Modelling of carbon erosion and re-deposition for the EAST movable limiter

The movable limiter at the mid-plane of the Experimental Advanced Superconducting Tokamak(EAST)with carbon coatings on the surface was exposed to edge plasma to study the material erosion and re-deposition.After the experiments,the carbon erosion and re-deposition is modelled using the 3D Monte Carlo code ERO.The geometry of the movable limiter,3D configuration of the plasma parameters and electromagnetic fields under both limiter and divertor configurations have been implemented into the code.In the simulations,the main uncertain parameters such as carbon concentrationρ_c in the background plasma and cross-field transport coefficient D_⊥in the vicinity of surface according to the'funneling model',have been studied in comparison with experiments.The parameterρ_c mainly influences the net erosion and deposition profiles of the two sides of the movable limiter,while D_⊥mostly changes the profiles on the top surface.     

Preliminary design and performance study of EAST liquid lithium limiter based on CPS

Lithium has the ability of H recycling suppression and impurities absorption and it can be used as plasma facing material (PFM) in tokamaks. Lithium conditioning experiments were launched on EAST, HT-7 and some other tokamaks for many years by using the methods of GDC, IRCF and evaporation. Liquid lithium has better performances in effective lifetime and heat removal aspects compared to non-liquid lithium. While, applying liquid lithium in the tokamak would cause the safety problem as the lithium can react with many substances violently and the magnetohydrodynamic behavior is difficult to be handled. EAST liquid lithium limiter (LLL) system is under developing and will be applied in EAST to study the main technologies of the liquid lithium application. The normal operation temperature of the limiter is expected as 230–550 °C under the active cooling of water. Capillary porous system (CPS) is used to prevent the lithium from splashing under large electromagnetic force by increasing the surface tension of the lithium. In order to investigate the cooling performance of the cooling design, the thermal-hydraulic analysis was done which shows that with 3 m/s flowing velocity, the water can keep the limiter under 550 °C all the time if the heat flux is lower than 0.7 MW/m². Under heat flux of 1 MW/m², the limiter should be retreated within 7 s to avoid erosion. The pressure drop of the coolant under 3 m/s is less than 40 kPa with temperature difference nearly 34 °C which meet the design requirements very well. The key manufacture process and technologies like vacuum bonding between the CuCrZr heat sink and 316L guide plate were well studied in the R&D process. The heating test on the test bench showed that the CPS can be heated efficiently by the heaters attached into the heat sink.     

Modeling of heat deposition on the W/Cu movable limiter in HT-7

The thermal behavior of a directly water-cooled W/Cu movable poloidal limiter was investigated in HT-7, a medium-sized superconducting tokamak with limiter configuration, major radius R = 1.22 m, and minor radius a = 0.27 m. The W/Cu movable limiter (ML) was exposed to the plasma at various radial positions at r < a. The surface and bulk temperatures were monitored by an IR-camera and the thermocouples, respectively. The heat flux deposited on the limiter was evaluated by an ANSYS code using the measured surface temperatures as boundary conditions. It was found that the maximum heat flux incident on the ML was less than 1 MW/m² in the Ohmic discharges, but reached up to 5-7 MW/m² in the discharges with lower hybrid current drive (LHCD). A simple model was developed to understand heat transport to the W/Cu ML, taking into account the “funnel effect”.     

Primary research of heat flux deposition pattern on the surface of HT-7 movable limiter

This paper presents a method of characterizing the heat flux deposition pattern on HT-7 movable limiter, a new component in the spring experimental campaign of year 2006. A new modeling establishment combines with heat transfer simulation with ANSYS code, and a shot of long pulse discharge is chosen with small radius 265 mm of movable limiter, which is 5 mm smaller than that of the main toroidal limiters. Both parallel and perpendicular heat flux are taken into account and different ratios of them are also tested in simulations. The simulation temperature values match well with the results of the IR-camera. Temperature distribution shows that ?_||/?_⊥ = 5 is quite suitable in HT-7 device, and the maximum heat flux is about 10 MW/m² and average heat flux is about 5 MW/m² on the movable limiter. This method can be also used in other tokamak devices with limiter configuration and circular cross-section shape.     

