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Equilibrium issues encountered in the design of the National Compact Stellarator Experiment (NCSX) are discussed, focusing particularly on equilibrium magnetic islands. Significant improvements have been made to the VMEC equilibrium code to deal with numerical challenges at the low aspect ratios characterizing the NCSX design. Modifications to the PIES code have increased its speed, allowing routine evaluation of flux surfaces for candidate configurations. An optimizer has been built around the PIES code for healing magnetic islands, modifying the coil shapes to suppress resonant components of the magnetic field while preserving desired physics and engineering properties. The modified coils produce improved flux surface quality for a range of configurations. Neoclassical effects, which are not included in the PIES calculations, are estimated using a cylindrical model and are found to further reduce island widths significantly.

Recent experiments (Synakowski et al 2004 Nucl. Fusion 43 1648, Lloyd et al 2004 Plasma Phys. Control. Fusion 46 B477) on the Spherical Tokamak (or Spherical Torus, ST) (Peng 2000 Phys. Plasmas 7 1681) have discovered robust plasma conditions, easing shaping, stability limits, energy confinement, self-driven current and sustainment. This progress has encouraged an update of the plasma conditions and engineering of a Component Test Facility (CTF), (Cheng 1998 Fusion Eng. Des. 38 219) which is a very valuable step in the development of practical fusion energy. The testing conditions in a CTF are characterized by high fusion neutron fluxes Γn ≈ 8.8 × 1013 n s−1 cm−2 ('wall loading' WL ≈ 2 MW m−2), over size-scale >105 cm2 and depth-scale >50 cm, delivering >3 accumulated displacement per atom per year ('neutron fluence' >0.3 MW yr−1 m−2) (Abdou et al 1999 Fusion Technol. 29 1). Such conditions are estimated to be achievable in a CTF with R0 = 1.2 m, A = 1.5, elongation ~3, Ip ~ 12 MA, BT ~ 2.5 T, producing a driven fusion burn using 47 MW of combined neutral beam and RF heating power. A design concept that allows straight-line access via remote handling to all activated fusion core components is developed and presented. The ST CTF will test the lifetime of single-turn, copper alloy centre leg for the toroidal field coil without an induction solenoid and neutron shielding and require physics data on solenoid-free plasma current initiation, ramp-up to and sustainment at multiple megaampere level. A systems code that combines the key required plasma and engineering science conditions of CTF has been prepared and utilized as part of this study. The results show high potential for a family of relatively low cost CTF devices to suit a range of fusion engineering and technology test missions.

The National Compact Stellarator Experiment (NCSX) is a quasi-axisymmetric facility that combines the high beta and good confinement features 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. At this time, the project has completed construction of several prototype components which validate the fabrication and inspection processes that are planned for the production coils. In addition, some advanced techniques for error-field compensation and assembly simulation using computer-aided design (CAD) have been developed.

The quasi-poloidal stellarator (QPS) hybrid has been developed using a stellarator optimization approach that has proven to be compatible with both low aspect ratio and significantly reduced neoclassical transport relative to anomalous levels. A unique characteristic of this type of quasi-symmetry is a reduced viscous damping level for poloidal plasma flows. Since the plasma-generated E × B and diamagnetic flows are nearly poloidal, minimal parallel flows (and viscous stress) are required to achieve parallel pressure balance in comparison with configurations such as the tokamak, in which the plasma induced flows are nearly perpendicular to the direction of minimum viscosity and relatively larger parallel flows are required. In addition to this impact on neoclassical flows it is also anticipated that quasi-poloidal symmetry will minimize resistance to self-organized plasma-turbulence-driven shear flows and ease access to enhanced confinement states. In order to test these and other transport issues, the QPS device has been designed with a high degree of flexibility by allowing variable current capability not only in its vertical and toroidal coilsets but also in each separate modular coil group. Numerical optimizations have demonstrated that this flexibility can be used not only to modify transport properties, such as the poloidal viscosity, but also to directly suppress magnetic islands.

