Optical Design of ECEI Diagnostic System for HT-7 Tokamak

Electron cyclotron emission imaging system in the frequency range of 95 GHz-125 GHz is going to be constructed for a two-dimensional diagnosis of the electron temperature profiles and fluctuations on the HT-7 Tokamak.The optical design for the ECEI diagnostic system is completed.Because of the superconducting technology used in HT-7,the vacuum chamber is rather thick (630mm),the height of the horizontal windows is limited(maximum450mm),which constrains greatly the ECE imaging Gaussian beam that passing through the windows.We herecomes to make a design compromise between the number of the beams that can pass through the windows and the spatial resolution (around 1.1 cm).We also find that due to the field curvature of the optical system,the gaussian beams of edge channels are always overlapped.To flatten the field curvature,it is needed to insert a concave made of a material with a low refractive index(compared with the one used in the convex).But the suitable material has not been available so far,therefore the deterioration of the resolution in some channels(e.g.the edge channels) is acceptable.     

The Experiment of Modulated Toroidal Current on HT-7 and HT-6M Tokamak

The Experiments of Modulated Toroidal Current were done on the HT-6M tokamak and HT-7 superconducting tokamak. The toroidal current was modulated by programming the Ohmic heating field. Modulation of the plasma current has been used successfully to suppress MHD activity in discharges near the density limit where large MHD m = 2 tearing modes were suppressed by sufficiently large plasma current oscillations. The improved Ohmic confinement phase was observed during modulating toroidal current (MTC) on the Hefei Tokamak-6M (HT-6M) and Hefei superconducting Tokamak-7 (HT-7). A toroidal frequency-modulated current, induced by a modulated loop voltage, was added on the plasma equilibrium current. The ratio of A.C. amplitude of plasma current to the main plasma current △Ip/Ip is about 12% ～ 30%. The different formats of the frequency-modulated toroidal current were compared.     

Plasma Shape and Current Control Simulation of HT-7U Tokamak

This paper describoes the discharge simulation of HT-7U tokamak plasma equilibrium and plasma current by solving MHD equations and surface average transport equations using an equilibrium evolution code. The simulated result shows the evolution of plasma parameter versus time .The simulated result can play an important role in the design of the plasma equilibrium and control system of a tokamak.     

Study of Liquid Phase Shifter for ICRF on EAST

A method of current drive with Ion Cyclotron Range of Frequency （ICRF） on Experimental Advanced Superconducting Tokomak （EAST） is described. A variety of liquid silicon oil heights in the phase shifter will bring the phase difference to the current drive. It is found that the current drive can be achieved by using the phase shifter. The liquid phase shifter is one of the impedance matching systems too.     

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.     

The Circulating Current Analysis and Parameter Design of Interphase—Reactor of HT—7U Toroidal Power Supply

This paper analyzes the circulating current which is produced by HT-7U superconducting toroidal power supply-two sets of two-reverse-star converter with an interphase-reactor in parallel running on the basis of the output voltage mathematical equation of three-phaase semi-wave converter circuit.A new iden of omitting interphase-reactor between two converters is proposed,and the parameter design of interphase-reactor of HT-7 toroidal power supply is presented.Simulated results demonstrate the validity of this new project.     

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

The HT-TU 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 [1]. 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 resultsof 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 Positions and Number of Flux Loops and Magnetic Probes for the HT—7U Superconducting Tokamak

The plasma shape and other paremeters such as βp,li is important for the tokamak deveice where the palsma has a non-circular cross-section of sufficient elongation.The measuered signals of magnetic probes and flux loops are used to reconstruct the plasma shape and the current profile in device operation and plasma shape feed back control system.So the number and positions of magnetic probes and flux loops provides the basis of the plasma reconstruction.This paper instroduce how to use EFIT code (equilibrium fitting code)to determine the number and positions of the magnetic probes and flux loops.The simulation result is given also.     

Using Poloidal Flux as Feedback Control Signal for Injected Power of LHCD in Long Pulse Discharge on HT-7 Tokamak

Long pulse discharge is one of the important goals of HT-7 superconducting tokamak experiments. For ITER （International Thermonuclear Experimental Reactor） or a tokamak reactor, carrying out a steady operation is one of the main techniques. For long pulse discharges on HT-7 the poloidal flux is used as the feedback signal to control the injected power of LHCD （Low Hybrid Current Drive） system. Experimental results are presented.     

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.     

