D.M. Gao

Progress of the Engineering Design for the HT-7U Steady-State Superconducting Tokamak

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.

The Superconducting Magnets for EAST Tokamak

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.

Plasma facing conponents of EAST

EAST plasma facing components (PFCs) have the function of protecting the vacuum vessel, heating systems and diagnostic components from the plasma particles and heat loads, and also additional to this particles and heat loads handling. They are installed in the vacuum vessel together with in-vessel coils, cryopump and diagnostic components. The design, fabrication and assembly have been finished. The PFCs are designed up-down symmetry to accommodate with both double null and single null plasma configuration. All PFCs use graphite tile for plasma facing surfaces affixed to copper alloy heat sink. A special deep hole drilling technology was developed to drill cooling channels directly on heat sink for high efficient heat removal. All Heat sink are installed onto the base alignment rails through stainless steel supports. As the benchmark of assembly for PFCs, the base rails are installed and measured precise based on a new alignment method integrating the optical instruments and a mechanical template. And so is a mechanical check template for checking the surface of first wall. As indicated, all the first wall components were fabricated and assembled successfully and meet the design requirement for the plasma operation.

Winding technique for HT-7U Tokamak magnet coils

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 fabrication and preassembly of HT-7 Superconducting Tokamak

The HT-7 Superconducting Tokamak is being built in Hefei, China. It is a device reconstructed and upgraded from T-7 Tokamak of Russia. This project is a co-operating item of fusion research between Russia and China. The Research and Manufacturing Center (RMC), ASIPP is responsible for the fabrication and assembly of HT-7. The design of HT-7 was carried out by both of Russian and Chinese engineers in 1991, and the main parts of the machine were fabricated in 1992. The preassembly of HT-7 was completed in April, 1993. This paper emphasizes on the manufacturing steps necessary to produce the HT-7 and a series of technological methods to overcome the key techniques.

The R&D progress of the EAST (HT-7U) superconducting tokamak

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.