A broadly based study of the fusion engineering and plasma science conditions of a Component Test Facility (CTF), using the Spherical Torus or Spherical Tokamak (ST) configuration, have been carried out. The chamber systems testing conditions in a CTF are characterized by high fusion neutron fluxes {gamma}{sub n} > 4.4 x 10{sup 13} n/s/cm{sup 2}, over size scales > 10{sup 5} cm{sup 2} and depth scales > 50 cm, delivering > 3 accumulated displacement per atom (dpa) per year. The desired chamber conditions can be provided by a CTF with R{sub 0} 1.2 m, A = 1.5, elongation {approx} 3, I{sub p} {approx} 9 MA, B{sub T} {approx} 2.5 T, producing a driven fusion burn using 36 MW of combined neutral beam and RF power. Relatively robust ST plasma conditions are adequate, which have been shown achievable without active feedback manipulation of the MHD modes. The ST CTF will test the single-turn, copper alloy center leg for the toroidal field coil without an induction solenoid and neutron shielding, and require physics data on solenoid-free plasma current initiation, ramp-up, and sustainment to multiple MA level. A new systems code that combines the key required plasma and engineering science conditions of CTFmore » has been prepared and utilized as part of this study. The results show high potential for a family of lower-cost CTF devices to suit a variety of fusion engineering science test missions.« less

Recent progress(1) in plasma science of the Spherical Tokamak (or Spherical Torus, ST)(2) has indicated relatively robust plasma conditions in a broad number of topical area including strong shaping, stability limits, energy confinement, self-driven current, and sustainment. This progress has enabled an extensive update of the plasma science and fusion engineering conditions of a Component Test Facility (CTF)(3), which is potentially a necessary step in the development of practical fusion energy. The chamber systems testing conditions in a CTF are characterized by high fusion neutron fluxes n > 4.4 1013 n/s/cm2, over sizescale > 105 cm2 and depth-scale > 50 cm, delivering > 3 accumulated displacement per atom (dpa) per year(4). Such chamber conditions are calculated to be achievable in a CTF with R0 = 1.2 m, A = 1.5, elongation ~ 3, Ip ~ 9 MA, BT ~ 2.5 T, producing a driven fusion burn using 36 MW of combined neutral beam and RF power. The ST CTF will test the life time of single-turn, copper alloy center leg for the toroidal field coil without an induction solenoid and neutron shielding, and require physics data on solenoid-free plasma current initiation, ramp-up, and sustainment to multiple MA level. A newmore » systems code that combines the key required plasma and engineering science conditions of CTF has been prepared and utilized as part of this study. The results show high potential for a family of relatively low cost CTF devices to suit a range of fusion engineering science test missions.« less

An important goal for a stellarator design is to incorporate enough flexibility to experimentally test a range of physics issues. The proposed Quasi-Poloidal Stellarator device achieves this by allowing independently variable currents in the modular, vertical field, and toroidal coil sets. Numerical optimizations and modeling show that this can allow significant tests of neoclassical cross-field transport rates, reduced poloidal flow damping (relative to the tokamak), and magnetic island width control. This flexibility is achieved in a unique, very low aspect ratio (R 0 / = 2.7) two-field period (racetrack-shaped) configuration that generates rotational transform from a combination of internal plasma currents and external shaping.

Second ballooning stability is examined in quasipoloidally symmetric, compact stellarator configurations. These high-β (volume-average β>4%) free-boundary equilibria are calculated using a reference Quasi-Poloidal Stellarator (QPS) configuration. QPS plasmas have low-shear, stellarator-like rotational transform profile with |B| that is approximately poloidally symmetric. The high-β QPS equilibria are similar in their magnetic configuration to previously studied tokamak-stellarator hybrids which have a high-shear, tokamak-like rotational transform profile. Both types of configurations have strong magnetic wells and consequently high interchange stability β limits. Free-boundary QPS equilibria have regions of second stability at high β. For infinite-n ballooning modes in QPS plasmas, the boundary for first instability is 〈β〉∼2% and the boundary for second stability is 〈β〉∼6%. Finite-n ballooning mode calculations show higher β limits, 〈β〉>5%. Increasing plasma current (for fixed plasma pressure) can lower t...

A compact quasi-poloidally symmetric stellarator (QPS) plasma and coil configuration is described that has desirable physics properties and engineering feasibility with a very low aspect ratio plas...

