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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.

The National Compact Stellarator Experiment (NCSX) is under construction at Princeton Plasma Physics Laboratory (PPPL). The stellarator is a complex, 3-D assembly that has been designed by a team of engineers from the partner institutions, Oak Ridge National Laboratory (ORNL) and PPPL. The complex, 3-D nature of the stellarator design and the multi-institutional, geographically dispersed nature of the team have created special management challenges. Management systems were implemented to address these challenges. The lessons learned from our experience on NCSX thus far are discussed

The Compact Ignition Tokamak (CIT) is planned as the next major step in the US fusion program. Its overall objective is to produce ignited plasma discharges. In order to realize this objective, plasma parameters which are consistent with recognized theoretical and operational limits and which constitute the most cost-effective path to ignition must be established. The key parameters include the plasma safety factor, elongation, plasma current, toroidal field and aspect ratio. Each of the parameters was considered in optimizing the present baseline wedged design. The CIT design team is presently developing a bucked design with enhanced plasma performance. A similar effort to optimize the bucked design is underway. The optimization procedure and optimization of the wedged and bucked designs are presented. >

The Tokamak Physics Experiment (TPX) is designed to develop the scientific basis for a compact and continuously operating tokamak fusion reactor. TPX has a long pulse (1000s) capability, can accommodate high divertor heat loads, has a flexible PF system, and auxiliary heating and current drive systems that make it an ideal test bed for development of attractive reactor concepts. The design incorporates superconducting magnets in both the toroidal and poloidal field magnets. Long pulse deuterium operation will produce 6/spl times/10/sup 21/ neutrons per year requiring remote maintenance of the in-vessel hardware. This paper provides an overview of the TPX Project and describes the design approach with emphasis on salient features of the tokamak.

Compact stellarators have the potential to make steady-state, disruption-free magnetic fusion systems with beta approximately 5% and relatively low aspect ratio (R/ < 4.5) compared to most drift-optimized stellarators. Magnetic quasi-symmetry can be used to reduce orbit losses. The National Compact Stellarator Experiment (NCSX) is designed to test compact stellarator physics in a high-beta quasi-axisymmetric configuration and to determine the conditions for high-beta disruption-free operation. It is designed around a reference plasma with low ripple, good magnetic surfaces, and stability to the important ideal instabilities at beta approximately 4%. The device size, available heating power, and pulse lengths provide access to a high-beta target plasma state. The NCSX has magnetic flexibility to explore a wide range of equilibrium conditions and has operational flexibility to achieve a wide range of beta and collisionality values. The design provides space to accommodate plasma-facing components for divertor operation and ports for an extensive array of diagnostics.

The Tokamak Physics Experiment (TPX) is designed to develop the scientific basis for a compact and continuously operating tokamak fusion reactor. TPX has a long pulse (1000s) capability, can accommodate high divertor heat loads, has a flexible poloidal field (PF) system, and auxiliary heating and current drive systems that make it an ideal test bed for development of attractive reactor concepts. The design incorporates superconducting magnets in both the toroidal field (TF) and poloidal field (PF) systems. Long pulse deuterium operation will produce 6 {times} 10{sup 21} neutrons per year requiring remote maintenance of the in-vessel hardware. This paper provides an overview of the TPX design with the emphasis on developments in the tokamak design since the Conceptual Design Review (CDR) in March, 1993.