Thermal Hydraulic Research Of Internal Components And Fluid Transmission And Distribution Network For Fusion Device

Tokamak is the most promising fusion reactor for the development and utilization of fusion energy.Due to the fusion of deuterium and tritium under magnetic confinement should be produced at 100 million centigrade,the components other than blanket in fusion reactor need to bear hundreds of MW order heat load,so the thermal hydraulic system is one of the core systems related to the safe operation of fusion device.According to the characteristics of material,structure,welding and installation,there are both strict requirements for the heat load and temperature control of fusion device internal components,during fusion device operating.In order to provide reference and experience for the thermal hydraulic research and engineering design of fusion device,based on the EAST tokamak experimental device,the thermal hydraulic analysis method and measurement method suitable for the engineering structure characteristics of fusion device are selected to carry out the integrated and systematic thermal hydraulic research for the internal components and fluid transmission and distribution network of fusion experiment device.Based on the theory of flow and heat transfer,the internal component cooling circuit and the external fluid transmission and distribution network thermal hydraulic analysis method(FRTHAM)is constructed for the fusion experimental device.The EAST thermal hydraulic measurement system is built.The thermal hydraulic experiment and performance analysis of the internal components and the fluid transmission and distribution network are carried out base on them,after that a set of thermal hydraulic model and integrated thermal hydraulic system engineering design method for the internal components and fluid transmission and distribution pipe network are obtained,which are in good agreement with the experiment.These works can provide reference and experience for the thermal hydraulic system design of fusion device in the future.The divertor is the most important high heat load component in the magnetic confinement fusion device except the blanket.It needs to bear the non-nuclear high heat load during fusion device operation.The relationship between the temperature of the target plate and the heat load is simulated and analyzed,under the condition of magnetic surface broadening in the new concept of divertor configuration which are proposed in recent ten years,and the heat flux carrying capacity of them are evaluated.The heat load carrying performance of the dynamic magnetic divertor configuration,the fishtail divertor(FTD),which is proposed by ASIPP,is simulated,and analyzed in detail.The Fishtail divertor is more suitable for future fusion reactor,for very narrow high-intensity heat flow,the heat flow swing mode can significantly reduce the temperature of divertor target plate.The simulation results show that the new technology of Fishtail Divertor has excellent ability to relieve the transient high heat load caused by the Edge Local Mode(ELM)under the condition of hundred Hertz heat flow swing.