For the (non-axisymmetric) stellarator class of plasma confinement devices to be feasible candidates for fusion power stations it is essential that, to a good approximation, the magnetic field lines lie on nested flux surfaces; however, the inherent lack of a continuous symmetry implies that magnetic islands responsible for breaking the smooth topology of the flux surfaces are guaranteed to exist. Thus, the suppression of magnetic islands is a critical issue for stellarator design, particularly for small aspect ratio devices. Pfirsch–Schluter currents, diamagnetic currents and resonant coil fields contribute to the formation of magnetic islands, and the challenge is to design the plasma and coils such that these effects cancel.Magnetic islands in free-boundary high-pressure full-current stellarator magnetohydrodynamic equilibria are suppressed using a procedure based on the Princeton Iterative Equilibrium Solver (Reiman and Greenside 1986 Comput. Phys. Commun. 43 157) which iterates the equilibrium equations to obtain the plasma equilibrium. At each iteration, changes to a Fourier representation of the coil geometry are made to cancel resonant fields produced by the plasma. The changes are constrained to preserve certain measures of engineering acceptability and to preserve the stability of ideal kink modes. As the iterations continue, the coil geometry and the plasma simultaneously converge to an equilibrium in which the island content is negligible, the plasma is stable to ideal kink modes, and the coils satisfy engineering constraints. The method is applied to a candidate plasma and coil design for the National Compact Stellarator eXperiment (Reiman et al 2001 Phys. Plasma 8 2083).

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.

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.

The lack of axisymmetry in stellarators guarantees that in general magnetic islands and chaotic magnetic field lines will exist. As particle transport is strongly tied to the magnetic field lines, magnetic islands and chaotic field lines result in poor plasma confinement. For stellarators to be feasible candidates for fusion power stations it is essential that, to a good approximation, the magnetic field lines lie on nested flux-surfaces, and the suppression of magnetic islands is a critical issue for stellarator coil design, particularly for small aspect ratio devices. A procedure for modifying stellarator coil designs to eliminate magnetic islands in free-boundary full-pressure magnetohydrodynamic equilibria is presented. Islands may be removed from coil–plasma free-boundary equilibria by making small changes to the coil geometry and also by variation of trim coil currents. A plasma and coil design relevant to the National Compact Stellarator Experiment is used to illustrate the technique.

A method is presented for designing coils for compact stellarators. In contrast to methods that select a finite number of coils from an optimal continuous surface current distribution, the COILOPT code solves for the optimal parameters in an explicit representation of modular coils on a toroidal winding surface that is well separated from the plasma boundary, together with the coefficients of the winding surface. The problem is posed as a balance between approximating a prescribed magnetic configuration and satisfying certain critical engineering requirements. Results are presented for quasi-axisymmetric and quasi-poloidal compact stellarator designs.

Compact optimized stellarators offer novel solutions for confining high-beta plasmas and developing magnetic confinement fusion. The three-dimensional plasma shape can be designed to enhance the magnetohydrodynamic (MHD) stability without feedback or nearby conducting structures and provide drift-orbit confinement similar to tokamaks. These configurations offer the possibility of combining the steady-state low-recirculating power, external control, and disruption resilience of previous stellarators with the low aspect ratio, high beta limit, and good confinement of advanced tokamaks. Quasi-axisymmetric equilibria have been developed for the proposed National Compact Stellarator Experiment (NCSX) with average aspect ratio 4-4.4 and average elongation similar to 1.8. Even with bootstrap-current consistent profiles, they are passively stable to the ballooning, kink, vertical, Mercier, and neoclassical-tearing modes for beta > 4%, without the need for external feedback or conducting walls. The bootstrap current generates only 1/4 of the magnetic rotational transform at beta = 4% (the rest is from the coils); thus the equilibrium is much less non-linear and is more controllable than similar advanced tokamaks. The enhanced stability is a result of 'reversed' global shear, the spatial distribution of local shear, and the large fraction of externally generated transform. Transport simulations show adequate fast-ion confinement and thermal neoclassical transport similar to equivalent tokamaks. Modular coils have been designed which reproduce the physics properties, provide good flux surfaces, and allow flexible variation of the plasma shape to control the predicted MHD stability and transport properties.

