托卡马克HT-7U低温超导磁体的绝缘系统

主要介绍低温超导托卡马克HT-7U中磁体的绝缘结构,绝缘材料的特性,绝缘结构形成的工艺,绝缘层的电性能和力学性能等。     

Recent Progress of HT-7U Superconducting Tokamak

HT-7U is a superconducting tokamak. which is being constructed in Institute of Plasma Physics, Chinese Academy of Sciences. The mission of the HT-7U project is to develop a scientific and engineering basis of the steady state operation of advanced tokamak.The engineering design of the device has been optimized. The R&D program is going on. Short samples of the conductor and a CS model coil were tested. All the TF and PF coils will be manufactured and tested in Institute of Plasma Physics. Therefore, a 600-meter long jacketing line for cable-in-conduit conductors along with two winding machines, a set of VPI equipment and a test facility for the TF and PF coils are ready in ASIPP now. In this paper, the recent progress of the HT-7U is described.     

Insulating process for HT-7U central solenoid model coils

The HT-7U superconducting Tokamak is a whole superconducting magnetically confined fusion device. The insulating system of its central solenoid coils is critical to its properties. In this paper the forming of the insulating system and the vacuum-pressure-impregnating (VPI) are introduced, and the whole insulating process is verified under the superconducting experiment condition.     

Finite Element Analyses and Instrumentation Layout for Single Coil Testing of TF Coils in HT-7U

The HT-7U tokamak is a magnetically-confined full superconducting fusion device, consisting of superconducting toroidal field (TF) coils and superconducting poloidal field (PF) coils. These coils are wound with cable-in-conductor (CICC) which is based on UNK NbTi wires made in Russian '. A single D-shaped toroidal field magnet coil will be tested for large and expensive magnets systems before assembling them in the toroidal configuration. This paper describes the layout of the instrumentation for a superconducting test facility based on the results of a finite element modeling of the single coil of toroidal magnetic field (TF) coils in HT-7U tokamak device. At the same time, the design of coil support structure in the test facility is particularly discussed in some detail.     

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.     

Structural Analysis of Central Solenoid for HT—7U

The central solenoid(CS) composed of cable-in-conduit(CIC) type conductor is an important part of the HT-7U device,The CS coils can be considered from the viewpoint of micromechanics as a compostie material.And then,the residual stiffness is computed according to the micro-damage modeling of continuum damage mechanics,These material properties have been used as input data for finite element method(FEM) analysis.In this paper the computational analysis of the stress and the displacement on the central solenoid are made by the finite element analysis system COSMOS/M2.0 under operating temperature.According to the analytical results.the CS coils are all satisfied with designed safety criteria.     

Design of the Cryostat for HT—7U superconducting Tokamak

The cryostat of HT-7U tokamak is a large vacuum vessel surrounding the entire basic machine with a cylindrical shell,a dished top and a flat bottom.The main function of HT-7U cryostat is to provide a thermal barrier between an ambient temperature test hall and a liquid helium-cooled superconducting magnet.The loads applied to the cryostat are from sources of vacuum pressure,dead weight,seismic events and electromagnetic forces originated by eddy currents.It also provides feed-through penetrations for all the conecting elements inside and outside the cryostat.The main material selected for the cryostat is stainless steel 304L.The structural analyses including buckling for the cryostat vessel under the plasma operation condition have been carried out by using a finite element code.Stress analysis results show that the maximum stress intensity was below the allowable value.In this paper,the structural analyses and design of HT-7U cryostat are emphasized.     

核聚变实验装置HT-7U停机辐射剂量率三维计算与分析

在基于三维蒙特卡罗方法的聚变装置停机剂量率计算方法“严格两步法”(R2S)的基础上,首先建立了核聚变托卡马克实验装置HT-7U三维精确模型,然后对HT-7U各种D-D放电模式下的停机剂量率进行了详细的三维计算与分析,从而为装置实验方案及实验维修人员的安全操作规程的制定提供了理论基础,也对装置的辐射防护问题具有参考价值。     

The Paralleling of High Power High Frequency Amplifier Based on Synchronous and Asynchronous Control

The vertical position of plasma in the HT-7U Tokamak is inherently unstable. In order to realize active stabilization, the response rate of the high-power high-frequency amplifier feeding the active control coils must be fast enough. This paper analyzes the paralleling scheme of the power amplifier through two kinds of control mode. One is the synchronous control; the other is the asynchronous control. Via the comparison of the two kinds of control mode, both of their characteristics are given in the text. At last, the analyzed result is verified by a small power experiment.     

