Search

China Fusion Engineering Test Reactor (CFETR) is a tokamak reactor; one design option under the consideration of the China National Integration Design Group employs superconducting magnets. The fusion power is at the range of 50-200 MW and the duty cycle (or burning time) was envisioned as 30%-50%. The plasma current will be 10 MA and the major and minor radii are 5.7 and 1.6 m, respectively. The concept engineering design including the magnet system, vacuum vessel (VV) system, and maintenance method has been carried out in the past years. The toroidal magnetic field strength at R0 is 5 T and the maximum flux swing provided by central solenoid winding will be 160 VS. The main design work, including the electromagnetic analysis of the magnet system, has already been carried out. This paper mainly probes into the VV design and optimization based on three types of maintenance ports. Furthermore, the maintenance method, counted as one of the most important design studies, is presented specifically in this paper. It includes the design of the maintenance ports and the remote handling system design, and so on. The next design stage will mainly include mechanical design, conductor stability, different types of divertor system design, and so on.

The experimental advanced superconducting tokamak (EAST) is a superconducting tokamak, which successfully achieved the first plasma discharge in 2006. The major radius of EAST plasma is 1.9 m, and its central magnet field is 3.5 T. In the past few years, EAST has made many achievements, such as 60-s double-null divertor configuration plasma and 1-MA plasma. In 2012, 411-s long-pulse discharge with 0.28-MA plasma current and 32-s H-mode operation was achieved at the seventh campaign. In addition, various means for mitigating ELMs have also been demonstrated to facilitate long-pulse operation. The experimental results and engineering experience obtained from EAST can be used as a reference for fusion next step. To make further progress, much optimization work should be done in next step. Based on the present EAST heating system, two sets of 4-MW neutral beam injection, one set of electron cyclotron resonance heating, and lower hybrid current drive will be installed to achieve more than 30-MW total heating power. The upper divertor will be updated to W-Cu divertor, which consists of monoblock structure and tungsten armor that can withstand 10- MW·m-2 heat load. To increase the upper divertor's ash removal efficiency, one new cryopump will be added on the behind of it. Aiming to improve plasma configuration and enhance its stabilization, the EAST resonant magnetic perturbation (RMP) coils will be designed, fabricated, and installed. The RMP coils will integrate the functions of error field correction, edge localized mode, and resistive wall mode. About 45 kinds of diagnostics will be used in the next campaign and all of them will be integrated on six ports as port plug. In addition, some fusion technologies are also considered in EAST to validate for ITER and future reactor.