Physics issues are discussed for compact stellarator configurations which achieve good confinement by the fact that the magnetic field modulus |B| in magnetic co-ordinates is dominated by poloidally symmetric components. Two distinct configuration types are considered: (1) those which achieve their drift optimization and rotational transform at low β and low bootstrap current by appropriate plasma shaping; and (2) those which have a greater reliance on plasma β and bootstrap currents for supplying the transform and obtaining quasi-poloidal symmetry. Stability analysis of the latter group of devices against ballooning, kink and vertical displacement modes has indicated that stable β values on the order of 15% are possible. The first class of devices is being considered for a low β near term experiment that could explore some of the confinement features of the high β configurations.

Experimental devices for the study of the physics of high beta (β 4%), low aspect ratio (A 4.5) stellarator plasmas require coils that will produce plasmas satisfying a set of physics goals, provide experimental flexibility and be practical to construct. In the course of designing a flexible coil set for the National Compact Stellarator Experiment, several innovations have been made that may be useful in future stellarator design efforts. These include: the use of singular value decomposition methods for obtaining families of smooth current potentials on distant coil winding surfaces from which low current density solutions may be identified; the use of a control matrix method for identifying which few of the many detailed elements of a stellarator boundary must be targeted if a coil set is to provide fields to control the essential physics of the plasma; the use of a genetic algorithm for choosing an optimal set of discrete coils from a continuum of potential contours; the evaluation of alternate coil topologies for balancing the trade-off between physics objectives and engineering constraints; the development of a new coil optimization code for designing modular coils and the identification of a `natural' basis for describing current sheet distributions.

The scientific parameters and the technology issues for a modest size spherical torus (ST) at 10 MA plasma current are discussed. This class of devices includes a DT-capable ST experiment (DTST, R0 = 1.2 m) for extended plasma performance tests for limited pulse lengths and neutron fluences, and a volume neutron source (VNS, R0 = 1.1 m) for steady state energy technology testing to high neutron fluences. The scientific issues of interest for DTST include non-inductive ramp-up of plasma current on a limited timescale (~30 s), the confinement needed for high Q burn, the behaviour of energetic particles, the physics and techniques to handle intense plasma exhaust, and the possibility of high performance plasma regimes free of disruptions or large disruption impact. Of further interest for the VNS would be steady state operation using large external current drive, possibly at a modest Q (~1-2), achieving significant neutron wall loading (~1 MW/m2) and a configuration relatively amenable to remote maintenance. A much longer timescale would be permitted in a VNS for non-inductive current ramp-up. The centre leg of the toroidal field coils, possibly multiturn for DTST and necessarily single turn for a VNS without significant nuclear shielding, presents technical and material issues of unique importance to the ST. Positive ion neutral beam injection and high harmonic fast wave (~80 MHz) heating and current drive systems already available are likely to be adequate for DTST following pulse length extension to ~50 s. Given an adequate physics database, the remaining enabling technologies needed for the VNS appear largely similar in nature to those of the ITER EDA design.

The mission of the National Spherical Torus Experiment (NSTX) is to prove the principles of spherical torus physics by producing high-beta toroidal plasmas that are non-inductively sustained, and whose current profiles are in steady-state. NSTX will be one of the first ultra low a[P(input) up to 11 MW] in order to produce high-beta toroidal (25 to 40%), low collisionality, high bootstrap fraction (less than or equal to 70%) discharges. Both radio-frequency (RF) and neutral-beam (NB) heating and current drive will be employed. Built into NSTX is sufficient configurational flexibility to study a range of operating space and the resulting dependences of the confinement, micro- and MHD stability, and particle and power handling properties. NSTX research will be carried out by a nationally based science team.

A cost and performance analysis of a spherical torus-based volume neutron source (ST-VNS) is performed using the SUPERCODE systems code. An ST-VNS, proposed as a reduced-cost option for testing the nuclear technologies necessary for tokamak power reactors, will cost about 25% of the construction cost of the International Thermonuclear Experimental Reactor Engineering Design Activities (ITER-EDA) when it is operating in an initial test phase with 0.5 MW m 2 wall load. Advanced operation at 5 MW m -2 wall load is possible with surface elongation κ = 3, safety factor q 95 = 10, confinement enhancement factor H = 2.3, and Troyon beta limit β N = 5%, at a direct cost of about 30% that of the ITER-EDA. There is a narrow region β N and H space which is useful for attaining high neutron wall loads.

