G.Z. Zuo

Comparison of various wall conditionings on the reduction of H content and particle recycling in EAST

Reductions in H content and particle recycling are important for the improvement of ion cyclotron range of frequency (ICRF) minority heating efficiency and the enhancement of plasma performance of the EAST superconducting tokamak. During recent years several techniques of surface conditioning such as baking, glow discharge cleaning/ICRF discharge cleaning, surface coatings, such as boronization, siliconization and lithium coating, have all been attempted in order to reduce the H/(H+D) ratio and particle recycling in EAST. Even though boronization and siliconization were both reasonably effective methods to improve plasma performance, lithium coatings were observed to reduce the H content and particle recycling to levels low enough to allow the attainment of enhanced plasma parameters and operating modes on EAST. For example, by accomplishing lithium coating using either vacuum evaporation or the real-time injection of fine lithium powder, the H/(H+D) ratio could be routinely decreased to about 5%, which significantly improved ICRF minority heating efficiency during the autumn campaign of 2010. Due to the reduced H/(H+D) ratio and lower particle recycling, and a reduced H-mode power threshold, improved plasma confinement and the first EAST H-mode plasma were obtained. Furthermore, with increasing accumulation of deposited lithium, several new milestones of EAST performance, such as a 6.4 s-long H-mode, a 100 s-long plasma duration and a 1 MA plasma current, were achieved in the 2010 autumn campaign.

First observations of ELM triggering by injected lithium granules in EAST

The first results of edge-localized mode (ELM) pacing using small spherical lithium granules injected mechanically into H-mode discharges are reported. Triggering of ELMs was accomplished using a simple rotating impeller to inject sub-millimetre size granules at speeds of a few tens of meters per second into the outer midplane of the EAST fusion device. During the injection phase, ELMs were triggered with near 100% efficiency and the amplitude of the induced ELMs as measured by Dα was clearly reduced compared to contemporaneous naturally occurring ELMs. In addition, a wide range of granule penetration depths was observed. Moreover, a substantial fraction of the injected granules appeared to penetrate up to 50% deeper than the 3 cm nominal EAST H-mode pedestal width. The observed granule penetration was, however, less deep than suggested by ablation modelling carried out after the experiment. The observation that ELMs can be triggered using the injection of something other than frozen hydrogenic pellets allows for the contemplation of lithium or beryllium-based ELM pace-making on future fusion devices. This change in triggering paradigm would allow for the decoupling of the ELM-triggering process from the plasma-fuelling process which is currently a limitation on the performance of hydrogen-based ELM mitigation by injected pellets.

First results of the use of a continuously flowing lithium limiter in high performance discharges in the EAST device

As an alternative choice of solid plasma facing components (PFCs), flowing liquid lithium can serve as a limiter or divertor PFC and offers a self-healing surface with acceptable heat removal and good impurity control. Such a system could improve plasma performance, and therefore be attractive for future fusion devices. Recently, a continuously flowing liquid lithium (FLiLi) limiter has been successfully designed and tested in the EAST superconducting tokamak. A circulating lithium layer with a thickness of <0.1 mm and a flow rate ~2 cm3 s−1 was achieved. A novel in-vessel electro-magnetic pump, working with the toroidal magnetic field of the EAST device, was reliable to control the lithium flow speed. The flowing liquid limiter was found to be fully compatible with various plasma scenarios, including high confinement mode plasmas heated by lower hybrid waves or by neutral beam injection. It was also found that the controllable lithium emission from the limiter was beneficial for the reduction of recycling and impurities, for the reduction of divertor heat flux, and in certain cases, for the improvement of plasma stored energy, which bodes well application for the use of flowing liquid lithium PFCs in future fusion devices.

First results of flowing liquid lithium limiter in HT-7

Two different types of flowing liquid lithium limiters were firstly installed and successfully tested in HT-7 tokamaks in 2012 and some encouraging results were obtained. Two limiters of the first type, called FLiLi limiters, used a thin lithium layer flowing under gravity. The other type had lithium-metal infused trenches (LIMIT) for thermoelectric magnetohydrodynamic drive of the liquid metal flow. The surface of one of the FLiLi limiters was coated by evaporated lithium before liquid lithium was injected by Ar pressure into a special distributor of the limiter. Then the liquid lithium could slowly move along the plasma facing guide surface of the limiter due to gravity. For LIMIT, it was found that liquid lithium could flow along the trenches as expected with a velocity of about 3.7 ± 0.5 cm s−1 driven by the electromagnetic force, which came from the interaction between the thermoelectric current and magnetic field. Use of flowing liquid lithium limiters in HT-7 resulted in reduction of particle recycling, suppression of impurity emission and improvement of the confinement.

Primary Results of Lithium Coating for the Improvement of Plasma Performance in EAST

First lithium coating associated with ion cyclotron range of frequency (ICRF) plasma was performed successfully in EAST. Results in reduction of both residual impurity and deuterium in the vacuum vessel were obtained. Particularly the partial pressure of deuterium after the lithium coating was reduced by about a factor of 5. Impurity radiation in the plasma was reduced and electron temperature increased by about 50%. Moreover, reproducible plasma discharges with high parameters, such as higher plasma current and density, could be easily obtained. These results showed that plasma performance was improved. Even though only 2 g of lithium were injected, the effective lifetime of the Li film was raised up to 40 shots.

