2 edition of Steady-state resistive toroidal-field coils for tokamak reactors found in the catalog.
Work supported by the U.S. Department of Energy contract no. EY-76-C-02-3073.Paper presented at the Eighth Symposium on the Engineering Problems of Fusion Research (San Francisco) 13-16 November 1979.Dec. 1979.Includes bibliographical references.
|Statement||Dept. of Energy, Plasma Physics Laboratory|
|Publishers||Dept. of Energy, Plasma Physics Laboratory|
|The Physical Object|
|Pagination||xvi, 125 p. :|
|Number of Pages||70|
|2||PPPL (Series) -- 1620.|
|3||PPPL ; 1620|
nodata File Size: 8MB.
In this chapter, the optimal control problem of nonlinear PDE systems is considered. The main heating systems on MAST are two neutral beam PINI injectors each with a maximum capability to inject 2. Therefore, a hollow current profile has been considered in the SSTR concept . In , Xu et al. Let's take a look behind the scenes at the last-stage fabrication activities that are mobilizing the expertise and skill of heavy industry specialists under the responsibility of Japanese QST, the National Institutes for Quantum and Radiological Science and Technology.
The Fusion Engineering Design Center FEDC is part of a national design team that is developing the conceptual design of the Compact Ignition Tokamak CIT.
It was obvious from Steady-state resistive toroidal-field coils for tokamak reactors outset that the compact and complex geometry of the joint design presented a special challenge in the areas of reliability, assembly, maintenance, disassembly, and cost. Transport analysis and simulation suggest that the combination of high density gradient and high Shafranov shift allows turbulence stabilization and higher confinement.
A design requirement on the TF coils is that they contain readily demountable joints to facilitate replacement of components inside the bore of the coil. 1 and a typical magnetic equilibrium for MAST is shown in Fig. Aspects of cooling, magnetic stress, and construction are addressed for several reference designs.
At LLNL, a detail model of the TF coil straight leg near the equator was used to obtain stresses and displacements during TF operation only. An adequate shield is determined to be 10 cm of zirconium borohydride, which reduces the nuclear heating of the TF coils by a factor of 11.
Recent advances in tokamak physics indicate the spherical tokamak may offer a magnetic fusion development path that can be started with a small size pilot plant and progress smoothly to larger power plants.
Using on-axis ECH injection, tungsten accumulation is avoided on EAST, and this is reproduced in modeling. The evaluations examined fabrication feasibility and complexity, thermal-electrical performance at approximately two-thirds of the steady-state design conditions, and installation and assembly processes. Results of the thermal-electrical tests were analyzed and extrapolated to predict performance at peak design parameters.
The tensile load on the vertical leg can be eliminated if the demountable joints can slide.
The reversed field experiment machine the largest RFP device in operation in Padova, Italy is used to study the physical problems arising from the RFP configuration.
The results of this study show the possibility of designing a tokamak reactor with reduced size in comparison with other INTOR like devices, still gaining some margins in front of the uncertainties in the scaling laws for plasma physics parameters and retaining the presence of a blanket with a tritium breeding ratio of about 1.
50 B t T 4.