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 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.     

A Hybrid Heating Method for the HT-7U Coils during Vacuum-Pressure Impregnation

1. IntroductionA program to design and fabricate a tokamakphysics experimental device of HT-7U using superconducting coils has been undertaken by the institute of Plasma Physics, the Chinese Academy of Sciences .The aim of HT-7U is to develop a scientificbasis for a low--cost, clean and continuously operating TOkamak fusion reator. The HT-7U magnet system consists of 16 magnets in the toroidal field (TF)system and 6 pairs of magnets in the poloidal field(PF) system located symmetricall…     

How should we test the ITER TF coils?

The international thermonuclear experimental reactor (ITER) toroidal field (TF) magnet system consists of 18 superconducting coils using a 68 kA Nb₃Sn conductor. In order to guarantee the performances of these coils prior to their installation, the test of at least one prototype coil at liquid helium temperature and full current is required. The test of all coils in the two-coil test configuration, with successive charging of each coil to nominal current is recommended. This requires a large test facility.     

Present Status of the KSTAR Superconducting Magnet System Development

The mission of Korea Superconducting Tokamak Advanced Research (KSTAR) project is to develop an advanced steady-state superconducting tokamak for establishing a scientific and technological basis for an attractive fusion reactor. Because one of the KSTAR mission is to achieve a steady-state operation, the use of superconducting coils is an obvious choice for the magnet system. The KSTAR superconducting magnet system consists of 16 Toroidal Field (TF) coils and 14 Poloidal Field (PF) coils. Internally-cooled Cable-In-Conduit Conductors (CICC) are put into use in both the TF and PF coil systems. The TF coil system provides a field of 3.5 T at the plasma center and the PF coil system is able to provide a flux swing of 17 V-sec. The major achievement in KSTAR magnet-system development includes the development of CICC,the development of a full-size TF model coil, the development of a coil system for background magnetic-field generation , the construction of a large-scale superconducting magnet and CICC test facility. TF and PF coils are in the stage of fabrication to pave the way for the scheduled completion of KSTAR by the end of 2006.     

Feasibility Study on Welding Structure of the HT-7U Toroidal Field Coil Case

1. IntroductionThe TF magnet system of HT-7U consists of 16coils arrayed toroidally and spaced 22.5" apart oneanother. The designed TF coils provide a magneticfield, which is necessary to maintain plasma in atokamak configuration with a current up to I MA.The TF coils have a D--shaped geometry of 3.5 meters high and 2.5 meters wide, producing a magneticfield of 3.5 T on 1.7 m of main radius. The TF coilcases, enclosing the TF winding packs, are the mainstructural components of the magn…     

Structural design of cryogenic component support for EAST （HT-7U）

EAST is a full superconducting tokamak with an elongated plasma cross-section. It consists of superconducting poloidal field (PF) magnet system, toroidal field (TF) magnet system, vacuum vessel with inner parts, thermal shields and cryostat vessel. The mission of the project is to widely investigate both physics and technologies of advanced tokamak operations, especially the mechanism of power and particle handling for steady-state operations. The cryogenic component is mainly composed of superconducting TF and superconducting PF coils that ensure the ability of sustaining magnetic field for plasma confinement, control and shaping in steady-state. This report describes the process of the structure design of cryogenic component support for EAST.     

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.     

The Alignment and Assembly for EAST Tokamak Device

EAST （HT-7U） is a large fusion experimental device. It is a full superconducting tokamak with 1 MA of plasma current, 1000 s of plasma duration, high elongation and triangularity. It mainly consists of superconducting magnets of poloidal and toroidal field （PF ＆ TF）, vacuum vessel （VV）, thermal radiation shield （TRS） and cryostat vessel （CV）. The significant difficulty for assembly of EAST is tight installation tolerances, which are in the order of several tenth of a millimeter. In particular, the alignment of plasma facing components to the magnetic axis of the device is less than ±0.5 mm. At present, a reasonable assembly process of EAST has been defined, and based on it, the alignment method for EAST, including the survey control network, the location of the main components in different directions, the magnetic axis determination and the accurate positioning of the plasma facing components inside of the vacuum vessel and so on, has been defined by using the sophisticated optical metrology system （SOMS）. This paper describes the assembly procedure of EAST and the installation tolerances associated with the main components. Meanwhile, how to establish the assembly survey control network, magnetic axis determination methods, are introduced in detail.     

Design of a Wedge-Shaped Toroidal Field Winding for KTX Device

The KTX device is a reversed field pinch(RFP)device currently under construction.Its maximum plasma current is designed as 1 MA with a discharge time longer than 100 ms.Its major radius is 1.4 m and its minor radius is 0.55 m.One of the most important problems in the magnet system design is how to reduce the TF magnetic field ripple and error field.A new wedgeshaped TF coil is put forward for the KTX device and its electromagnetic properties are compared with those of rectangular-shaped TF coils.The error field Bn/Btof wedge-shaped TF coils with6.4 degrees is about 6%as compared with 8%in the case of a rectangular-shaped TF coil.Besides,the wedge-shaped TF coils have a lower magnetic field ripple at the edge of the plasma region,which is smaller than 7.5%at R=1.83 m and 2%at R=1.07 m.This means that the tokamak operation mode may be feasible for this device when the plasma area becomes smaller,because the maximum ripple in the plasma area of the tokamak model is always required to be smaller than 0.4%.Detailed analysis of the results shows that the structure of the wedged-shape TF coil is reliable.It can serve as a reference for TF coil design of small aspect ratio RFPs or similar torus devices.     

