[en] The HT-7U superconducting tokamak is under construction. It is expected to complete the fabrication and installation in the next 2-3 years. Overall configuration of the device is described in other paper. The vacuum vessel is full welded with 'D' shaped cross-section and double wall. It consists of 16 segments and supported by eight flexible multiple plate supports. Ports are distributed at top, bottom and middle plan. Low stiffness supports, and two sections of bellows on each port allow the vacuum vessel can move a little in radial direction to accommodate the expansion during the vessel bake-out to 250 deg. C. Some R and D of the vacuum vessel is completed, and fabrication is in progress

[en] The vacuum vessel of the HT-7U superconducting tokamak has a full welded structure with double wall which will be filled with borated water is used for neutron shielding, non-circular cross-section is used for plasma elongating, horizontal and vertical ports are used 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 deg. C and plasma facing components (PFCs) baked out at 350 deg. C to get a cleaning wall. When the machine is in operation, hot wall (vacuum vessel wall is around 100 deg. C and first wall is around 150 deg. C) and cold wall (vacuum vessel wall and first wall is in normal equilibrium temperature) are both considered. The stress caused by thermal deformation and electromagnetic (EM) loads caused by 1.5 MA plasma disrupt in 3.5 T magnetic field are important in the design of the HT-7U vacuum vessel and PFCs. Finite element method was accepted for structure analysis of the vacuum vessel

[en] DSD (Double seal door) design concept was introduced. 3-D model work was performed for DSD in the three typical regions, such as upper port, equatorial port, divertor port. The numerical analysis for some typical components was done based on Finite Element (FE) method by using ANSYS code, especially for the optimization activities. The rescue procedures of the DSD was discussed which could benefit a little for future engineering implementation. The design and analysis work can support and be the important reference for future procurement. (authors)

[en] In-vessel components are important parts of EAST superconducting tokamak. It is include plasma facing components, passive plates, cryo-pumps, in-vessel coils, etc. The structure optimized design, analyses and related R and D has been complete. Divertor is up down symmetric to accommodate both double null and single null operation. Passive plates can supply around 100 ms time constant during plasma vertical movements. In-vessel coils are used for plasma vertical movements active control. Each cryo-pumps can supply around 45 m³/s pumping speed for particle exhaust. Analyses shows when 1 MA plasma current disrupt in 3 ms EM loads caused by halo current (I_halo 25%I_p, TPF=2) will not bring unacceptable stress on divertor structure. Graphite tiles bolted to heat sink. Cooling channel consist of holes on heat sink. The bolted divertor thermal structure can sustain 2 MW/m² up to 60 s operation if we limited first surface temperature 1500 deg C. Thermal testing and structure optimised testing have been made to demonstrate the analyses result. All the in vessel components are under fabrication and components for first plasma will be complete around July 2006. (author)

[en] The HT-7U superconducting Tokamak device is in the stage of initial engineering design. As the device is a complicated mechanical system, many mechanical analyses must be performed so as to optimize the design. During the design of the HT-7U device, we have made some static structural analyses on some key parts of HT-7U. The finite element method was mainly used in the analyses and the analyses were very helpful for our engineering design. (authors)

[en] The first wall at inner part of the vacuum vessel in EAST is one of the plasma facing components (PFCs), which needs to be cooled during plasma operation and baked during wall conditioning. Water and hot nitrogen gas, respectively, will be used during cooling and baking of the first wall. The thermal analysis is done by software FLUENT. Compared with two arrangements of tubes in heat sink, the result is helpful to the design of PFCs' structure. (authors)

[en] SOLPS-ITER is used for modeling divertor plasmas in the experimental shot on the EAST tokamak with a carbon divertor. The modeling uses real divertor geometry and an MHD equilibrium from the shot. The modelled divertor configurations include upper single null (USN), disconnected double null (DDN) and lower single null (LSN). D⁰, D⁺¹, e ⁻ and carbon impurity species sputtered from the first wall are included in the multi-fluid simulations. Neutral particles are traced with EIRENE in which neutral–neutral collisions in triangular grids are considered. The default physics models in the code are employed. The cross-field transport coefficients D, and deduced from the profile measurements at the targets are used with the assumption of , some of the deduced D are modified in the SOL and private regions and D, , in the core are adjusted in order to match better the experimental measurement at the target plates. Employed D, and drop from the core to the SOL, they may drop to neoclassic level within the edge transport barriers (ETBs). A ballooning effect on the transport coefficients is introduced. The computational contour profiles of electron temperature in the divertor regions and the computational profiles of plasma pressure and electron temperature at the mid-plane (the upstream) and at the target plate show the sheath limited (low recycling) operation regime. The calculated profiles of electron density n _e, electron temperature T _e, particle flux f , parallel ion saturation current and parallel power flux P along the target plates of the divertor are compared to the experimental measurements. Using the deduced and adjusted cross-field transport coefficients, when the measurement errors are considered the calculated and measured profiles have a relatively good match, the computational profiles are similar to the measured profiles, the profiles peaked at the separatrix strike points. The most peak values f _s of calculated profiles are in the range of the measured error-bars , only in several matched profiles f _s is out of the range of the error-bar, but, relative error does not exceed . When full drifts are taken into account in the entire SOL and the private flux regions the calculated profiles have better match to the measured profiles and the modeling results show the drifts enhance the divertor asymmetry. (paper)

[en] The EAST upper divertor was updated to tungsten divertor in 2014, but the lower divertor was still left carbon divertor. The upper divertor heat load exhausting capability can be up to 10MW/m² , but only 2MW/m² for the lower divertor. To meet the device high performance plasma operation and high plasma heating power. The lower divertor planed to be updated to tungsten divertor. The lower tungsten divertor concept design started from 2016. Considering plasma configurations, neuter particles exhausting, field expansion, cost reduction and so on, geometry of lower divertor was optimized and will be different from upper tungsten divertor geometry.

[en] To operate divertor experiment, an in-vessel cryopump was installed on the EAST tokamak in 2008. It can limit gas impurity recycling from divertor region into core plasma area, and provide plasma density control with toroidally distributed high pumping speed. In this paper, the designing and manufacturing is basically described. Most parts are manufactured in ASIPP, except for some procedures such as laser cutting, plasma-spray coating, and pipe annealing. For this first in-vessel cryopump, liquid helium and nitrogen supplying system is upgraded. Functional tests for this cryopump show a good radiation shield and pumping capability. A campaign utilizing this device for divertor physics research has been successful.

[en] In-vessel components are important parts of EAST superconducting tokamak. It is include plasma facing components, passive plates, cryo-pumps, in-vessel coils, etc. The structure design, analyses and related R and D had been completed. Divertor is designed as up down symmetric to accommodate both double null and single null plasma operation. Passive plates can supply around 100ms time constant during plasma vertical movements. In-vessel coils are used for plasma vertical movements active control. Each cryo-pumps can supply around 45m³/s pumping speed for particle exhaust. Analyses shows when 1MA plasma current disrupt in 3ms EM loads caused by halo current will not bring unacceptable stress on divertor structure. Bolted divertor thermal structure can sustain 2MW/m² up to 60s operation if plasma facing surface temperature is limited to 1500 deg. C. Thermal testing and structure optimised testing have been made to demonstrate the analyses result. (author)