Superconducting Busline Design for HT—7U Tokamak

Superconducting (SC) tokamak HT-7U has seven pairs of buslines connecting toroidal/Poloidal coils and the current leads,These SC buslines(SCBLs) share a common cryostat and are made of the calbe in conduit conductors(CICCs) arranged as a decoupling configuration.In order to reduce the heat loads conducted from the seven current leads with a capacity of 15kA during the magnets cooldown.the buslines with a much lower thermal conduction were employed in comparison with the current leads,and a special cooling loop was designed.     

EAST托卡马克的中性束注入方案

高能中性束注入(Neutral beam injection,NBI)是核聚变装置托卡马克采用的芯部辅助加热和非感应电流驱动主要手段之一.本文介绍了国家大科学工程全超导托卡马克实验装置(Experimental advanced super-conductingtokamak,EAST)上的高能NBI加热方案及注入器的工程要求,并讨论了中性束在EAST等离子体中的传输等相关问题.     

Design and development of a device management platform for EAST cryogenic system

EAST cryogenic system is one of the critical sub-systems of the EAST tokamak device. It is a large scale helium cryoplant, which adopts distributed control system to realize monitoring and control of the cryogenic process and devices. However, the maintenance and management of most field devices are still in the corrective maintenance or traditional preventive maintenance stage. Under maintained or over maintained problems widely exist, which could cause devices fault and increase operation costs. Therefore, a device management platform is proposed for a safe and steady operation as well as fault diagnosis and predictive maintenance of EAST cryogenic system.This paper presents the function design and architecture design of the cryogenic device management platform. This platform is developed based on DeltaV DCS and acquires monitoring data through OPC protocol. It consists of three pillars, namely device information management, device condition management, and device performance monitoring. The development and implementation of every pillar are illustrated in detail in this paper. Test results and discussions are presented in the end.     

Assessment of Radiation Damage to the Structural Material of EAST Tokamak

Radiation damage to structural material of fusion facilities is of high concern for safety. The superconducting tokamak EAST will conduct D-D plasma experiments with the neutron production of 1015 neutrons per second. To evaluate the material radiation damage a programme system has been devised with the Monte Carlo transport code MCNP-4C, the inventory code FISPACT99, a specific interface, and the fusion evaluated nuclear data library FENDL-2. The key nuclear responses, i.e. fast neutron flux, displacement per atom, and the helium and hydrogen production, are calculated for the structural material SS-316L of the first wall, and the vacuum vessel, using this programme. The results demonstrate that the radiation damage to the structural material is so little that it will not lead to any significant change of material properties according to the reference design. This indicates that there is a large potential space for EAST to test advanced operation regime from the viewpoint of structural material safety.     

Design,Analysis and R＆D of the EAST In-Vessel Components

In-vessel components are important parts of the EAST superconducting tokamak. They include the plasma facing components, passive plates, cryo-pumps, in-vessel coils, etc. The structural design, analysis and related R＆D have been completed. The divertor is designed in an up-down symmetric configuration to accommodate both double null and single null plasma operation. Passive plates are used for plasma movement control. In-vessel coils are used for the active control of plasma vertical movements. Each cryo-pump can provide an approximately 45 m^3/s pumping rate at a pressure of 10^-1 Pa for particle exhaust. Analysis shows that, when a plasma current of 1 MA disrupts in 3 ms, the EM loads caused by the eddy current and the halo current in a vertical displacement event （VDE） will not generate an unacceptable stress on the divertor structure. The bolted divertor thermal structure with an active cooling system can sustain a load of 2 MW/m^2 up to a 60 s operation if the plasma facing surface temperature is limited to 1500 ℃. Thermal testing and structural optimization testing were conducted to demonstrate the analysis results.     

The batch production for superconducting magnet coils of EAST (HT-7U)

The experimental advanced superconducting tokamak (EAST) is being constructed at the Institute of Plasma Physics of Chinese Academy of Sciences (CASIPP). The EAST project, approved by the Chinese government as a national mega project of science research is a fully superconducting tokamak. The most key component for EAST is the superconducting magnet coils (SMCs), which consists of 16 toroidal field coils (TFCs) and 14 poloidal field coils (PFCs). In 2003, three prototypes, one TFC and two PFCs, were successfully completed and passed a series of cryogenic tests. Batch production, needed for the SMCs has begun at CASIPP since 2002. Up to now, all 58 CIC conductors with a total length of 32 km, 12 TFCs out of 16 and 10 PFCs out of 14 have been fabricated. This paper emphasizes on the various technology issues that must be faced and solved for four R&D lines of SMCs after transforming to batch production. Quality control methods in process are also described.