G. Zhuang

The China Fusion Engineering Test Reactor (CFETR) is the next device in the roadmap for the realization of fusion energy in China, which aims to bridge the gaps between the fusion experimental reactor ITER and the demonstration reactor (DEMO). CFETR will be operated in two phases. Steady-state operation and self-sufficiency will be the two key issues for Phase I with a modest fusion power of up to 200 MW. Phase II aims for DEMO validation with a fusion power over 1 GW. Advanced H-mode physics, high magnetic fields up to 7 T, high frequency electron cyclotron resonance heating and lower hybrid current drive together with off-axis negative-ion neutral beam injection will be developed for achieving steady-state advanced operation. The recent detailed design, research and development (R&D) activities including integrated modeling of operation scenarios, high field magnet, material, tritium plant, remote handling and future plans are introduced in this paper.

The Chinese Fusion Engineering Testing Reactor (CFETR), complementing the ITER facility, is aiming to demonstrate fusion energy production up to 200 MW initially and to eventually reach DEMO relevant power level 1 GW, to manifest a high duty factor of 0.3–0.5, and to pursue tritium self-sufficiency with tritium breeding ratio (TBR) >1. The key challenge to meet the missions of the CFETR is to run the machine in steady state (or long pulse) and high duty factor. By using a multi-dimensional code suite with physics-based models, self-consistent steady-state and hybrid mode scenarios for CFETR have been developed under a high magnetic field up to 6.5 T. The negative-ion neutral beam injection together with high frequency electron cyclotron wave and lower hybrid wave (and/or fast wave) are proposed to be used to drive the current. Subsequently the engineering design of CFETR including the magnet system, vacuum system, tritium breeding blanket, divertor, remote handling and maintenance system will be introduced. Some research and development (R&D) activities are also introduced in this paper.

The impact of plasma shaping on electron heat transport is investigated in TCV L-mode plasmas. The study is motivated by the observation of an increase in the energy confinement time with decreasing plasma triangularity which may not be explained by a change in the temperature gradient induced by changes in the geometry of the flux surfaces. The plasma triangularity is varied over a wide range, from positive to negative values, and various plasmas conditions are explored by changing the total electron cyclotron (EC) heating power and the plasma density. The mid-radius electron heat diffusivity is shown to significantly decrease with decreasing triangularity and, for similar plasma conditions, only half of the EC power is required at a triangularity of -0.4 compared with +0.4 to obtain the same temperature profile. Besides, the observed dependence of the electron heat diffusivity on the electron temperature, electron density and effective charge can be grouped in a unique dependence on the plasma effective collisionality. In summary, the electron heat transport level exhibits a continuous decrease with decreasing triangularity and increasing collisionality. Local gyro-fluid and global gyro-kinetic simulations predict that trapped electron modes are the most unstable modes in these EC heated plasmas with an effective collisionality ranging from 0.2 to 1. The modes stability dependence on the plasma triangularity is investigated.

The TEXT/(TEXT-U) tokamak, formerly built and operated by the University of Texas at Austin in USA, was dismantled and shipped to China in 2004, and renamed as the Joint TEXT (J-TEXT) tokamak. The reconstruction work, which included reassembly of the machine and development of peripheral devices, was completed in the spring of 2007. Consequently, the first plasma was obtained at the end of 2007. At present, a typical J-TEXT ohmic discharge can produce a plasma with flattop current up to 220 kA and lasting for 300 ms, line-averaged density above 2 × 10 19 m −3 , and an electron temperature of about 800 eV, with a toroidal magnetic field of 2.2 T. A number of diagnostic devices used to facilitate the routine operation and experimental scenarios were developed on the J-TEXT tokamak. Hence, the measurements of the electrostatic fluctuations in the edge region and conditional analysis of the intermittent burst events near the last closed flux surface were undertaken. The observation and simple analysis of MHD activity and disruption events were also performed. The preliminary experimental results and the future research plan for the J-TEXT are described in detail.