The mechanical and thermal design issues of the centerpost (CP) for a spherical torus based volumetric neutron source (ST-VNS) is described. It is intended to provide a test bed for developing nuclear technologies as well as for qualifying blanket designs for use in fusion devices. As scoped out, the VNS is capable of staged operation from a neutron wall loading of 0.5–5.0 MW/m 2 as the physics and engineering design assumptions are raised from modest to aggressive levels. Design margins are ensured since operation will be adequate at a wall loading of 2 MW/m 2 . The device has a major radius of 1.07 m, a minor radius of 0.77 m for an aspect ratio of 1.4, an elongation of 3, triangularity of 0.6 and has a naturally diverted plasma. It can be driven with neutral beams (NB) or ICRF. In the NB driven case, the plasma current is 11.1 MA, and in the ICRF case it is 13.2 MA. The magnetic field at the plasma major radius is 2.13 T and 2.55 T for the NB and the ICRF cases, respectively. This ST-VNS utilizes a single turn unshielded normal conducting CP fabricated from dispersion strengthened Cu, is 15.5 m long, has a diameter of 0.55 m at the midplane, a diameter of 1.2 m at the extremities and, in the ICRF driven case, carries a current of 13.6 MA. For the same case, the ohmic heating at the start of operation is 153 MW, increasing to 178 MW after 3 full power years. Thermal hydraulics, stress analysis and effects of transmutation on the CP have been determined, and a procedure for replacing the CP with minimal down time is described.

The tokamak systems code (SuperCode) is used to identify lower-cost ITER options. Superconducting coil, lower-cost options are found by: (1) reducing the ITER technical objectives (e.g., driven bur...

Progress is given on the investigation of a low cost, scientifically attractive, and technologically feasible volumetric neutron source (VNS) based on the spherical torus (ST) concept. The ST-VNS h...

The National Spherical Tokamak Experiment (NSTX) is an ultra low aspect ratio device with a plasma current of 1 MA. The tokamak features auxiliary heating and current drive with a close-fitting conducting shell to maximize the plasma pressure. NSTX is designed for an experimental pulse length that will demonstrate quasi-steady state non-inductively driven advanced tokamak operation. The design also takes maximum advantage of existing facilities and components from previous Princeton devices to reduce the overall program costs.

Control of the poloidal field (PF) in the Tokamak Physics Experiment (TPX) is critical to achieving its mission of advanced tokamak research. Extensive examination of the plasma equilibrium; plasma start-up; plasma position, shape, and current control; and plasma shape reconstruction have been performed as part of the design process. This paper reports the progress in this area. The PF coils have been designed to produce a wide range of plasmas. Plasma start-up can be achieved for multiple conditions. Fast plasma position control coils inside the vacuum vessel are used for short timescale control of the plasma vertical and radial position. Shape and total plasma-current control are provided by the PF coils over a slower timescale. A new algorithm for shape control of a few critical plasma boundary points is described and used in simulations using the Tokamak Simulation Code. Fast magnetostatic reconstruction of the plasma shape is examined to determine the impact of measurement locations and their quality.

The Tokamak Physics Experiment is designed to develop the scientific basis for a compact and continuously operating tokamak fusion reactor. It is based on an emerging class of tokamak operating modes, characterized by beta limits well in excess of the Troyon limit, confinement scaling well in excess of H-mode, and bootstrap current fractions approaching unity. Such modes are attainable through the use of advanced, steady state plasma controls including strong shaping, current profile control, and active particle recycling control. Key design features of the TPX are superconducting toroidal and poloidal field coils; actively-cooled plasma-facing components; a flexible heating and current drive system; and a spacious divertor for flexibility. Substantial deuterium plasma operation is made possible with an in-vessel remote maintenance system, a lowactivation titanium vacuum vessel, and shielding of ex-vessel components. The facility will be constructed as a national project with substantial participation by U.S. industry. Operation will begin with first plasma in the year 2000.

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. Four ARIES visions are currentl...

