Weng Pei-de

First Engineering Commissioning of EAST Tokamak

Experimental Advanced Superconducting Tokamak (EAST) is the first fully superconducting tokamak. The first commissioning started on Feb. 1st of 2006 and finished on March 30th of 2006 at the Institute of Plasma Physics, Chinese Academy of Sciences. It consists of leakage testing at both room temperature and low temperature, pumping down, cooling down all coils, current leads, bus bar and the thermal shielding, exciting all the coils, measuring magnetic configuration and warming up the magnets. The electromagnetic, thermal hydraulic and mechanical performance of EAST Toroidal Field (TF) and Poloidal Field (PF) magnets have also been tested. All sub-systems, including pumping system, cryogenic system, PF& TF power supply systems, magnet instrumentation system, quench detection and protection system, water cooling system, data acquisition system, main control system, plasma control system (PCS), interlock and safety system have been successfully tested.

Design, Analysis and R&D of the EAST In-Vessel Components

In-vessel components are important parts of the EAST superconducting tokamak. They include the plasma facing components, passive plates, cryo-pumps, in-vessel coils, etc. The structural design, analysis and related R&D have been completed. The divertor is designed in an up-down symmetric configuration to accommodate both double null and single null plasma operation. Passive plates are used for plasma movement control. In-vessel coils are used for the active control of plasma vertical movements. Each cryo-pump can provide an approximately 45 m3/s pumping rate at a pressure of 10−1 Pa for particle exhaust. Analysis shows that, when a plasma current of 1 MA disrupts in 3 ms, the EM loads caused by the eddy current and the halo current in a vertical displacement event (VDE) will not generate an unacceptable stress on the divertor structure. The bolted divertor thermal structure with an active cooling system can sustain a load of 2 MW/m2 up to a 60 s operation if the plasma facing surface temperature is limited to 1500 °C. Thermal testing and structural optimization testing were conducted to demonstrate the analysis results.

Static and Dynamic Mechanical Analyses for the Vacuum Vessel of EAST Superconducting Tokamak Device

EAST (experimental advanced superconducting tokamak) is an advanced steady-state plasma physics experimental device, which is being constructed as the Chinese National Nuclear Fusion Research Project. During the plasma operation the vacuum vessel as one of the key component will withstand the electromagnetic force due to the plasma disruption, the Halo current and the toroidal field coil quench, the pressure of boride water and the thermal load due to 250 oC baking by pressurized nitrogen gas. In this paper a report of the static and dynamic mechanical analyses of the vacuum vessel is made. Firstly the applied loads on the vacuum vessel were given and the static stress distribution under the gravitational loads, the pressure loads, the electromagnetic loads and thermal loads were investigated. Then a series of primary dynamic, buckling and fatigue life analyses were performed to predict the structures dynamic behavior. A seismic analysis was also conducted.

The Test Facility for the EAST Superconducting Magnets

A large facility for testing superconducting magnets has been in operation at the Institute of Plasma Physics of the Chinese Academy of Sciences since the completion of its construction that began in 1999. A helium refrigerator is used to cool the magnets and liquefy helium which can provide 3.8 K-4.5 K, 1.8 bar-5 bar, 20 g/s-40 g/s supercritical helium for the coils or a 150 L/h liquefying helium capacity. Other major parts include a large vacuum vessel (3.5 m in diameter and 6.1 m in height) with a liquid nitrogen temperature shield, two pairs of current lead, three sets of 14.5 kA-50 kA power supply with a fast dump quench protection circuitry, a data acquisition and control system, a vacuum pumping system, and a gas tightness inspecting devise. The primary goal of the test facility is to test the EAST TF and PF magnets in relation to their electromagnetic, stability, thermal, hydraulic, and mechanical performance. The construction of this facility was completed in 2002, followed by a series of systematic coil testing. By now ten TF magnets, a central solenoid model coil, a central solenoid prototype coil, and a model coil of the PF large coil have been successfully tested in the facility.

The HT-7U Project and its preliminary engineering design

The HT-7U superconducting tokamak is an advanced steady state plasma experimental device to be built in the Institute of Plasma Physics, the Chinese Academy of Sciences (ASIPP). The mission of the HT-7U Project is to develop scientific issues on the sustainment of a nonburning plasma scenario for the steady-state operation and engineering issues on establishing technology basis of superconducting tokamak. HT-7U will have a long pulse (60-1000 s) capability, a flexible poloidal field system, and auxiliary heating and current drive systems, and will be able to accommodate divertor heat loads that make it an attractive test. The engineering conceptual design and the R and D program for prototypes of HT-7U has begun in 1995. The design of HT-7U incorporates superconducting magnet systems of the toroidal field (TF) and poloidal field (PF). In the beginning of this year, some developments in tokamak design have been made. The R&D programs, including prototype coils of TF and central solenoid (CS), the coordinate test facilities, are being readied. In this paper, an overview of the HT-7U Project and a brief introduction of the general tokamak system with the emphasis on the features structures of TF system, PF system, vacuum vessel, radiation shields and cryostat vessel are provided. The R&D program status are introduced, also.

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.

Temperature Field and Thermal Stress Analysis of the HT-7U Vacuum Vessel

The HT-7U vacuum vessel is an all-metal-welded double-wall interconnected with toroidal and poloidal stiffening ribs. The channels formed between the ribs and walls are filled with boride water as a nuclear shielding. On the vessel surface facing the plasma are installed cable-based Ohmic heaters. Prior to plasma operation the vessel is to be baked out and discharge cleaned at about 250°C. During baking out the non-uniformity of temperature distribution on the vacuum vessel will bring about serious thermal stress that can damage the vessel. In order to determine and optimize the design of the HT-7U vacuum vessel, a three-dimensional finite element model was performed to analyse its temperature field and thermal stress. The maximal thermal stress appeared on the round of lower vertical port and maximal deformation located just on the region between the upper vertical port and the horizontal port. The results show that the reinforced structure has a good capability of withstanding the thermal loads.

The Analysis and Calculation for the Toroidal Magnetic Field of HT-7U

The HT-7U super-conducting tokamak is a full super-conducting magnetically confined fusion device, It mainly consists of super-conducting toroidal field (TF) coils and super conducting poloidal field (PF) coils. This paper describes the distribution of magnetic field, ripple and electromagnetic loads of TF system, some results are necessary to analyze and calculate the stresses and deformation on TF system by a finite element method. Meanwhile, in this paper, the main scope of the calculation is carried out for the case of constant magnetic field on conductor of the TF coil winding in order to provide electromagnet parameters for the quench analysis of Cable-in-Conduit Conductor (CICC) of TF system in HT-7U.

FE Stress Analysis of the Toroidal Field Magnets of HT-7U

In this paper, the mechanical strength of the toroidal field (TF) magnets of HT-7U with the electromagnetic force in different magnetic fields is emphatically analyzed by means of finite element method. The model and feasible method of computation are put forward. Some important conclusions are made available for reference in the design and construction of TF for HT-7U.

Research of the CICC Stability by the Numerical Code Gandalf

The basic approach to computer analysis of the CICC in superconducting Tokamak HT-7U is given and discussed. We will apply a 1-D mathematical model (Gandalf) to investigate the stability of CICC at real operating modes of Tokamak. 1-D model can be directly adopted to follow the evolution of the zone when the energy input is large enough and the coil quenches. In this report, we will analyze the stability of typical CICC (including pure copper) and discuss effect of copper on the stability of CICC.