P.D. Weng

The engineering design of the HT-7U tokamak

The HT-7U tokamak is a Chinese National Meg-science Engineering Project designed to develop scientific and technological basis for the tokamak fusion reactor. HT-7U has a long pulse (1000 s) operating capability, flexible PF system and auxiliary heating and current drive system making it a good facility for advanced performance experiments. This paper provides an overview of the HT-7U project and describes the engineering design of the tokamak and sub-systems.

HT-7U TF and PF conductor design

Abstract The HT-7U tokamak is a Chinese National Project. The mission of the project is to build an advanced full superconducting tokamak. NbTi strands are selected as the superconducting material. The copper ratio of the strands is 3.5 times less than that required by optimization; therefore, increasing the copper fraction in the cable is a key issue for conductor stability. Several versions of CICC with different configurations and surface treatment have been investigated and experimented with. Using segregated copper is a cost-saving option available to the HT-7U project.

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 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.

Reliability of Force Cooled Superconducting Magnets for Fusion

The latest superconducting magnets (SM) for fusion are mostly force-cooled, mainly because it allows reliable electrical insulation of the coils using vacuum pressure impregnation (VPI). SM of this type have many leads, feeders and coolant tubes, located in cryostat vacuum, which must sustain high voltages, induced on them by fast current changes. However vacuum loss can spoil their insulation. A few such cases occurred during the T-15 tokamak coils testing, initially having bare leads relying upon vacuum. But its loss generated a coil quench, a protecting current dump at high voltage, followed by breakdown and arc. Even leads insulation by Teflon and fiberglass tape wrap proved to be insufficient. Nevertheless, similar tape wrap insulation of leads and feeders (ILF) was used in EAST, KSTAR, SST-1 and W-7X. So far, seven breakdowns occurred during their coil tests at operating voltage ~<;3 kV. Breakdowns never initiated in the coils, but always on their leads, feeders and sensor lines, indicating that their insulation made by tape wrap were too weak. Instead of ILF improvement some projects undertake Paschen tests. These are planned as the baseline for ITER too. But these tests are valid for the coil with open insulated surface, but are not appropriate for the final tests, when insulation should not be exposed to vacuum. Up to now ILF final tests have been done in all devices at 10-21 kV, but only in good vacuum in spite of the fact that such tests could not guarantee safe operation in case of vacuum loss. We propose to increase ILF strength to the same level, as in the coils, using vacuum-tight grounded stainless steel casings filled up by VPI over magnet leads. This will provide reliable and easily testable solid insulation. Besides, casings would exclude He leaks, providing the second vacuum tight barrier over the ILF. Thus it would increase the magnet reliability and would make it possible to avoid the needs of all single coils test.

Operational region and sawteeth oscillation in the EAST tokamak

The first plasma discharges were successfully achieved on the experimental advanced superconducting tokamak (EAST) in 2006. The sawteeth behaviours were observed by means of soft x-ray diagnostics and ECE signals in the EAST. The displacement and radius of the q = 1 surface was studied and compared with the result of equilibrium calculation. The density sawtooth oscillation was also observed by the HCN laser interferometer diagnostics. The structure of the EAST operational region was studied in detail. Plasma performance was obviously improved by the boronization and wall conditioning. It was observed that lower qa and a wider stable operating region is extended by the GDC boronization.

Spring-back simulation of sheet metal forming for the HT-7U vacuum vessel

Abstract The HT-7U vacuum vessel is an all-metal welded double wall toroidal structure, which has characteristics of ultra-high vacuum and thin shell. Some of the forming tools will be fabricated according to its outline dimensions. In order to design a optimal shape for the forming tools and avoid an undesirable local deformation at the head of sheet, the software package ansys/ls-dyna and dynaform were utilized to simulate the process of the deep drawing in sheet metal forming and develop a parameter optimization system for the design of the die and punch. During the numerical simulation the updated LaGrange formulation and elastic–plastic constitutive equation were adopted to solve the problem of large strain and large deformation in sheet forming process. According to the simulation analysis results the optimum shape of the die and punch surface was finally determined.

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 modifications of conductor design for HT-7U TF magnets

The Tokamak HT-7U project has been funded as a Chinese national project since 1998. The machine is designed by the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). The main object of the project is to build a nuclear fusion experimental device with divertor configuration. It is a fully superconducting device, consisting of superconducting toroidal field (TF) coils and super-conducting poloidal field (PF) coils. This paper describes the design of TF conductors, including pool boiling cooled conductor (PBCC) and cable-in-conduit conductor (CICC). The CICC design is based on UNK NbTi wires made in Russia, cooled with supercritical helium. The other (PBCC) is based on SSC cables. Both versions can satisfy the design requirements. The modification of the conductor design including their main parameters, the configuration, and analysis of stability for both conductors are given.

Structural analyses and preliminary design of HT-7U cryostat

The cryostat of HT-7U Tokamak is a large vacuum vessel surrounding the entire Basic Machine with cylindrical shell, dished top and flat bottom. The main function of HT-7U cryostat provides the thermal barrier between the ambient temperature testing hall and the liquid helium cooled superconducting magnet. The loads applied to the cryostat are vacuum pressure, dead weight, seismic events and electromagnetic forces originated by eddy currents. It also provides feed through penetrations for all the connecting elements inside and outside the cryostat. The main material selected for the cryostat is stainless steel 304L. The structural analyses including buckling for the cryostat vessel under the plasma operation condition have been carried out using a finite element code. Stress analysis results show that the maximum stress intensity was below the allowable value. In this paper, we emphasize on the structural analyses of HT-7U cryostat. The preliminary design of the cryostat is also described.

Fabrication of the Toroidal Field Superconducting Coils for the EAST Device

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