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. Three ARIES visions are currently planned for the ARIES program. The ARIES-I design is a DT-burning reactor based on modest extrapolation from the present tokamak physics data base; ARIES-II is a DT-burning reactor which will employ potential advances in physics; and ARIES-III is a conceptual D-3He reactor. The first design to be completed is ARIES-I, a 1000 MWe power reactor. The key features of ARIES-I are: (1) a passively safe and low environmental impact design because of choice of low activation material throughout the fusion power core, (2) an acceptable cost of electricity, (3) a plasma with performance as close as possible to present-day experimental achievements, (4) a high performance, low activation, SiC composite blanket cooled by He, and (5) an advanced Rankine power cycle as planned for near term coal-fired plants. The ARIES-I research has also identified key physics and technology areas with the highest leverage for achieving attractive fusion power system.

The plasma shaping flexibility of the Compact Ignition Tokamak (CIT) poloidal field (PF) coil set is demonstrated through MHD equilibrium calculations of optimal PF coil current distributions and their variation with poloidal beta, internal inductance, plasma 95% elongation, and 95% triangularity. Calculations of the magnetic stored energy are used to compare solutions associated with various plasma parameters. The Control Matrix (CM) equilibrium code, together with the nonlinear equation and numerical optimization software packages HYBRD, and VMCON, respectively, are used to find equilibrium coil current distributions for fixed divertor geometry, volt-seconds, and plasma profiles in order to isolate the dependence on individual parameters. A reference equilibrium and coil current distribution are chosen, and correction currents dI are determined using the CM equilibrium method to obtain other specified plasma shapes. The reference equilibrium is the {kappa} = 2 divertor at beginning of flattop (BOFT) with a minimum stored energy solution for the coil current distribution. The pressure profile function is fixed.

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

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 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

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 requirements of (1) an external poloidal-field coil (PFC) system with minimum stored energy, (2) double-null divertor plasmas with elongated D shape, (3) adequate passive stabilization of plasma vertical displacement by a vacuum vessel located behind the blanket zone, and (4) an enhanced plasma beta limit in the first stability regime are incorporated in the Advanced Reactor Innovation and Evaluation Study-I (ARIES-I) concept for a high-field tokamak reactor with high aspect ratio (A=4.5). The plasma current and pressure profiles are also made consistent with enhanced bootstrap current and reduced current drive power by means of ion cyclotron wave or neutral beam injection. These lead to plasmas characterized by an elongation of 1.8 to the divertor X-point, a triangularity of 0.7, a safety factor of approximately 1.5, a safety factor at the edge >4, a plasma beta of approximately 2%, and a poloidal beta of approximately 0.5. With a plasma current of 11 MA, a toroidal field of 13 T at the major radius of 6.5 m, and over 3.5 m of clearance between the PFCs and the plasma edge, the stored energy in the PFC system ranges from 20 GJ during plasma operation at low beta, to 12 GJ during plasma operation at high beta. >

The free-boundary equilibrium code VEQ provides equilibrium data that are used by the Tokamak Simulation Code (TSC) in designing and analyzing the poloidal-field (PF) system for the Compact Ignition Tokamak (CIT). VEQ serves as an important design tool for locating the PF coils and defining coil current trajectories and control systems for the TSC. VEQ and its role in the TSC analysis of the CIT PF system are described. >

A description of the ARIES-III research effort is presented, and the general features of the ARIES-III reactor are described. The plasma engineering and fusion-power-core design are summarized, including the major results, the key technical issues, and the central conclusions. Analyses have shown that the plasma power-balance window for D-/sup 3/He tokamak reactors is small and requires a first wall (or coating) that is highly reflective to synchrotron radiation and small values of tau /sub ash// epsilon /sub e/ (the ratio of ash-particle to energy confinement times in the core plasma). Both first and second stability regimes of operation have been considered. The second stability regime is chosen for the ARIES-III design point because the reactor can operate at a higher value of tau /sub ash// tau /sub E// tau /sub E/ approximately=2 (twice that of a first stability version), and because it has a reduced plasma current (30 MA), magnetic field at the coil (14 T), mass, and cost (also compared to a first-stability D-/sup 3/He reactor). The major and minor radii are, respectively 7.5 and 2.5 m. >

