The Korea Superconducting Tokamak Advanced Research (KSTAR)
project is the major effort of the national fusion programme of the Republic of Korea. Its aim is
to develop a steady state capable advanced superconducting tokamak to
establish a scientific and technological basis for an attractive fusion
reactor. The major parameters of the tokamak are: major radius 1.8 m, minor
radius 0.5 m, toroidal field 3.5 T and plasma current 2 MA, with a
strongly shaped plasma cross-section and double null divertor. The initial
pulse length provided by the poloidal magnet system is 20 s, but the pulse
length can be increased to 300 s through non-inductive current drive. The
plasma heating and current drive system consists of neutral beams,
ion cyclotron waves, lower hybrid waves and electron cyclotron waves for
flexible profile control in advanced tokamak operating modes. A
comprehensive set of diagnostics is planned for plasma control,
performance evaluation and physics understanding. The project has
completed its conceptual design and moved to the engineering design and
construction phase. The target date for the first plasma is 2002.
This article describes a reliable gateway for in-vehicle networks. Such networks include local interconnect networks, controller area networks, and FlexRay. There is some latency when transferring a message from one node (source) to another node (destination). A high probability of error exists due to different protocol specifications such as baud-rate, and message frame format. Therefore, deploying a reliable gateway is a challenge to the automotive industry. We propose a reliable gateway based on the OSEK/VDX components for in-vehicle networks. We also examine the gateway system developed, and then we evaluate the performance of our proposed system.
DIII-D physics research addresses critical challenges for the operation of ITER and the next generation of fusion energy devices. This is done through a focus on innovations to provide solutions for high performance long pulse operation, coupled with fundamental plasma physics understanding and model validation, to drive scenario development by integrating high performance core and boundary plasmas. Substantial increases in off-axis current drive efficiency from an innovative top launch system for EC power, and in pressure broadening for Alfven eigenmode control from a co-/counter-I
p steerable off-axis neutral beam, all improve the prospects for optimization of future long pulse/steady state high performance tokamak operation. Fundamental studies into the modes that drive the evolution of the pedestal pressure profile and electron vs ion heat flux validate predictive models of pedestal recovery after ELMs. Understanding the physics mechanisms of ELM control and density pumpout by 3D magnetic perturbation fields leads to confident predictions for ITER and future devices. Validated modeling of high-Z shattered pellet injection for disruption mitigation, runaway electron dissipation, and techniques for disruption prediction and avoidance including machine learning, give confidence in handling disruptivity for future devices. For the non-nuclear phase of ITER, two actuators are identified to lower the L–H threshold power in hydrogen plasmas. With this physics understanding and suite of capabilities, a high poloidal beta optimized-core scenario with an internal transport barrier that projects nearly to Q = 10 in ITER at ∼8 MA was coupled to a detached divertor, and a near super H-mode optimized-pedestal scenario with co-I
p beam injection was coupled to a radiative divertor. The hybrid core scenario was achieved directly, without the need for anomalous current diffusion, using off-axis current drive actuators. Also, a controller to assess proximity to stability limits and regulate β
N in the ITER baseline scenario, based on plasma response to probing 3D fields, was demonstrated. Finally, innovative tokamak operation using a negative triangularity shape showed many attractive features for future pilot plant operation.
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