Thermal Spray Coating of Tungsten for Tokamak Device

Thermal spray, such as direct current （d.c.） plasma spray or radio frequency induced plasma spray, was used to deposit tungsten coatings on the copper electrodes of a tokamak device. The tungsten coating on the outer surface of one copper electrode was formed directly through d.c. plasma spraying of fine tungsten powder. The tungsten coating/lining on the inner surface of another copper electrode could be formed indirectly through induced plasma spraying of coarse tungsten powder. Scanning electron microscopy （SEM） was used to examine the cross section and the interface of the tungsten coating. Energy Dispersive Analysis of X-ray （EDAX） was used to analyze the metallic elements attached to a separated interface. The influence of the particle size of the tungsten powder on the density, cracking behavior and adhesion of the coating is discussed. It is found that the coarse tungsten powder with the particle size of 45-75μm can be melted and the coating can be formed only by using induced plasma. The coating deposited from the coarse powder has much higher cohesive strength, adhesive strength and crack resistance than the coating made from the fine powder with a particle size of 5μm.

The Design of Circuit for Checking Short in HT-7U Superconducting Tokamak Device

The tokamak HT-7U project has been funded as a Chinese national project since 1998. The main object of the project is to build a nuclear fusion experimental device with divertor configuration, which is designed by the Institute of Plasma Physics, the Chinese Academy of Sciences (ASIPP). It is a full superconducting device, consisting of superconducting toroidal field (TF) coils and superconducting poloidal field (PF) coil. During the operation of the device, the operational parameter of device should be checked by technical diagnosis. This paper describes the design of circuit for checldng short between every two parts of the HT7U device. The main contents of design include circuit of data acquisition and data processing of computer.

Double-sided friction stir welding of Nitronic-40 stainless steel for application in tokamak devices

Nitrogen containing austenitic stainless steels(N-ASS)are widely utilized to fabricate various structural components in tokamak type fusion reactors owing to their suitable mechanical and functional proper-ties.These components are exposed to a range of temperatures(4-500 K)and interact closely with the magnetic fields that are used to control and contain the plasma within the tokamak systems.Nitronic-40(N40)or XM-11 stainless steel is one such N-ASS used for fabricating structural components in the mag-netic and vacuum vessel systems in tokamak devices.Fabrication of most of the larger components in the magnetic and vacuum vessel systems typically involves some type of fusion-based welding process.This study presents a double-sided friction stir welding(FSW)approach as an alternative to fusion welding processes to join 12 mm thick N40 plates to obtain joints with a low fraction ofδferrite(a detrimen-tal ferromagnetic phase),high joint efficiency,no sensitization and loss of hardness in the heat affected zone,and minimal nitrogen desorption from the weld nugget.The double-sided FSW approach yielded superior weldments when compared to similar joints accomplished by fusion welding for application in tokamak devices.

An approach to explore the eddy currents of the new type divertor for EAST device using ANSYS code

An effective method for eddy current calculation has been developed for EAST’s new divertor by using ANSYS.A 3D model of a double null divertor for the EAST device was built to evaluate eddy currents and electromagnetic(EM)forces on these components.The main input to the model is the plasma current and poloidal field coil currents,which are loaded into the model using experimental data measured from the EAST discharges.These currents generate magnetic fields that match those producing an EAST discharge,and the time variation of these fields produces the eddy currents in the divertors,along with from the resulting EM forces.In addition,the first 10 time steps were discussed for the eddy current generation and changing trend.It indicates that a static analysis before a transient mode start can solve the eddy current origination in the initial time steps.With this method,the EM transient response of EAST’s new divertor can be predicted based on ANSYS simulations.Furthermore,the method is also an effective approach to estimate the EM results for the in-vessel components of a fusion reactor during a disruption.