P. Goranson

Design of the national compact stellarator experiment (NCSX)

Abstract The National Compact Stellarator Experiment (NCSX) [ http://www.pppl.gov/ncsx/Meetings/CDR/CDRFinal/EngineeringOverview_R2.pdf ] is being designed as a proof of principal test of a quasi-axisymmetric compact stellarator. This concept combines the high beta and good confinement features of an advanced tokamak with the low current, disruption-free characteristics of a stellarator. NCSX has a three-field-period plasma configuration with an average major radius of 1.4 m, an average minor radius of 0.33 m and a toroidal magnetic field on axis of up to 2 T. The stellarator core is a complex assembly of four coil systems that surround the highly shaped plasma and vacuum vessel. Heating is provided by up to four, 1.5 MW neutral beam injectors and provision is made to add 6 MW of ICRH. The experiment will be built at the Princeton Plasma Physics Laboratory, with first plasma expected in 2007.

Engineering design of the National Spherical Torus Experiment

NSTX is a proof-of-principle experiment aimed at exploring the physics of the ‘spherical torus’ (ST) configuration, which is predicted to exhibit more efficient magnetic confinement than conventional large aspect ratio tokamaks, among other advantages. The low aspect ratio (R:a, typically 1.2‐2 in ST designs compared to 4‐5 in conventional tokamaks) decreases the available cross sectional area through the center of the torus for toroidal and poloidal field coil conductors, vacuum vessel wall, plasma facing components, etc., thus increasing the need to deploy all components within the so-called ‘center stack’ in the most efficient manner possible. Several unique design features have been developed for the NSTX center stack, and careful engineering of this region of the machine, utilizing materials up to their engineering allowables, has been key to meeting the desired objectives. The design and construction of the machine has been accomplished in a rapid and cost effective manner thanks to the availability of extensive facilities, a strong experience base from the TFTR era, and good cooperation between institutions.

Engineering Accomplishments in the Construction of NCSX

The National Compact Stellarator Experiment (NCSX) was designed to test a compact, quasi-axisymmetric stellarator configuration. Flexibility and accurate realization of its complex 3D geometry were key requirements affecting the design and construction. While the project was terminated before completing construction, there were significant engineering accomplishments in design, fabrication, and assembly. The design of the stellarator core device was completed. All of the modular coils, toroidal field coils, and vacuum vessel sectors were fabricated. Critical assembly steps were demonstrated. Engineering advances were made in the application of CAD modeling, structural analysis, and accurate fabrication of complex-shaped components and sub-assemblies. The engineering accomplishments of the project are summarized.

Progress in NCSX Construction

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 projects 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.

NSTX high temperature sensor systems

The design of the more than 300 in-vessel sensor systems for the National Spherical Torus Experiment (NSTX) has encountered several challenging fusion reactor diagnostic issues involving high temperatures and space constraints. This has resulted in unique miniature, high temperature invessel sensor systems mounted in small spaces behind plasma facing armor tiles, and they are prototypical of possible high power reactor first-wall applications. In the Center Stack, Divertor, Passive Plate, and vessel wall regions, the small magnetic sensors, large magnetic sensors, fluxloops, Rogowski coils, thermocouples, and Langmuir probes are qualified for 600 /spl deg/C operation. This rating will accommodate both peak rear-face graphite tile temperatures during operations and 350/spl deg/C bake-out conditions. Similar sensor systems including flux loops, on other vacuum vessel regions are qualified for 350 /spl deg/C operation. Cabling from the sensors embedded in the graphite tiles follows narrow routes to exit the vessel. The detailed sensor design and installation methods of these diagnostic systems developed for high-powered ST operation are discussed.