The application of limiter target electrostatic measurement system in J-TEXT tokamak

The limiter target electrostatic measurement system including limiter grounding current sensors and Langmuir probes have been newly developed for the measurement of the limiter target area on the Joint-Texas Experimental tokamak(J-TEXT).Current sensors fixed between graphite limiters and the vacuum vessel walls are used to measure the currents between limiters and vessel wall.Simultaneously,a rectangular poloidal array containing 54 Langmuir probes is embedded in the graphite tiles of limiters for a more localized measurement.Based on this system,the effect of both the plasma's inherent behavior,including plasma motion and the 2/1 tearing mode,and the electrode biasing on probe and sensor signals have been observed and analyzed in the experiments.     

The design, analysis and alignment of EAST divertor

The divertor of EAST is composed of inner-target, outer-target and dome. Every kind of target has 16 separate modules along toroidal direction. The structure of divertor includes support, heat sink and the first wall graphite. The shape and dimension of divertor are optimized for flexible operation of different elongation and triangularity of plasma. The eddy current and halo current interact with high toroidal magnetic field will produce large forces on divertor, so the structure of divertor must bear these electromagnetic forces. The first wall of divertor must have high assembly accuracy of 0.5 mm. The traditional method using optical instruments to survey directly is impractical because of the especial location of divertor. The benchmark templet integrates with optical instruments to align the divertor is proved simple and effective.     

Evidence of vapor shielding effect on heat flux loaded on flowing liquid lithium limiter in EAST

A lithium (Li) vapour layer was formed around a flowing liquid Li limiter to shield against the plasma incident power and reduce limiter heat flux in the EAST tokamak. The results revealed that after a plasma operation of a few seconds, the layer became clear, which indicated a strong Li emission with a decrease in the limiter surface temperature. This emission resulted in a dense vapour around the limiter, and Li ions moved along the magnetic field to form a green shielding layer on the limiter. The plasma heat flux loaded on the limiter, measured by the probe installed on the limiter, was approximately 52% lower than that detected by a fast-reciprocating probe at the same radial position without the limiter in EAST. Additionally, approximately 42% of the parallel heat flux was dissipated directly with the enhanced Li radiation in the discharge with the liquid metal infused trenches (LIMIT) limiter. This observation revealed that the Li vapour layer exhibited an excellent shielding effect to liquid Li on plasma heat flux, which is a possible benefit of liquid-plasma-facing components in future fusion devices.     

Mechanical design of the second ICRF antenna for EAST

In order to satisfy the requirements of heating plasma on EAST project, 3 MW ion cyclotron range of frequency (ICRF) heating system will be available at the second stage. Based on this requirement, the second ICRF antenna, has been designed for EAST. The antenna which is planned to operate with a frequency ranging from 30 MHz to 110 MHz, comprises four poloidal current straps. The antenna has many cooling channels inside the current straps, faraday shield and baffle to remove the dissipated RF loss power and incoming plasma heat loads. The antenna is supported via a cantilever support box to the external support structure. Its assembly is plugged in the port and fixed on the support box. External slideway and bellows allow the antenna to be able to move in the radial direction. The key components of the second ICRF antenna has been designed together with structural and thermal analysis presented.     

The design and the manufacturing process of the superconducting toroidal field magnet system for EAST device

The toroidal field (TF) magnet system of EAST (HT-7U), which consists of 16 superconducting coils enclosed in steel cases, has been manufactured to generate the magnetic field of 3.5 T at the plasma center to maintain plasma in a tokamak configuration with a current up to 1 MA. The TF coils have an approximately D-shape geometry of 2.6 m wide and 4.0 m high, and operate at a maximum field of 5.8 T. The conductor used in the TF coil is NbTi/Cu cable-in conduit (CIC) conductor, and its operating current is 14.3 kA.In March 2006, the first cooling down of the EAST device has been carried out successfully. The total of TF magnet system has been cooled down from room temperature to 4.5 K, and the TF system has been energized up to 8.2 kA with 5 A/s ramp rate. In September 2006, full performances of the TF magnet system have been reached, and the device of EAST has delivered its first plasma. In addition, the TF magnet system has been routinely operated with a current maintained constant on a whole day basis, for a preliminary program of more than 500 shots.In this paper, the main parts of the design, developmental tests, and the fabrication and assembly of TF coils are described in detail.     

