Wu Weiyue

Prototype of the Superferric Dipoles for the Super-FRS of the FAIR-Project

The FAIR China Group (FCG), consisting of the Institute of Modern Physics (IMP Lanzhou), the Institute of Plasma Physics (ASIPP, Hefei) and the Institute of Electric Engineering (IEE, Beijing) developed and manufactured in cooperation with GSI, Germany a prototype of a superferric dipole for the Super-Fragment-Separator of the FAIR-project. The dipole magnets of the separator will have a deflection radius of 12.5 m, a field up to 1.6 T, a gap of at least 170 mm and an effective length of more than 2 meters to bend ion beams with a rigidity from 2 T · m up to 20 T · m. The magnets operate at DC mode. These requirements led to a superferric design with a yoke weight of more than 50 tons and a maximum stored energy of more than 400 kJ. The principles of yoke, coil and cryostat construction will be presented. We will also show first results of tests and measurements realized at ASIPP and at IMP.

Design of the Magnetic Field for HT-7U Tokamak

The HT-7U superconducting tokamak is an advanced steady-state plasma physics experimental device to be built in the Institute of Plasma Physics, the Chinese Academy of Sciences (IPP-CAS). The plasma current is 1 MA and the major and minor radius are 1.78 m and 0.4m respectively, with an elongation of 1.85. The preliminary engineering design of the poloidal field (PF) and toroidal field (TF) magnet systems have been done. The PF system is composed of twelve superconducting coils located symmetrically about the equator plane. The central solenoid (CS) assembly is formed by six coils. The TF system consists of 16 superconducting coils. The NbTi cable-in-conduit conductor or (CICC) cooled by a supercritical helium at 4.5 K is chosen as a superconductor for all of the PF and TF coils. At this temperature, the peak magnetic field on the PF magnets is about 4.51 T. The maximum volt-second capacity and the duration of plasma inductive discharge are about 10 Vs and 10 seconds respectively. The stray field in plasma initial region is quite low (? 1.5 ? 10-3 T). The magnetic field on the TF magnet is 5.8 T while the toroidal field at the center of the device (R = 1.7 m) is 3.5 T and the ripple of the TF is less than 0.62% at the outer plasma surface (R = 2.1 m). All of the PF and TF magnets are stable during all modes of operation including the plasma disruption. The final design of the PF system is the result of an iterative process involving the use of equilibrium code EQT, magnetic code EFFI, and other codes, which have been developed by our designing group.

Design of the PF system for EAST(HT-7U) tokamak

EAST(HT-7U) superconducting tokamak is an advanced steady-state plasma physics experimental device. The poloidal field (PF) coil system is used with superconducting conductor. The NbTi cable-in-conduit conduct (CICC) cooled by supercritical helium at 4.5 K is chosen as superconductor for all of the PF magnets. It was consisted of fourteen superconducting coils, located symmetrically about the equator plane of the machine. The PF system was supported by the case of TF coils. The final engineering design of PF system will be report in this paper. The maximum capacity of the volt seconds for EAST(experimental advanced superconducting tokamak) PF system is about 10 Web and the peak magnetic field in PF coils is about 4.5 T.

Design of the poloidal field system and plasma equilibrium of HT-7U tokamak

Abstract HT-7U superconducting tokamak is an advanced steady-state plasma physics experimental device will be built in Hefei at the Chinese Academy of Sciences, Institute of Plasma Physics (CASIPP). The HT-7U device is a full superconducting Tokamak which means that the toroidal field (TF) coils system and poloidal field coils (PF) systems are used with superconducting conductor. There are fourteen superconducting PF coils, including central selenoid, located symmetrically about the equator plane. The selenoid is supported by the case of TF coils. The NbTi cable-in-conduit conduct (CICC) cooled by supercritical helium at 4.5 K is chosen as superconductor for all of the PF and TF coils. The plasma current is ∼1 MA and the duration of the plasma is ∼10 s for ohmic heating discharge. The preliminary engineering design of PF and the results of the calculating for plasma equilibrium will be reported in this paper. The maximum capacity of the magnetic flux for PF is ∼10 Web and the stray field in plasma initiate region is quite low. The peak magnetic field in PF coils is ∼4.5 T. Three type shapes of plasma can be chosen during the operating which are shown in the result of our calculations.

Preliminary Engineering Design of Toroidal Field Magnet System for Superconducting Tokamak HT-7U

HT-7U is a large fusion experimental device. It will be built in 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 a 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 and stress analysis, heat load evaluation are given.

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.

Development of Experimental Superconducting Magnet for the Collector Ring of FAIR Project

A pool cooled experimental magnet based on the copper stabilized NbTi superconducting wire was designed, fabricated and tested, in order to evaluate the engineering design of the dipole superconducting magnet for the collector ring (CR) of the facility for antiproton and ion research (FAIR) project. In this paper, the experimental setup including quench protection system was presented. Performance of the liquid helium pool cooled test was introduced. All of the results indicate both the performance of conductor and the experimental superconducting magnet under low temperature is stable, which suggests the engineering design are feasible for the formal magnet in CR of the FAIR project.

An Optimized Structure Design of the HT-7U Vacuum Vessel

The vacuum vessel of the HT-7U superconducting tokamak will be a fully-welded structure with a double-wall. The space between the double-wall will be filled with borated water for neutron shielding. Non-circular cross-section is designed for plasma elongating. Horizontal and vertical ports are designed for diagnosing, vacuum pumping, plasma heating and plasma current driving, etc. The vacuum vessel consists of 16 segments. It will be baked out at 250°C to obtain a clean wall. When the machine is in operation, both the hot wall (the wall temperature is around 100°C) and the cold wall (wall temperature is in normal equilibrium) are considered. The stress caused by thermal deformation and the electromagnetic (EM) loads caused by 1.5 MA plasma disruption in 3.5 T magnetic field have to be taken into account in the design of the HT-7U vacuum vessel. Finite element method was employed for structure analysis of the vacuum vessel.

Collaborative Simulation and Testing of the Superconducting Dipole Prototype Magnet for the FAIR Project

The superconducting dipole prototype magnet of the collector ring for the Facility for Antiproton and Ion Research (FAIR) is an international cooperation project. The collaborative simulation and testing of the developed prototype magnet is presented in this paper. To evaluate the mechanical strength of the coil case during quench, a 3-dimensional (3D) electromagnetic (EM) model was developed based on the solid97 magnetic vector element in the ANSYS commercial software, which includes the air region, coil and yoke. EM analysis was carried out with a peak operating current at 278 A. Then, the solid97 element was transferred into the solid185 element, the coupled analysis was switched from electromagnetic to structural, and the finite element model for the coil case and glass-fiber reinforced composite (G10) spacers was established by the ANSYS Parametric Design Language based on the 3D model from the CATIA V5 software. However, to simulate the friction characteristics inside the coil case, the conta173 surface-to-surface contact element was established. The results for the coil case and G10 spacers show that they are safe and have sufficient strength, on the basis of testing in discharge and quench scenarios.

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 composite 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.