Oclator: The one coil low aspect toroidal reactor

A new approach to construct a tokamak-type reactor(s) is presented here. Basically, the return conductors of toroidal field coils are eliminated and the toroidal field coil is replaced by one single large coil, around which will be placed several tokamaks or other toroidal devices. The elimination of return conductors should, in addition to other advantages, improve the accessibility and maintainability of the tokamaks, and offer a possible alternative to the search for special materials to withstand large neutron wall loading, as the frequency of changeover would be increased due to minimum downtime. It also makes it possible to have a low aspect ratio tokamak, which should improve the limit, so that a low toroidal magnetic field strength might be acceptable, meaning that NbTi superconducting wire could be used. This system is named OCLATOR (One Coil Low Aspect Toroidal Reactor).     

Conceptual Design of Alborz Tokamak Poloidal Coils System

The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. One of the most important parts of tokamak design is the design of the poloidal field system. This part includes the numbers, individual position, currents and number of coil turns of the magnetic field coils. Circular cross section tokamaks have Vertical Field system but since the elongation and triangularity of plasma cross section shaping are important in improving the plasma performance and stability, the poloidal field coils are designed to have a shaped plasma configuration. In this paper the design of vertical field system and the magnetohydrodynamic equilibrium of axisymmetric plasma, as given by the Grad-Shafranov equation will be discussed. The poloidal field coils system consists of 12 circular coils located symmetrically about the equator plane, six inner PF coils and six outer PF coils. Six outer poloidal field coils (PF) are located outside of the toroidal field coils (TF), and six inner poloidal field coils are wound on the inner legs and are located outside of a vacuum vessel.     

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.     

Preparation of the manufacture of the ITER correction coils

The ITER correction coils (CC) include three sets of six coils each, distributed symmetrically around the tokamak and inserted between the toroidal field (TF) and the poloidal field (PF) coils. Each pair of coils located on opposite sides with respect to the plasma is series connected with polarity such to produce asymmetric fields. These superconducting coils use a cable-in-conduit conductor, insulated, wound into multiple pancakes and inserted inside an austenitic stainless steel case. The requirements and the main features of the design are presented and the selected options reviewed in terms of their criticality in achieving the specified tolerances. The requested qualification trials are identified and reports the results obtained so far.     

A coil test facility for the cryogenic tests of the JT-60SA TF coils

In the framework of the Broader Approach Activities, the EU will deliver to Japan the 18 superconducting coils, which constitute the JT-60SA Toroidal field magnet. These 18 coils, manufactured by France and Italy, will be cold tested before shipping to Japan. For this purpose, the European Joint Undertaking for ITER, the Development of Fusion Energy (“Fusion for Energy”, F4E) and the European Voluntary Contributors are collaborating to design and set-up a coil test facility (CTF) and to perform the acceptance test of the 18 JT-60SA Toroidal Field (TF) coils. The test facility is designed to test one coil at a time at nominal current and cryogenic temperature. The test of the first coil of each manufacturer includes a quench triggered by increasing the temperature.The project is presently in the detailed design phase.     

Construction of a 2kW／4K Helium Refrigerator for HT—7U

Superconducting magnets of toroidal field (TF) and poloidal field(PF) of HT-7U tokamak are all made of NbTi/Cu Cable-in-Conduit Conductor (ClCC),and cooled with a forced flow supercritical helium of 3.8K.A helium refrigerator with an equivalent capacity of 2kW/r K will be constructed.This paper presents the design of the helium refrigerator process.The thermodynamics of the refrigeration cycle and the refrigerator equipments.     

Investigation of Effects of Toroidal Field Ripple on Plasma Poloidal Beta in IR-T1 Tokamak

We present an investigation of effect of Toroidal Field (TF) ripple (due to finite number of the toroidal field coils) on the plasma poloidal Beta in IR-T1 Tokamak. For this purpose, array of magnetic probes and also a diamagnetic loop with its compensation coil were designed, constructed, and installed on the outer surface of IR-T1. Amplitude of the TF ripple is obtained 0.01, and also the effect of the TF ripple on the poloidal Beta discussed. In the high field side region of tokamak chamber, the TF ripple effect is decreasing of the poloidal Beta, whereas the low field side has inverse situation.     

A Study on Main Breaker for Quench Protection of HT-7U Toroidal Superconducting Tokamak

This paper proposes a quench protection project of HT-7U toroidal superconductingtokamak through a forced commutation analysis of DC circuit breaker (DCCB) paralleling fuse.Based on the requirement of quench protection, main parameters are selected. Experimentalresults demonstrate the validity of this proposed project.     

Research and Development of Superconducting Magnet for Fusion Power

This paper describes mainly technological achievment of superconducting magnet for fusion power for the latest 10 years in Japan. The magnet development had been devoted to tokamak Fusion Experimental Reactor (FER). The major results obtained up to now are as follows.

In toroidal coil program, 12 T field generation, which is requested in a reactor toroidal coil, was realized with 6 kA multifilamentary Nb3Sn conductor in 1 m bore. For scaling-up of toroidal coil, half size coils of FER, LCT coils were tested up to 9 T.

In poloidal coil program, Demo Poloidal Coil project is now under way and coil testing will be started in spring of 1989. The stored energy of this coil is around 40 MJ.

In cryogenic technology program, fabrication and operation of large helium refrigerator technology were well established. A supercritical helium pump of 500 g/s was tested for forced flow coil.     

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.