Comparison of the Feasible Schemes for HT-7U Assembly Simulation

The whole superconducting HT-7U Tokamak is a high-cost and large-scale compli-cated device.The assembly requiremem of HT-7U device is arduous and strict.At present,therehave been no guiding principle for the assembly of the device,but assembly simulation can help theengineer plan and make decision by an intuitional and visual way before its actual assembly.The     

Feedback Control of Plasma Current and Horizontal Position in HT-7

1. IntroductionA superconducting tokamak HT-7 has been estab-lished at ASIPP, Hefei, China. The machine.was de-signed to mainly investigate the reactor-relevant ls-sues, such as edvanced operation modes and plasmawall interastions in the near-steady-state condition.Its poloidal fie1d coils include ohmic heatlng coi1s'bias field coils' vertical field coils and horizontalfie1d coi1s (See Fig.1), being connected to indlvldualpower supplies which are all the thyristor--controlledrectifier unlt…     

Analysis of charge-exchanged neutral particles during IBW heating in the HT-7 tokamak

The charge-exchange neutral particles fluxes and energy distribution in IBW heated plasma were investigated in the HT-7 tokamak. The RF frequency was 30 MHz and with an injecting power up to 200 kW. It is observed that the plasma performance is obviously enhanced by IBW heating. The electron temperature was increased by 0.5 keV and the central line averaged electron density was doubled. The neutral particle fluxes of high-energy increased and the bulk ions were heated during IBW heating. The ion temperature was increased by 0.3 keV and the ion heating efficiency of (2–3) eV kW⁻¹ × 10¹³ cm⁻³ was achieved. The velocity distribution of charge-exchanged neutral particles appears to be Maxwellian without high-energy tail ions up to the maximum RF power.     

二维中子-光子输运计算程序DOT3.5在核聚变实验装置HT-7U辐射屏蔽设计中的应用

利用二维中子－光子输运程序ＤＯＴ３．５计算和分析ＨＴ－７Ｕ在Ｄ－Ｄ放电运物时实验大厅内外的中子、光子通量及剂量当量率的空间分布,并评估大厅顶部不同防护材料的屏蔽效果,为辐射防护设计提供参考。     

HT-7 超导托卡马克上辐射剂量测量与分析

HT-7超导托卡马克是我国正在运行中的中型准稳态磁约束聚变实验研究装置,运行时产生一个复杂的瞬发混合辐射场。本文介绍了该装置上中子辐射剂量率的测量方法以及在欧姆、波辅助加热、长脉冲等放电模式下的剂量率特点,并且给出了X与7射线的总辐射剂量测量方法和分析结果。在严格遵守安全操作规程的前提下,目前的等离子体物理实验对人员产生的总辐射剂量低于国家辐射安全标准的规定。     

HT-7U NbTi 管内电缆导体瞬态稳定性的实验研究与分析

在SULTAN 测试设备上进行了含分离铜股线CICC瞬态稳定性的实验研究,应用脉冲场对样品的脉冲场区域(390 mm)进行感应加热,发现设计的含分离铜股线CICC能够经受住很大的瞬态磁扰动,分析了这个现象的原因,并就股线上的电阻层对稳定性的影响进行了分析,对4个CICC导体的稳定性差异进行分析和稳定性裕度的理论计算,由理论计算值和实验测量值进行比较分析,为HT-7U纵场和极向场NbTi CICC的选择提供实验依据.     

