion ∇B drift directions, suggesting that the ion Pfirsh-Schlüter (PS) flow is the most feasible mechanism to drive the parallel plasma flow at the midplane against the ion ∇B drift direction [37]. On EAST, the characteristics of measured SOL parallel plasma flow under various discharge conditions on the low field side midplane are consistent with the PS flow, with measured parallel Mach number M || = 0.3 − 0.5, as shown in figures 12 and 14 in [38,39]. These findings suggest that the SOL parallel flow on the low field side midplane is dominated by the PS flow and that the direction of the PS flow component of the SOL parallel flow depends on the ion ∇B drift direction.…”

Section: Role Of Field-dependent Sol Parallel Flow On the Transition supporting

confidence: 59%

Study on the L–H transition power threshold with RF heating and lithium-wall coating on EAST

Chen

Nielsen

et al. 2016

Nucl. Fusion

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The power threshold for low (L) to high (H) confinement mode transition achieved by radio-frequency (RF) heating and lithium-wall coating is investigated experimentally on EAST for two sets of walls: an all carbon wall (C) and molybdenum chamber and a carbon divertor (Mo/C). For both sets of walls, a minimum power threshold P thr of ~0.6 MW was found when the EAST operates in a double null (DN) divertor configuration with intensive lithium-wall coating. When operating in upper single null (USN) or lower single null (LSN), the power threshold depends on the ion ∇B drift direction. The low density dependence of the L-H power threshold, namely an increase below a minimum density, was identified in the Mo/C wall for the first time. For the C wall only the single-step L-H transition with limited injection power is observed whereas also the so-called dithering L-H transition is observed in the Mo/C wall. The dithering behaves distinctively in a USN, DN and LSN configuration, suggesting the divertor pumping capability is an important ingredient in this transition since the internal cryopump is located underneath the lower divertor. Depending on the chosen divertor configuration, the power across the separatrix P loss increases with neutral density near the lower X-point in EAST with the Mo/C wall, consistent with previous results in the C wall (Xu et al 2011 Nucl. Fusion 51 072001). These findings suggest that the edge neutral density, the ion ∇B drift as well as the divertor pumping capability play important roles in the L-H power threshold and transition behaviour.

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Section: Role Of Field-dependent Sol Parallel Flow On the Transition supporting

confidence: 59%

Study on the L–H transition power threshold with RF heating and lithium-wall coating on EAST

Chen

Nielsen

et al. 2016

Nucl. Fusion

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“…The errors of electron temperature T e measured by divertor triple probes are estimated to be roughly ±50%, based on the comparison with divertor single probe measurements in previous EAST experiments [30,31]. According to the error transfer formula, the errors of electron density n e and electron pressure p e are about ±35% [30,31]. The particle flux in equation ( 16)…”

Section: Modeling Without Driftsmentioning

confidence: 98%

SOLPS-ITER modeling of SOL-divertor plasmas with different configurations in EAST

et al. 2020

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SOLPS-ITER is used for modeling divertor plasmas in the experimental shot on the EAST tokamak with a carbon divertor. The modeling uses real divertor geometry and an MHD equilibrium from the shot. The modelled divertor configurations include upper single null (USN), disconnected double null (DDN) and lower single null (LSN). D 0 , D +1 , e − and carbon impurity species sputtered from the first wall are included in the multi-fluid simulations. Neutral particles are traced with EIRENE in which neutral-neutral collisions in triangular grids are considered. The default physics models in the code are employed. The cross-field transport coefficients D, χ e and χ i deduced from the profile measurements at the targets are used with the assumption of χ i = χ e , some of the deduced D are modified in the SOL and private regions and D, χ e , χ i in the core are adjusted in order to match better the experimental measurement at the target plates. Employed D, χ e and χ i drop from the core to the SOL, they may drop to neoclassic level within the edge transport barriers (ETBs). A ballooning effect on the transport coefficients is introduced. The computational contour profiles of electron temperature in the divertor regions and the computational profiles of plasma pressure and electron temperature at the mid-plane (the upstream) and at the target plate show the sheath limited (low recycling) operation regime. The calculated profiles of electron density n e , electron temperature T e , particle flux f , parallel ion saturation current j sat and parallel power flux P along the target plates of the divertor are compared to the experimental measurements. Using the deduced and adjusted cross-field transport coefficients, when the measurement errors are considered the calculated and measured profiles have a relatively good match, the computational profiles are similar to the measured profiles, the profiles peaked at the separatrix strike points. The most peak values f s of calculated profiles are in the range of the measured error-bars f e ± d, only in several matched profiles f s is out of the range of the errorbar, but, relative error error p does not exceed 46.55%. When full drifts are taken into account in the entire SOL and the private flux regions the calculated profiles have better match to the measured profiles and the modeling results show the drifts enhance the divertor asymmetry.

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“…As aforementioned, the experimental studies of the DOIA of heat flux for L-mode discharges 24 – 26 and the DOIA of particle flux for H-mode discharges 27 , 28 have been carried out in EAST in recent years. There is still a lack of study on the DOIA of heat flux for EAST H-mode discharges due to the unreliability of heat flux measurement from the divertor Langmuir probes.…”

Section: Effects Of Drifts On the Doias Of Particle And Heat Fluxes A...mentioning

confidence: 99%

“…EAST is the first superconducting tokamak in China with ITER-like divertor and configuration. Experimental studies on the DOIA of heat flux for the L-mode discharges 24 – 26 and DOIA of particle flux for the H-mode discharges 27 , 28 have been carried out recently. The results also show that the DOIAs could be reversed by reversing the toroidal magnetic field.…”

Section: Introductionmentioning

confidence: 99%

Modeling study of divertor particle and heat flux asymmetries for EAST H-mode discharges

Deng

Lin

2022

Sci Rep

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The BOUT++ transport code is run to study the effects of plasma drifts on the divertor out-in asymmetries (DOIAs) of particle and heat fluxes and their decay widths for EAST lower single null H-mode discharges. The diamagnetic drift seems to have no effects on the DOIAs of total particle and heat fluxes due to its divergence-free nature. However, it could significantly increase the DOIAs of peak particle and heat fluxes and the flux decay widths. The E × B drift is found to induce a large plasma flow to the divertor region, enhancing the DOIAs of both total and peak particle and heat fluxes and the flux decay widths. Both the radial and poloidal components of the E × B drift are necessary in increasing the DOIAs, however, the radial E × B drift seems to play a more important role. The effects on the DOIAs caused by both diamagnetic and E × B drifts are reversed with the reverse of toroidal magnetic field. The heat flux decay width λq and spreading width Sq are important physical and engineering parameters for the divertors and could be obtained by fitting the heat flux profiles at divertor targets. The λq at the outer target from the simulation case with all drifts could well match with the multi-machine scaling proposed by Eich and the DOIA of λq is in reasonable agreement with the scaling proposed by Goldston.

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Effects of heating power on divertor in-out asymmetry and scrape-off layer flow in reversed field on Experimental Advanced Superconducting Tokamak

Cited by 5 publications

References 59 publications

Study on the L–H transition power threshold with RF heating and lithium-wall coating on EAST

Study on the L–H transition power threshold with RF heating and lithium-wall coating on EAST

SOLPS-ITER modeling of SOL-divertor plasmas with different configurations in EAST

Modeling study of divertor particle and heat flux asymmetries for EAST H-mode discharges

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