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[en] The first dedicated divertor physics experiments were carried out on Experimental Advanced Superconducting Tokamak (EAST) in April 2009. Detailed measurements of plasma parameters at the divertor targets were made using an extensive array of divertor diagnostics, and some insights into several divertor physics issues were gained. This work was focused primarily on ohmic discharges. The behavior of divertor plasma, power asymmetry between inner and outer targets, and different fueling methods were investigated in both single null (SN) and double null (DN) divertor discharges. Divertor plasma detachment was achieved, for the first time on EAST, by ramping up the density during the discharge with well controlled, steady divertor configuration. The screening of carbon has been studied by puffing methane into the divertor. In addition, radiative divertor experiments were performed by localized argon injection, leading to a significant reduction in the heat flux on the target plates due to enhanced radiation in the divertor. This provides a means to avoid excess heat load on divertor components, specifically the divertor target tiles, which is essential for high power and long pulse operations, as envisioned for EAST.
[en] A new method, by using eigenmodes to reduce the fitting parameters and pre-calculated eddy current based on a lump parameter circuit equation, is applied to reconstruct the vacuum field for EAST plasma startup. (magnetically confined plasma)
[en] For low single-pass absorption of ion cyclotron range frequency (ICRF) wave in the EAST plasma cavity modes are expected to be excited between the low field side (LFS) antenna and the hybrid cut-off layer. The toroidal spectrum for D(H) minority heating scenarios in EAST is modeled by using FELICE (finite elements ion cyclotron emulator), a full wave code based on plane-stratified geometry. The excitation of cavity modes is studied. The methods for suppressing cavity modes are also discussed, to increase the efficiency of minority ion heating.
[en] Technical diagnosis system (TDS) is one of the important subsystems of EAST (experimental advanced superconducting tokamak) device, main function of which is to monitor status parameters in EAST device. Those status parameters include temperature of different positions of main components, resistance of each superconducting (SC) coils, joint resistance of SC coils and high-temperature superconducting (HTS) current leads, strain of cold-quality components endured force, and displacement and current of toroidal field (TF) coils in EAST device, which are analog input signals. In addition there are still some analog and digital output signals. The TDS monitors all of those signals in the period of EAST experiments. TDS data monitoring is described in detail for it plays important role during EAST campaign. And how to protect the SC magnet system during each plasma discharging is presented with data of temperature of coolant inlet and outlet of SC coils and feeders and cases of the TF coils and temperature in the upper and middle and bottom of the TF coil case. During construction of the TDS primary difficulties come from installation of Lakeshore Cernox temperature sensors, strain measurement of central solenoid coils support legs and installation of co-wound voltage sensors for quench detection. While during operation since the first commissioning big challenges are from temperature measurement changes in current leads and quench detection of PF coils. Those difficulties in both stages are introduced which are key to make the TDS reliable. Meanwhile analysis of experimental data like temperature as a back up to testify quench occurrence and stress on vacuum vessel thermal shield and vacuum vessel have also been discussed.
[en] Full text of publication follows. The scientific mission of EAST is to explore the reactor relevant regimes with long pulse lengths and high plasma core confinement and to develop and verify solutions for power exhaust and particle control under steady state operation. One of the main research goals of EAST involves long steady-state high-performance plasma pulse. It is a fusion engineering challenge to handle steady-state high heat flux form plasma. To accomplish this aim, 3 generation divertors have been designed for EAST to withstand high heat flux whose peak values are increasing rapidly. The first generation divertors were used on the initial phase of the plasma burning as there was no any cooling for divertors. The first generation divertors are just stainless plate 5 mm in thickness bolted on supports which had been applied since 2006 to 2007. From 2008 to 2013 the longest EAST pulse has reached 400 s and the peak heat flux on divertor exceed 2 MW/m2. The second generation divertors were used during this phase. The divertors consist of graphite tiles, heat sink and supports. The graphite tiles coated with SiC were mounted on heat sink. The heat sink has several cooling channels and cooling water was pumped in from pipes to remove thermal power. The third generation divertors are ITER-like W mono-block structure. They are going to be used in 2014. The newest design is expected to withstand the higher heat flux that is more than 10 MW/m2. Material W is a hard melt metal with large strength. The cooling CuCrZr pipe connects with W mono-blocks directly. Simulation and test proved that such structure have higher reliability and more duty time than the past design. We are making every effort to improve thermal extraction technology of divertor by comparing and practice different designs. The process of divertor heat transfer has been studied to define key design variables and explored the relationship between design variable and heat transfer efficiency. All such efforts made in EAST can bring experiences and answers for ITER or any next divertor fusion device on nuclear phase. (author)
[en] A new loop and fold wave-guide ion cyclotron resonance frequency (ICRF) antenna with a power output of 3 MW, which can operate at a frequency in the range of 30 MHz to 110 MHz, was designed. The design of key components of the new ICRF antenna and the characters of the new prototype ICRF antenna are presented. The thermo-mechanical analysis of both the Faraday shield and the current straps was conducted, and the stresses due to heat loads are studied in detail with different cooling-water velocities considered. In addition, the movability of prototype ICRF antenna under vacuum condition by the driving system was tested. An engineering commissioning was successfully performed on the prototype ICRF antenna using the original transmitter. The results are close to the expected.
[en] A strange magnetohydrodynamic (MHD) phenomenon has been observed on EAST tokamak. The waveform of the MHD oscillations is similar to that of fishbone mode. The observation and analyzing results suggest that the appearance of the strange oscillations is associated with intensive impurity radiation.
[en] An efficient technique for building plasma equilibria with preset or constrained q profiles by solving the nonlinear Grad–Shafranov equation iteratively is proposed. The technique uses a set of dynamic virtual poloidal field probes in the plasma core region. Numerical examples are shown for monotonic q profiles and negative central magnetic shear equilibria which are obtained with this technique. As a proof of principle, the technique is applied on the EAST tokamak, to constrain the plasma equilibrium reconstruction with the experimental q = 1 surface information from the ECE diagnostic in addition to the external magnetic measurements. (letter)
[en] Radio frequency (RF) heating in the ion cyclotron range of frequencies (ICRF) is one of the primary auxiliary heating methods for EAST. The ICRF system provides 6 MW power in primary phase and will be capable of 10 MW later. Three 1.5 MW ICRF systems in a frequency range of 25 MHz to 70 MHz have already been in operation. The ICRF heating launchers are designed to have two current straps with each driven by a RF power source of 1.5 MW. In this paper a brief introduction of the ICRF heating system capability in EAST and the preliminary results in EAST are presented. (magnetically confined plasma)
[en] A fast plasma boundary reconstruction technique based on a local expansion method is designed for EAST. It represents the poloidal flux distribution in the vacuum region by a limited number of expansions. The plasma boundary reconstructed by the local expansion method is consistent with EFIT/RT-EFIT results for an arbitrary plasma configuration. On a Linux server with Intel (R) Xeon (TM) CPU 3.2 GHz, the method completes one plasma boundary reconstruction in about 150 μs. This technique is sufficiently reliable and fast for real-time shape control.