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 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.     

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.     

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.     

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。     

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.     

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.     

Shutdown dose rate calculation code system and its application to EAST tokamak

For safe operation and maintenance of nuclear devices, it is very important to predict the dose rate distribution after shutdown. Based on the rigorous two-step (R2S) method, a new shutdown dose rate calculation code system was developed for nuclear devices with large dimension and complex geometries. By coupling the Monte Carlo particle transport calculation code MCNP with the activation simulation code FISPACT, the dose rate calculation codes integrate the functions of neutron transport calculation, activation calculation and decay gamma transport calculation. This code system was applied to the shutdown dose analysis for experimental advanced superconducting tokamak (EAST). Three-dimensional dose rate distributions of the EAST for different cooling times and spatial locations were obtained. The results can be used to assist radiation protection in EAST.     

Low-temperature operating regime of the tokamak evacuating limiter

Conclusions The conditions for realizing the regime of strong recycling of a cold dense plasma of an evacuating limiter were determined based on a previously proposed model for describing the limiter layer of a tokamak. The scaling for the dependence of the gas pressure in the evacuation system on the average plasma density in the limiter layer was found, and agreed quantitatively with the results of measurements on the Alcator and ISX-B tokamaks.For the tokamak reactor of the INTOR scale the calculations show that the low-temperature operating regime of the evacuating limiter can be realized with a quite low pumping rate. It has the advantages of reduced erosion of the limiter and small fluxes of impurities into the working volume of the reactor. In addition, the relative concentration of the helium ash in the limiter layer does not exceed 2–3%, but the density of the main plasma is comparable to the proposed average density in the reactor. The concept of a stochastic limiter is of interest for lowering the plasma density in the limiter layer and lowering the thermal loads on the limiter.Translated from Atomnaya Énergiya, Vol. 61, No. 6, pp. 440–443, December, 1986.     

三维停堆剂量率计算程序研发及其在EAST上的应用

为了精确分析核装置停机后周围空间的三维辐射剂量场分布情况,本文基于停堆剂量计算的严格二步法思想,发展了基于蒙特卡罗输运计算程序MCNP和欧洲活化计算程序FlSPACT的耦合三维停堆剂量计算程序,实现了中子输运、材料活化和光子剂量计算的自动耦合.将该程序初步应用于EAST装置停堆剂量计算,得到了托卡马克装置停堆后周围空间...     

基于Oracle的高光谱数据管理平台设计与应用

高光谱遥感数据具有海量、高维、图谱合一等技术特点,随着数据量的增加,传统的文件管理方式越来越无法满足实际的需要。本文提出了将高光谱遥感数据按像元顺序或者波段顺序分割,按大型对象存储的技术思路,设计了高光谱遥感数据的存储结构,并开发了相应的数据管理平台。实际应用表明,该平台能够较好地完成高光谱数据的存储、管理与查询,取得了理想的应用效果。     

The application of elastic-plastic fracture mechanics parameters in fracture safe design

The report summarizes some of the methods which are currently used for assessing the fracture toughness of materials under elastic and elastic-plastic conditions. The main parameters which are considered are (1) plane strain fracture toughness (K_Ic), (2) equivalent energy (K_Icd), (3) contour integral (J) and (4) crack opening displacement (COD). Gross strain crack tolerance and stress concentration methods are also discussed.It is concluded that of these parameters, the contour integral and the crack opening displacement have most potential for future development. These two parameters are shown to be equivalent, however, at the present stage of development the COD concept has several advantages over the J concept. Firstly, the COD concept is able to take into account, secondary stresses, such as welding residual stresses. Because these stresses are in equilibrium, they do not appear in energy measurements to evaluate J. Secondly, the COD value is a physical measure of the crack tip conditions which includes the effect of stress state and thickness. It is, therefore, possible to measure and calculate COD levels for cracks in real structures. It is not possible to evaluate J for real structures since J methods are appropriate only to in-plane problems. This also means that partial wall (thumbnail) flaws are better characterized by the COD concept.The COD concept has been developed to a stage where it is possible to estimate the significance of flaws in welded structures provided the toughness of the material and the acting stresses or strains are known. This development is described and the method used to analyze tests on model pressure vessels with 6″ thick walls. A comparison is made with other methods, and it is concluded that although the COD analysis gives conservative estimates of the flaw size to cause failure, further work is necessary to be able to predict vessel burst conditions when failure is preceded by extensive plasticity and stable ductile tearing. A simple nomogram to determine COD levels to ensure leak before break conditions is also developed.     

The cryogenic control system of EAST

A large scale helium cryogenic system is one of the key components for the EAST tokamak device for the cooling of PF and TF coils, structures, thermal shields, buslines, current leads and cryopumps. Since the cooling scheme of the EAST cryogenic system is fairly complicated, a reliable and flexible control system is required for cryogenic system. The cryogenic control system is based on DeltaV DCS which is the process control software developed by Emerson Company. The EAST cryogenic system has been in operation for four years and has been proved to be safe, stable and energy saving by the past 7 experiments. This paper describes the redundant control network, hardware configuration, software structure, auxiliary system and the new development in the future.     

中子符合计数器在设计与应用时应注意的几个问题

在核材料衡算中,对铀、钚的分析可以采用化学的方法,但需要对样品进行破坏,且操作复杂,测量周期较长。由于铀、钚都可以自发裂变产生中子,可以进行中子测量,中子测量不需要破坏样品,可以对大体积样品进行快速分析,且误差较小。就中子符合计数器进行了参数优化实验,更好地理解了中子探测器的设计原理,包括中子计数统计、屏蔽厚度、不同材料对中子的散射、高压影响、门宽设置、死时间的修正、衰减时间确定、探测效率验证和γ干扰影响等;同时进行了测量条件实验,包括偶然事件影响等。为更好地使用此技术,具有一定的借鉴意义。     

The Test Facility for the EAST Superconducting Magnets

A large facility for testing superconducting magnets has been in operation at the Institute of Plasma Physics of the Chinese Academy of Sciences since the completion of its construction that began in 1999. A helium refrigerator is used to cool the magnets and liquefy helium which can provide 3.8 K-4.5 K, 1.8 bar-5 bar, 20g/s-40g/s supercritical helium for the coils or a 150 L/h liquefying helium capacity. Other major parts include a large vacuum vessel （3.5 m in diameter and 6.1 m in height） with a liquid nitrogen temperature shield, two pairs of current lead, three sets of 14.5 kA-50 kA power supply with a fast dump quench protection circuitry, a data acquisition and control system, a vacuum pumping system, and a gas tightness inspecting devise. The primary goal of the test facility is to test the EAST TF and PF magnets in relation to their electromagnetic, stability, thermal, hydraulic, and mechanical performance. The construction of this facility was completed in 2002, followed by a series of systematic coil testing. By now ten TF magnets, a central solenoid model coil, a central solenoid prototype coil, and a model coil of the PF large coil have been successfully tested in the facility.