Design and application of GDC on EAST Tokamak

On EAST Tokamak, DC glow discharge (GDC) is developed to clean the first wall of plasma. It can effectively control the recycling of H, C, O impurities and improve the wall conditions. There are four GDCs which distribute equally on the EAST Tokamak vacuum vessel wall. Each GDC is equipped with an anode, a stainless steel cover and four support legs. The anode is insulated from cover with Al₂O₃ ceramics. After a round of experiment, the value of insulation resistance of GDC decreases remarkably due to metallization. To protect the insulation parts and heighten the reliability, ceramic protection covers are used on the GDCs. The other measures which can heighten insulation grades are also taken. After upgrade, the insulation resistance of each GDC between anode and ground is raised highly. When the pressure reaches 4 Pa, H₂-GDC and He-GDC is strarted. Boronation and siliconization are also applied to the device wall conditioning. After GDC cleaning, the impurities and partial pressure of remainder gases in vacuum vessel (VV) is decreased greatly and vacuum degree of VV can reach high easily.     

Conceptual design of Remote Control System for EAST tokamak

The international collaboration becomes popular in tokamak research like in many other fields of science, because the experiment facilities become larger and more expensive. The traditional On-site collaboration Model that has to spend much money and time on international travel is not fit for the more frequent international collaboration. The Remote Control System (RCS), as an extension of the Central Control System for the EAST tokamak, is designed to provide an efficient and economical way to international collaboration. As a remote user interface, the RCS must integrate with the Central Control System for EAST tokamak to perform discharge control function. This paper presents a design concept delineating a few key technical issues and addressing all significant details in the system architecture design. With the aim of satisfying system requirements, the RCS will select rich Internet application (RIA) as a user interface, Java as a back-end service and Secure Socket Layer Virtual Private Network (SSL VPN) for securable Internet communication.     

EAST中性束注入器稳态偏转磁铁的参数设计与估算研究

中性束注入器偏转磁铁是剥离束流中剩余离子的关键设备,它与剩余离子吞食器等内部部件构成了中性束注入器的束偏转系统。束偏转系统的性能对中性束注入器束流的品质及其束传输效率发挥着重要作用。本文根据EAST(Experimental Advanced Superconducting Tokamak,EAST)中性束注入器对束偏转系统的要求,对其偏转磁铁各性能参数进行了估算。为中性束注入器设计了一台用以剩余离子180°偏转的偏转磁铁。该偏转磁铁采用H型二极电磁铁结构;其磁极端面设计为138cm×47cm的圆角矩形结构;其线圈设计为每侧2饼,每饼2层,每层8根的串联结构,导线选用外方内圆空心铜导体,以满足偏转磁铁稳态运行的需要。该设计的偏转磁铁在370 A励磁电流条件下,可提供80keV氘离子束偏转所需的磁场。实验测试结果显示:500 A励磁电流稳态运行条件下,偏转磁铁线圈冷却水温升约21.5℃,该设计结构的偏转磁铁满足EAST中性束注入器满参数稳态运行和未来运行参数逐步提高的需要。     

Physical design of a new set of high poloidal mode number coils in the EAST tokamak

An external resonant magnetic perturbation (RMP) field, which is an effective method to mitigate or suppress the edge localized mode (ELM), has been planned to be applied on the ELM control issue in ITER. A new set of magnetic perturbation coils, named as high m coils, has been developed for the EAST tokamak. The magnetic perturbation field of the high m coils is localized in the midplane of the low field side, with the spectral characteristic of high m and wide n, where m and n are the poloidal and toroidal mode numbers, respectively. The high m coils generate a strong localized perturbation field. Edge magnetic topology under the application of high m coils should have either a small or no stochastic region. With the combination of the high m coils and the current RMP coils in the EAST, flexible working scenarios of the magnetic perturbation field are available, which is beneficial for ELM control exploration on EAST. Numerical simulations have been carried out to characterize the high m coil system, including the magnetic spectrum and magnetic topology, which shows a great flexibility of magnetic perturbation variation as a tool to investigate the interaction between ELM and external magnetic perturbation.     

