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The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge National Laboratory (ORNL). Its mission is to develop the physics understanding of the compact stellarator and evaluate its potential for future fusion energy systems. Compact stellarators use 3D plasma shaping to produce a magnetic configuration that can be steady state without current drive or feedback control of instabilities. The NCSX has major radius 1.4 m, aspect ratio 4.4, 3 field periods, and a quasi-axisymmetric magnetic field. It is predicted to be stable and have good magnetic surfaces at beta > 4% and to have tokamak-like confinement properties. The device will provide the plasma configuration flexibility and the heating and diagnostic access needed to test physics predictions. Component production has advanced substantially since the first contracts were placed in 2004. Manufacture of the vacuum vessel was completed in 2006. All eighteen modular coil winding forms have been delivered, and twelve modular coils have been wound and epoxy impregnated. A contract for the (planar) toroidal field coils was placed in 2006 and manufacture is in progress. Assembly activities have begun and will be the project's main focus in the next few years. The engineering challenge of NCSX is to meet the requirements for complex geometries and tight tolerances within the cost and schedule constraints of a construction project. This paper will focus on how the engineering challenges of component production have been resolved, and how the assembly challenges are being met.

The National Compact Stellarator Experiment (NCSX) is a quasi-axisymmetric plasma experiment that combines the high beta and good confinement of an advanced tokamak with the low current, disruption-free characteristics of a stellarator. The experiment is based on a three field-period plasma configuration with an average major radius of 1.4 m, a minor radius of 0.3 m, and a toroidal magnetic field on axis of up to 2 T. The modular coils are one set in a complex assembly of four coil systems that surround the highly shaped plasma. There are six each of three coil types in the assembly, for a total of 18 modular coils. The coils are constructed by winding copper cable onto a cast stainless steel winding form that has been machined to high accuracy, so that the current center of the winding pack is within +/-1.5 mm of its theoretical position. The modular coils operate at a temperature of 80 K and are subjected to rapid heating and stress during a pulse. The final coil design has presented many challenges with its requirements for winding accuracy, good thermal performance, a robust supporting structure, and ease of assembly and maintenance

The design status of the stellarator core for the National Compact Stellarator Experiment (NCSX) is presented. The purpose, configuration, and possible manufacturing and assembly techniques of the various components of the core are described.

The National Compact Stellarator Experiment (NCSX) is proposed as a test of a low aspect ratio, quasi-axisymmetric plasma configuration that exhibits high beta and good confinement in a disruption-free environment. The experiment will be built at Princeton Plasma Physics Laboratory (PPPL) and utilize some ancillary equipment. The NCSX stellarator core is a complex assembly of four coil systems which provide the magnetic field for plasma shaping and position control, inductive current drive, and field error correction. The primary magnets are the modular coils, which provide up to 2-T at an average major radius of 1.4-m. Other magnets include toroidal field (TF) coils, poloidal field (PF) coils, and trim coils, which can be used to control resonant field errors. The magnets are supported by an integral shell structure, which also serves as the winding form for the modular coils. The coils and structure have been evaluated for thermal stress and electromagnetic loads during normal operating conditions. The results indicate that the performance of the modular coil system is acceptable.

The National Spherical Tokamak Experiment (NSTX) is an ultra low aspect ratio (R/a=1.25; R=80 cm) device designed for a plasma current of 1 MA. It features auxiliary heating and current drive and a close-fitting conducting shell to maximize plasma pressure (43% beta). NSTX is designed for a 5 sec. experimental pulse to demonstrate quasi-steady state non-inductively driven advanced tokamak operation. The design takes maximum advantage of existing components and facilities from previous devices at PPPL to reduce the program costs. The device will be sited in the former Princeton Large Torus (PLT) test cell and will utilize the PLT radiation shielding, base structure, and cell utilities. NSTX will utilize the S-1 Spheromak vacuum vessel, poloidal field coils, and capacitor banks (for helicity injection). The Poloidal Beta Experiment-Modified (PBX-M) power supplies will be shared to power the PF and TF coil systems. Existing RF hardware and infrastructure will be used for heating systems. TFTR data acquisition and diagnostics resources are planned to be used. In total, NSTX will utilize site credits with a value of /spl sim/$50 M, reducing base construction cost of the device to $18.6 M. Twelve water-cooled copper demountable toroidal field (TF) coils provide the 5.4 kg (pulsed) and 3.5 kg (long pulse >5 sec) toroidal field at the plasma center. Poloidal fields are generated by windings contained in the center column and four pairs water-cooled copper coils supported directly on the vacuum vessel. One of the most critical components of the device is the center stack, which consists of the inner legs of the TF coils overwrapped with ohmic heating and poloidal field windings. The ohmic heating coil windings are designed to optimize the V-s and together with the PF coils, produce a flux swing of 1 V-s.

The National Spherical Tokamak Experiment is a low-aspect-ratio tokamak facility currently in the conceptual design phase. The vacuum vessel, in addition to provides a high vacuum chamber for plasma experiments, also serves as a load-carrying element for supporting the center stack, the external poloidal field coils, and the toroidal field coils. Design requirements and structural considerations for the vacuum vessel and its support structure were described in the paper. Structural responses were calculated using the finite element method. Various loading conditions including in the analyses are dead weight, static electromagnetic loads, plasma disruptions, bakeout thermal effects, vacuum pressure, seismic excitations, and simultaneously occurrence of such loadings.