Measurement and analysis of Zeff in EAST tokamak

Effective ion charge (Zeff) is an important parameter in fusion research and can be used to quantify impurity levels. Investigation into Zeff helps us to understand the impurity generation and transport mechanisms in order to find an effective way to control impurity levels. This key parameter Zeff is measured by the visible bremsstrahlung diagnostic in the EAST tokamak. In this report, Zeff and impurity behavior in siliconization and lithium coating experiments have been investigated. Comparisons of averaged Zeff and Zeff profiles have been made between L-mode and H-mode discharges. It is found that the dependence of Zeff on averaged electron density complies with the hyperbolic scaling. Finally, Zeff behavior in auxiliary heating experiments is analyzed in detail. The relationship between Zeff and external power injected into plasma in lower hybrid wave and ion cyclotron resonance frequency heating has been summarized and analyzed.

Density limits investigation and high density operation in EAST tokamak

Increasing the density in a tokamak is limited by the so-called density limit, which is generally performed as an appearance of disruption causing loss of plasma confinement, or a degradation of high confinement mode which could further lead to a H → L transition. The L-mode and H-mode density limit has been investigated in EAST tokamak. Experimental results suggest that density limits could be triggered by either edge cooling or excessive central radiation. The L-mode density limit disruption is generally triggered by edge cooling, which leads to the current profile shrinkage and then destabilizes a 2/1 tearing mode, ultimately resulting in a disruption. The L-mode density limit scaling agrees well with the Greenwald limit in EAST. The observed H-mode density limit in EAST is an operational-space limit with a value of . High density H-mode heated by neutral beam injection (NBI) and lower hybrid current drive (LHCD) are analyzed, respectively. The constancy of the edge density gradients in H-mode indicates a critical limit caused perhaps by e.g. ballooning induced transport. The maximum density is accessed at the H → L transition which is generally caused by the excessive core radiation due to high Z impurities (Fe, Cu). Operating at a high density () is favorable for suppressing the beam shine through NBI. High density H-mode up to could be sustained by 2 MW 4.6 GHz LHCD alone, and its current drive efficiency is studied. Statistics show that good control of impurities and recycling facilitate high density operation. With careful control of these factors, high density up to 0.93 stable H-mode operation was carried out heated by 1.7 MW LHCD and 1.9 MW ion cyclotron resonance heating with supersonic molecular beam injection fueling.

Erratum: New Steady-State Quiescent High-Confinement Plasma in an Experimental Advanced Superconducting Tokamak [Phys. Rev. Lett. 114, 055001 (2015)].

In our recently published work, there is an error in the stated direction of the edge coherent MHDmode (ECM) associated with real-time Li injection. This error does not affect the main results of the paper. The direction of the ECMwas obtained from two high sampling-frequencyMirnov magnetic pickup probes. In the Letter, we reported that the ECM propagates in the ion diamagnetic drift direction. The ECM actually propagates in the electron diamagnetic drift direction, consistent with the identified direction in our previous study [1].

First Results of ELM Triggering With a Multichamber Lithium Granule Injector Into EAST Discharges

A critical challenge facing the basic long-pulse H-mode for ITER is to control edge-localized modes (ELMs). A new method using a multichamber lithium (Li) granule injector (LGI) for ELM triggering experiments has been developed in Experimental Advanced Superconducting Tokamak (EAST). First experimental results of the control of ELMs are obtained in EAST with a tungsten divertor. It is found that the injector has good capacities, i.e., allowing good flexibilities in granule size selection, injection rate, and injection velocity. LGI has successfully triggered ELMs during the H-mode. These results indicate the LGI would be a promising method to control ELMs in long-pulse steady-state tokamaks.

Compatibility of Molybdenum, Tungsten, and 304 Stainless Steel in Static Liquid Lithium Under High Vacuum

Molybdenum (Mo), tungsten (W), and stainless steel (SS) are widely used as important structure materials and first wall materials in fusion devices, while liquid lithium (Li) limiter/divertor can provide an attractive option for withstanding high heat load and solving life-time problem of first wall. Studying the compatibility of these materials exposed to liquid Li is significant for the application in Mo, W, and SS in fusion reactors. The corrosion behaviors of Mo, W, and 304SS exposed to static liquid Li at 600 K up to 1320 h under high vacuum with pressure 10−4 Pa were investigated. After exposure to liquid Li, it was found that the weight loss of Mo, W, and 304SS increases with corrosion time, but the total amount is moderate. 304SS specimens produce a non-uniform corrosion behavior because of Cr, Ni, and carbon (C) elements selectivity depletion and formation of carbides compound near surface. Mo and W surface microstructures are unchanged. 304SS surface hardness increases with corrosion products because these particles include C element, which increases by 49 HV after exposed to liquid Li for 1320 h, while Mo and W surface hardness are unchanged by the reason of their excellent corrosion resistance.