A new electromagnetic probe array diagnostic for analyzing electrostatic and magnetic fluctuations in EAST plasmas

A novel electromagnetic probe array(EMPA) diagnostic, which consists of a magnetic probe array and an electrostatic probe array, has recently been developed on EAST. The EMPA is fixed near the first wall at horizontal port P. The magnetic probe array of the EMPA consists of 24 identical magnetic probes, each of them capable of measuring toroidal, poloidal and radial magnetic fluctuations simultaneously, providing additional toroidal magnetic fluctuation measurements compared with the regular magnetic probes on EAST. With a higher sampling rate and self-resonant frequency, the EMPA magnetic probes can provide higher frequency magnetic fluctuation measurements. The magnetic probe array of the EMPA is composed of two parallel layers of magnetic probes with a radial distance of 63 mm, and each layer of magnetic probes is arranged in four poloidal rows and three toroidal columns. The compact arrangement of the EMPA magnetic probe array largely improves the toroidal mode number measurement ability from-8≤ n≤ 8 to-112≤ n≤ 112, and also improves the high poloidal wave number measurement ability of magnetic fluctuations compared with the regular high frequency magnetic probes on EAST. The electrostatic probe array of the EMPA consists of two sets of four-tip probes and a single-tip probe array with three poloidal rows and four toroidal columns. It complements the electrostatic parameter measurements behind the main limiter and near the first wall in EAST. The engineering details of the EMPA diagnostic, including the mechanical system, the electrical system, the acquisition and control system, and the effective area calibration, are presented. The preliminary applications of the EMPA in L-mode and H-mode discharges on EAST have demonstrated that the EMPA works well for providing information on the magnetic and electrostatic fluctuations and can contribute to deeper physical analysis in future EAST experiments.

Design and platform testing of the compact torus central fueling device for the EAST tokamak

Compact torus(CT)injection is a highly promising technique for the central fueling of future reactor-grade fusion devices since it features extremely high injection velocity and relatively high plasma mass.Recently,a CT injector for the EAST tokamak,EAST-CTI,was developed and platform-tested.In the first round of experiments conducted with low parameter settings,the maximum velocity and mass of the CT plasma were 150 km·s^(-1)and 90μg,respectively.However,the parameters obtained by EAST-CTI were still very low and were far from the requirements of a device such as EAST that has a strong magnetic field.In future,we plan to solve the spark problem that EAST-CTI currently encounters(that mainly hinders the further development of experiments)through engineering methods,and use greater power to obtain a more stable and suitable CT plasma for EAST.

Investigation of the compact torus plasma motion in the KTX-CTI device based on circuit analyses

Compact torus(CT)injection is one of the most promising methods for the central fuelling of next-generation reactor-grade fusion devices due to its high density,high velocity,and selfcontained magnetised structure.A newly compact torus injector(CTI)device in Keda Torus e Xperiment(KTX),named KTX-CTI,was successfully developed and tested at the University of Science and Technology in China.In this study,first,we briefly introduce the basic principles and structure of KTX-CTI,and then,present an accurate circuit model that relies on nonlinear regression analysis(NRA)for studying the current waveform of the formation region.The current waveform,displacement,and velocity of CT plasma in the acceleration region are calculated using this NRA-based one-dimensional point model.The model results were in good agreement with the experiments.The next-step upgrading reference scheme of the KTX-CTI device is preliminarily investigated using this NRA-based point model.This research can provide insights for the development of experiments and future upgrades of the device.

Conceptual design and heat transfer performance of a flat-tile water-cooled divertor target

The divertor target components for the Chinese fusion engineering test reactor(CFETR)and the future experimental advanced superconducting tokamak(EAST)need to remove a heat flux of up to20 MW m-2.In view of such a high heat flux removal requirement,this study proposes a conceptual design for a flat-tile divertor target based on explosive welding and brazing technology.Rectangular water-cooled channels with a special thermal transfer structure(TTS)are designed in the heat sink to improve the flat-tile divertor target’s heat transfer performance(HTP).The parametric design and optimization methods are applied to study the influence of the TTS variation parameters,including height(H),width(W*),thickness(T),and spacing(L),on the HTP.The research results show that the flat-tile divertor target’s HTP is sensitive to the TTS parameter changes,and the sensitivity is T>L>W*>H.The HTP first increases and then decreases with the increase of T,L,and W*and gradually increases with the increase of H.The optimal design parameters are as follows:H=5.5 mm,W*=25.8 mm,T=2.2 mm,and L=9.7 mm.The HTP of the optimized flat-tile divertor target at different flow speeds and tungsten tile thicknesses is studied using the numerical simulation method.A flat-tile divertor mock-up is developed according to the optimized parameters.In addition,high heat flux(HHF)tests are performed on an electron beam facility to further investigate the mock-up HTP.The numerical simulation calculation results show that the optimized flat-tile divertor target has great potential for handling the steady-state heat load of 20 MW m-2under the tungsten tile thickness<5 mm and the flow speed7 m s^(-1).The heat transfer efficiency of the flat-tile divertor target with rectangular cooling channels improves by13%and30%compared to that of the flat-tile divertor target with circular cooling channels and the ITER-like monoblock,respectively.The HHF tests indicate that the flat-tile divertor mock-up can successfully withstand 1000 cycles of20 MW m-2of heat load without visible deformation,damage,and HTP degradation.The surface temperature of the flat-tile divertor mock-up at the 1000th cycle is only930℃.The flat-tile divertor target’s HTP is greatly improved by the parametric design and optimization method,and is better than the ITER-like monoblock and the flat-tile mock-up for the WEST divertor.This conceptual design is currently being applied to the engineering design of the CFETR and EAST flat-tile divertors.