Revamped German stellarator should run longer, hotter and compete with tokamaks
Revamped German stellarator should run longer, hotter and compete with tokamaks
Concept Design of CFETR Tokamak Machine Song, Yun Tao; Wu, Song Tao; Li, Jian Gang ...
IEEE transactions on plasma science,
03/2014, Letnik:
42, Številka:
3
Journal Article
Recenzirano
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 R 0 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.
Abstract
Extended-MHD modeling of planned runaway electron mitigation coils (REMC) for SPARC and DIII-D is performed with the NIMROD code. A coil has been designed for each machine, with the two ...differing in shape and location, but both having n=1 symmetry (with n the toroidal mode number). Compared to previous modeling efforts, three improvements are made to the simulations boundary conditions. First a resistive wall model is used in place of an ideal wall. Second, the ThinCurr code is used to compute the time-dependent 3D fields used as magnetic boundary conditions for the simulations. Third, the simulation boundary is moved from the the first-wall location to the vacuum vessel (VV), which extends the boundary past the location of the internal REMC. To remove the 3D coil from the simulation domain, an equivalent set of 3D fields is calculated at the VV boundary that produce approximately the same field distribution at the last closed flux surface assuming vacuum between the two. Each of these three boundary condition improvements leads to an improvement in the predicted performance of the REMC for both machines. The resistive wall alone primarily effects the resonance of the coil with the plasma after the TQ, affecting the q-profile evolution in the SPARC modeling, and allowing the applied spectrum to be modified in response to the plasma in the DIII-D modeling. The movement of the simulation boundary has the most significant effect on the RE confinement overall, including in the early stages, particularly for a DIII-D inner wall limited equilibrium, where the RE loss fraction increases from 90% to >99%, with SPARC RE losses also occurring much earlier when the boundary is placed at the VV.
Spherical tokamaks (STs) exhibit significant promise as the foundation for compact fusion power plants, offering reduced aspect ratios and enhanced plasma performance that can potentially lower ...capital costs compared to conventional tokamak designs. The key to achieving an optimal design lies in understanding the sensitivity of the fusion power plant to plasma energy confinement times. However, due to the intricate nature of transport physics and the scarcity of data on highly radiative plasmas required for power plants, extrapolating performance from existing machines introduces substantial uncertainties. In this study, we employed the world-leading fusion power plant systems code, PROCESS, to explore the effects of different energy confinement time scalings on scoping and determining the design of a 1-<inline-formula> <tex-math notation="LaTeX">\text{GW}_{e}</tex-math> </inline-formula> net electric ST power plant. By comparing various commonly used scalings, we highlight the design impact of employing ST scalings versus those typically applied to conventional aspect ratios, considering both size and performance aspects. Our findings demonstrate that when allowed to freely optimize the choice of confinement scaling has negligible impact on the optimally found design point and is instead driven highly by engineering constraints. In a highly constrained scenario, the conventional IPB98(y,2) scaling consistently shows conservative values across a range of ST plasma performance scenarios. We recommend its utilization for future large design space exploration studies as a low-risk choice due to its intermediary performance between the broad scope of ST scalings and also as a proxy for addressing complex transport considerations in configuring initial ST concept designs.