EAST低温系统主运行模式的控制流程设计与分析

EAST托卡马克核聚变实验装置是世界上从事核聚变研究的先进科学设备。EAST低温系统是该装置的主要子系统之一,其相应DCS控制系统具有很高的稳定性和可扩展性,各部分的控制相互独立、并行执行。本文详细介绍和分析了EAST低温控制系统及在正常降温、稳态和失超模式下的控制流程设计。     

The Procedure for Assembling the EAST Tokamak

Due to the complicated constitution and high precision requirements of the EAST superconducting tokamak, a meticulous assembling procedure and measurement scheme must be established. The big size and mass of the EAST machine＇s components and complicated configuration with tight installation tolerances call for a highly careful assembling procedure. The assembling procedure consists of three main sub-procedures for the assembling of the base, of the tori of the VV, the vacuum vessel TS and the TF, and of the peripheral parts respectively. Before the assembly, a reference framework has been set up by means of an industrial measurement system with reference fiducial targets fixed on the wall of the test hall. In this paper, the assembling procedure is described in detail, the survey control system of the assembly is discussed, and progress in the assembly work is also reported.     

Optimal design of the first stage of the platefin heat exchanger for the EAST cryogenic system

The size of the heat exchanger is an important factor determining the dimensions of the cold box in helium cryogenic systems. In this paper, a counter-flow multi-stream plate-fin heat exchanger is optimized by means of a spatial interpolation method coupled with a hybrid genetic algorithm.Compared with empirical correlations, this spatial interpolation algorithm based on a kriging model can be adopted to more precisely predict the Colburn heat transfer factors and Fanning friction factors of offset-strip fins. Moreover, strict computational fluid dynamics simulations can be carried out to predict the heat transfer and friction performance in the absence of reliable experimental data. Within the constraints of heat exchange requirements, maximum allowable pressure drop, existing manufacturing techniques and structural strength, a mathematical model of an optimized design with discrete and continuous variables based on a hybrid genetic algorithm is established in order to minimize the volume. The results show that for the first-stage heat exchanger in the EAST refrigerator, the structural size could be decreased from the original2.200?×?0.600?×?0.627(m~3) to the optimized 1.854?×?0.420?×?0.340(m3), with a large reduction in volume. The current work demonstrates that the proposed method could be a useful tool to achieve optimization in an actual engineering project during the practical design process.     

The Application and Improvement of Helium Turbines in the EAST Cryogenic System

The helium cryogenic system supplies supercritical helium cooling capacity at 4.5 K for the superconducting magnets in the EAST superconducting tokamak.Four low-temperature helium turbines are used in the 2 kW helium refrigeration system,and their performance and reliability are critical for the continuous operation of the tokamak.The turbines were made by the Helium Mechanics Company in Russia.The start-up process is very unstable,easily broken,and thus testing and improvements are needed.In this paper,we analyze the structure of the helium turbine,make improvements,and describe the testing process and results of the improved helium turbines.Some of the operational experiences during the start-up process and tests are also presented.     

The Structure Analysis of the EAST Updated Divertor

After EAST_2012 campaign, the updated divertor (upper part) has been designed and installed in EAST. The updated upper divertor is designed as cassette structure like ITER divertor. The structure analysis has been carried out for divertor cassette design in the worst case loads of disruption with the maximum 1 MA of the plasmas current. A series of finite elements analysis indicates that the divertor cassette is able to withstand the electromagnetic (EM) load resulting from halo current, combined with EM load caused by eddy current, gravity and pressure of cooling water.