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The EAST is an Experimental Advanced Superconducting Tokamak. The mission of the EAST Project is to bring out scientific issues on the continuous nonburning plasma scenario of steady-state operation and engineering issues on establishing the basis of technology for superconducting tokamak. Superconducting magnets were chosen for all poloidal field (PF) and toroidal field (TF) systems since the engineering mission is to establish the technology basis of full superconducting Tokamak for future fusion reactors. The superconducting magnets of EAST consist of sixteen TF coils and fourteen PF coils (seven coil-pairs). To obtain the good performance of the superconducting magnets, all TF magnets and most PF magnets have been tested before assembly. The assembly of the main device was completed in the end of 2005 and at the beginning of 2006, we made successfully the first engineering commissioning of the EAST system. Up to now the EAST device has been used in 4 operation campaigns and has achieved good experimental results.

The construction of experimental advanced superconducting Tokamak (EAST) was completed at the Institute of Plasma Physics of Chinese Academy of Sciences (CASIPP) in January of 2006 after challenging techniques over the period of 8 years. The followed cooling-down test was also successfully performed and all superconducting magnets including two current leads and seven bus lines entered the cryogenic superconducting state smoothly. Total 260 shots have been fired. The longest TF current duration was 5000 s and the highest TF current was 8200 A (2T). All the testing results showed that the EAST machine and its sub-systems have been successfully built. EAST machine is a fully superconducting Tokamak. It consists of more 20,000 parts and possesses the configuration of 7600 mm in diameter, 6600 mm in height and the weight of 400 tonnes. The complicated general assembly of EAST began in July of 2003 and nearly took two and half years. During the process, not only a series of common difficulties for a large Tokamak but also many technique problems for a superconducting Tokamak itself needed to be solved properly. This paper emphasizes on some key technology issues that were faced and solved in EAST assembly process including optimization of assembly procedure, survey and alignment, ensuring of thermal insulation gap, installation of unique parts, treatment of electric insulation, improvement on heat conduction, control on deformation of vacuum vessel, etc. Besides, the quality control (QC), which made important contributions to the general assembly is also described.

EAST (HT-7U) is a large fusion experimental device being built at IPP, Hefei, China. Its superconducting magnet system consists of sixteen Toroidal Field (TF) coils and fourteen Poloidal Field (PF) coils. The TF coil includes the winding pack with a square cable-in-conduit (CIC) type superconductor in NbTi cooled by a force flow of supercritical helium, the welded case structure, the gravity support which is composed of pedestals with flexible plates, the coil joint with low resistance at high current and other components. At present, the production of all TF coils has been finished and all of the TF coils have been installed in the machine. This paper describes mainly the special fabrication processes of the TF coil

The HT-7U superconducting (SC) tokamak will have a long-pulse capability, a flexible poloidal field (PF) system, and auxiliary heating and current drive systems, and it will be able to accommodate divertor heat loads that make it an attractive test for the development of advanced tokamak operating modes. The greatest progress has been made on the engineering design of the HT-7U SC tokamak device, including the calculation and simulation of plasma shaping and control of the PF system as well as calculation and analyses of stress and deformation distribution on the main components caused by dynamic electromagnetic forces, vacuum pressure, temperature differences, etc. Significant research and development progress on the design and the testing of the cable-in-conduit conductor of the toroidal field and PF has been made. A test facility system for the SC magnets of HT-7U has been set up and operated.

HT-7U is a large fusion experimental device. It will be built at the Institute of Plasma Physics of Chinese Academy of Sciences. The mission of HT-7U is to develop the scientific basis for a continuously operating tokamak fusion reactor. This paper describes only the toroidal field (TF) superconducting magnet system of HT-7U. In this paper, design criteria of conductor and stability analysis, coil winding and support structure design of magnet system, mechanical calculation, stress analysis and heat load evaluation are given.

On-site winding of magnet coils of HT-7U started in September 1998, using a numerically controlled winding machine. Magnet coils are all forced-flow superconducting coils, consisting of superconducting Toroidal Field (TF) coils and superconducting Poloidal Field (PF) coils. All of TF and PF coils will use NbTi Cable-in-Conduit Conductor (CICC) cooled with supercritical helium. This paper reports on the set up of a new winding facility with unique capabilities for continuous winding of long length CICC. An analytical method used to predict conduit springback before winding is presented and the results are compared to the ones obtained during winding. Research and development of winding of the TF Dummy Coil (TFDC) has been carried out and a winding method has been developed by winding trials. Windings of the TFDC and CS Model Coil (CSMC) have been successfully completed by using the developed technique. In R&D, long dummy conductor and short sample jackets have been used to demonstrate a plastic deformation characteristic of CICC during the winding process. The developed winding method can be applied to fabricate all TF and PF coils.

The superconducting tokamak project HT-7U, aiming at steady-state advanced operation mode, will make a contribution to future steady-state tokamak reactors. The scientific and the engineering missions of the project are to study physics issues of the steady-state tokamak operation and to establish technology basis of full superconducting tokamaks. It features: superconducting toroidal field system and poloidal field system, non-inductive current drive and plasma heating systems, flexibility and reliability of plasma shaping control, J(r) and P(r) control, replaceability of plasma facing components and divertors for power and particle handling study in steady-state operation and advanced diagnostic measurements. The physics design and the engineering design have been completed essentially. The key R&D programs of the tokamak device have been successful. The assembly of the device has begun. It is planned to obtain the first plasma in 2005.

The HT-7U superconducting tokamak is an advanced steady-state plasma physics experimental device to be built at the Institute of Plasma Physics, the Chinese Academy of Sciences (ASIPP). HT-7U have a long pulse (60-1000 s) capability, a flexible PF system, and auxiliary heating and current drive systems, and will be able to accommodate divertor heat loads that make it an attractive test for the development of advanced tokamak operating modes. Now, the engineering design of the HT-7U device is in progress. The engineering design incorporates the superconducting toroidal field (TF) and poloidal field (PF) magnets, the vacuum vessel, the thermal shields, the cryostat and the current leads. This paper provides an overview of the HT-7U design emphasizing developments in the device design during the past year.