Evaluation of The Thermal Performance of Multi—Element Doped Graphite under Steady—State High Heat Flux

Multi-element doped graphite,GBST1308 has been developed as a plasma facing material(PFM) for high heat flux components of the HT-7U device.The thermal performance of the material under steady-state(SS) high heat flux was evaluated under actively cooling conditions,the specimens were mechanically joined to copper heat sink with supercarbon sheet as a compliant layer between the interfaces.The experiments have been performed in a facility of ACT (actively cooling test stand) with a 100kW electron gun in order to test the suitability and the loading limit of such materials.The surface temperature and bulk temperature distribtuion of the specimens were investigated.The experimental results are very encouraging that when heat flux is not more than 6 MW/m^2,the surface temperature of GBST1308 is less than 1000℃,which is the lowest,compared with IG-430U and even with CX-2002U(CFC),The primary results indicate that the mechanically-joined material system by such a proper design as thin tile.Super compliant layer,GBST as PFM and copper-alloy heat sink,can be used as divertor plater for HT-7U in the first phase.     

Technical diagnosis system for EAST tokamak

Technical diagnosis system (TDS) is one of the important subsystems of EAST (experimental advanced superconducting tokamak) device, main function of which is to monitor status parameters in EAST device. Those status parameters include temperature of different positions of main components, resistance of each superconducting (SC) coils, joint resistance of SC coils and high-temperature superconducting (HTS) current leads, strain of cold-quality components endured force, and displacement and current of toroidal field (TF) coils in EAST device, which are analog input signals. In addition there are still some analog and digital output signals. The TDS monitors all of those signals in the period of EAST experiments. TDS data monitoring is described in detail for it plays important role during EAST campaign. And how to protect the SC magnet system during each plasma discharging is presented with data of temperature of coolant inlet and outlet of SC coils and feeders and cases of the TF coils and temperature in the upper and middle and bottom of the TF coil case.During construction of the TDS primary difficulties come from installation of Lakeshore Cernox temperature sensors, strain measurement of central solenoid coils support legs and installation of co-wound voltage sensors for quench detection. While during operation since the first commissioning big challenges are from temperature measurement changes in current leads and quench detection of PF coils. Those difficulties in both stages are introduced which are key to make the TDS reliable. Meanwhile analysis of experimental data like temperature as a back up to testify quench occurrence and stress on vacuum vessel thermal shield and vacuum vessel have also been discussed.     

Magnet design considerations for Tokamak fusion reactors

Design problems for superconducting ohmic heating and toroidal field coils for large Tokamak fusion reactors are discussed. The necessity for making these coils superconducting is explained, together with the functions of these coils in a Tokamak reactor. Major problem areas include materials related aspects and mechanical design and cryogenic considerations. Projections and comparisons are made based on existing superconducting magnet technology. The mechanical design of large-scale coils, which can contain the severe electromagnetic loading and stress generated in the winding, are emphasized. Additional major tasks include the development of high current conductors for pulsed applications to be used in fabricating the ohmic heating coils. It is important to note, however, that no insurmountable technical barriers are expected in the course of developing superconducting coils for Tokamak fusion reactors.     

Preliminary design of KTX magnet system

KTX is a reversed field pinch magnetic confinement device of which the magnet system is designed in ASIPP and USTC. The main parameter of KTX is between RFX and MST. Its magnet system includes the toroidal field (TF) winding and poloidal field (PF) winding (ohmic heating winding and equilibrium field winding), which are less complex than tokamak device due to the fact that a tokamak requires a superconducting system to perform quasi-steady state operation and achieve Q > 10. However, the most important part of the magnet system design lies in how to keep the TF magnetic field ripple, as well as any kinds of stray field, to a minimum value. The main design activities of the KTX magnet system are presented as detailed as possible in this paper, and the main activities which have already been completed include magnet coils position and winding, insulation design, plasma modeling prediction, thermal analysis, magnetic field calculations were analyzed and so on. The magnet system design is one of the major activities for KTX device design, which is effective guarantee for the future R&D and manufacture. Besides, the detailed design activities should be continuously optimized and changed based on the results from future R&D and relevant tests.