Design of the quasi-poloidal stellarator experiment (QPS)

Abstract The Quasi-poloidal stellarator (QPS), currently in the early design phase, is a low-aspect-ratio (R/a=2.7), concept exploration experiment with a non-axisymmetric, near-poloidally-symmetric magnetic configuration. The QPS design parameters are 〈R〉=0.9 m, 〈a〉=0.33 m, B=1 T, and a 1 s pulse length. The QPS device will be located at the Oak Ridge National Laboratory. Lyon et al. [ http://qps.fed.ornl.gov/pvr/pdf/qpsentire.pdf , 2001] describes the physics and engineering features in detail. The QPS device is estimated to require 4 years from start of design to first plasma in 2007.

Application of high-performance aerogel insulating materials (analysis & test results)

The NCSX stellarator core design is built around a 3-period, highly shaped plasma with an assembly of four magnet systems, the TF coils (TF), the Modular Coils (MC), the PF Coils (PF), and the Trim Coils, that surrounds an all welded Vacuum Vessel (VV). The VV features approximately 100 ports for heating, pumping, diagnostics, and maintenance access. The entire system is surrounded by a cryostat to permit operation of the coils at liquid nitrogen temperature. The VV and coils are assembled in 120° segments. The VV segments must be placed inside the MC by sliding the coils over each end of the vessel subassembly. Installation of the port extensions is done after this operation. They are slipped through access holes in the MC onto port stubs and welded on from inside. Figures 1 & 2 illustrate the assembly of the MC over a VV section and installation of TF and port extensions to complete a sector of the VV.

Performance evaluation and analysis of critical interface features of the National Compact Stellarator Experiment (NCSX)

The (18) modular coils for the National Compact Stellarator Experiment (NCSX) are joined at assembly by interfaces to form a toroidal shell which serves as the structural backbone of the device. There are six each of three coil types (A, B, and C); consequently, there are 4 distinct interface designs (A-A, A-B, B-C, C-C). This paper describes the performance evaluations and analyses used in the development of these critical interfaces. Initial analyses indicated that the baseline interface designs did not provide adequate shear capability along the inner (unbolted) legs between the modular coils and did not adequately address assembly tolerance requirements. Consequently a design effort was undertaken to develop interfaces with adequate shear capability and which would facilitate the achievement of assembly tolerances. Analyses indicated that a friction coefficient of 0.3 is necessary for “no-slip” joints with a preload value of ∼320 kN in the outboard regions. Two types of compatible segmented friction shims were developed to meet the friction requirement. One type uses alumina coated stainless steel shims and the other uses G-10/ stainless steel/ G-10 “sandwich shims.” Analyses indicated that the time constant requirements for induced currents in the shell could still be achieved with welds along all the inner (unbolted) legs except at the C-C interface. Consequently, welded interfaces utilizing alternating MIG fillet welds on each end of shims between coil castings were developed to react the shear loads. This configuration minimizes distortion since it avoids direct weld shrinkage stress across the interfaces. Analyses indicates that a 12.7 mm fillet weld has adequate shear capability, with average stress through the welds of 90–125 MPa, compared to a static limit of 217 MPa. Custom sized compression pucks located in the middle of the welded shims react the compressive loads and have average stresses less than 137 MPa. Fatigue acceptability of the welded joints was examined using crack growth data of welded specimens and estimates the maximum acceptable initial flaw size in 12.7 mm welds to be 3.2 mm. At this size, a crack would propagate to an unacceptably large size in 500,000 (5 X life).

Engineering Design Status of the Quasi-Poloidal Stellarator (QPS)

The engineering design status of the quasi-poloidal stellarator experiment (QPS) is presented. The overall configuration and the design, manufacturing R&D and assembly techniques of the core components are described

NCSX Vacuum Vessel Fabrication

The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in conjunction with the Oak Ridge National Laboratory (ORNL). The goal of this experiment is to develop a device which has the steady state properties of a traditional stellarator along with the high performance characteristics of a tokamak. A key element of this device is its highly shaped Inconel 625 vacuum vessel. The vessel is being fabricated by Major Tool and Machine, Inc. (MTM) in three identical 120deg vessel segments, corresponding to the three NCSX field periods, in order to accommodate assembly of the device. This paper describes the manufacturing of the vacuum vessel