The ARIES-CS - A compact stellarator power plant

Stellarators have many desirable features as fusion power plants. A detailed and integrated study of compact stellarator configurations, ARIES-CS, was initiated recently to advance our understanding of attractive compact stellarator power plants and to define key R and D areas. The first goal of the S study was to develop configurations which are similar in size with advanced tokamaks as earlier stellarator power plant studies led to devices with large sizes ( = 14-20 m). This was achieved by a) developing configuration with a low plasma aspect ratio, and b) minimizing the space between plasma and the coils. We focused on quasi axisymmetric (QA) configurations as they are able to operate at a relatively low plasma aspect ratio (∼ 4-5). Three distinct classes of QA configuration is considered. First is NCSX-class configuration. Three variants of this configurations have been developed to explore trade-off between confinement of fast a particles and linear MHD stability. Progress has been made to reduce loss of a particles to < 5%. It appears that the introducing a bias in the principle mirror term in the magnetic spectrum, plays an important role in reducing a particle losses. Second is MHH2 which aims that developing a very low aspect ratio geometry (∼2.6) with relatively simpler coils. Third, SNS, is aimed at a configuration with excellent flux surface quantity and nearly flat rotational transform. In latter two cases, strict adherence to linear MHD stability is deemphasized. To reduce the blanket-coil spacing, a novel approach was developed in ARIES-CS in which the blanket at the critical areas of minimum stand-off is replaced by a highly efficient WC-based shield. In principle, by utilizing the shield-only region in strategic areas, we have been able to reduce the minimum stand-off, Δmin, by -30% compared to a uniform radial build that was assumed in previous studies. The reduced Δmin, together with the lower aspect ratio plasma lead to power plants that have similar size as advanced tokamak designs ( = 7-8 m). The device configuration, assembly, and maintenance procedures appear to impose severe constraints: Two distinct approaches were developed. Modular coils are designed to examine the geometric complexity and to understand the constraints imposed by the maximum allowable field, desirable coil-plasma separation, coil-coil spacing, and other coil parameters. (author)