A control matrix MHD (magnetohydrodynamic) equilibrium code, BEQ, and the Tokamak Simulation Code (TSC) are used to compute programmed double-null (DN) divertor sweep trajectories that maximize sweep distance while simultaneously satisfying a set of strict constraints: minimum lengths of the field lines between the X-point and strike points, minimum spacing between the inboard plasma edge and the limiter, maximum spacing between the outboard plasma edge and the ICRF (ion cyclotron range of frequencies) antennas, minimum safety factor, and linked poloidal flux. A sequence of DN diverted equilibria and a consistent TSC fiducial discharge simulation are used in evaluating the performance of the BPX (Burning Plasma Experiment) divertor shape and possible modifications. >

Physics and systems design analyses are carried out to estimate the desired fusion core parameters for an ST based Volume Neutron Source (ST-VNS) that utilizes a single-turn toroidal field coil (TFC). A design with a major radius R/sub 0/=1.07 m is estimated to have large margins in physics, technology, and engineering for the initial operation at moderate performance (neutron wall load, W/sub /spl lambda//=0.5-2.0 MW/m/sup 2/). The VNS therefore begins with technologies already assumed in the ITER EDA, and the relatively conservative physics to be tested initially by the ST proof-of-principle experiments presently being built. Given continued advances in technology via the VNS and in physics via the ST experiments, the design should permit upgrades to test components and operation at the level of future Pilot Plant and Power Plants (W/sub L/=5 MW/m/sup 2/). This approach to VNS places premium on modular components and remote maintenance, encourages continued innovation and optimization in ST fusion and plasma science, and enhances the practicality of the ST pathway to fusion power.

The mechanical design of a spherical torus based volumetric neutron source (ST VNS) is being studied under the support of a DOE-SBIR funding. A device capable of staged operation from a neutron wall loading of 0.5-5.0 MW/m/sup 2/ has been scoped out, as the physics and engineering design assumptions are raised from modest to aggressive levels. Margins in the design are ensured since operation of the VNS will be adequate at a wall loading of 2 MW/m/sup 2/. The device has a naturally diverted plasma with major radius of 1.07 m, a minor radius of 0.77 m for an aspect ratio of 1.4, an elongation of 3 and triangularity of 0.6. In the neutral beam driven version, the plasma current is 11.1 MA and the toroidal field at the plasma major radius is 2.13 T. The baseline fusion power is 151 MW giving an average neutron wall loading of 2 MW/m/sup 2/ on the outboard side over an accessible area of over 15 m/sup 2/ for blanket testing. The device utilizes a normal Cu conducting bell jar as the return leg of the toroidal field current, a concept developed at the Oak Ridge National Laboratory, The current is carried by an unshielded single-turn center post (CP) made of dispersion strengthened Cu which is cooled by water in a single pass from top to bottom. A special sliding electrical interface between the CP and the bed jar is provided on the upper end to allow for differential expansion and to isolate the CP from tensile and torsional forces from the bell jar. The ohmic heating in the CP is 153 MW at the start of operation and increases to 178 MW after 3 full power years of operation. Over this period the maximum Cu temperature does not exceed 160 C. This report primarily deals with the design of the CP, one of the most challenging issues of a low aspect ratio spherical torus. Maintenance approaches for the CP and the divertor assemblies have been determined and are addressed in the paper.

Magnetic islands in free-boundary stellarator equilibria are suppressed using a procedure that iterates the plasma equilibrium equations and, at each iteration, adjusts the coil geometry to cancel resonant fields produced by the plasma. The coils are constrained to satisfy certain measures of engineering acceptability and the plasma is constrained to ensure kink stability. As the iterations continue, the coil geometry and the plasma simultaneously converge to an equilibrium in which the island content is negligible. The method is applied with success to a candidate plasma and coil design for the National Compact Stellarator eXperiment [Physics of Plasma, 7 (2000) 1911].

Magnetic islands in free-boundary stellarator equilibria are suppressed using a procedure that iterates the plasma equilibrium equations and, at each iteration, adjusts the coil geometry to cancel resonant fields produced by the plasma. The coils are constrained to satisfy certain measures of engineering acceptability and the plasma is constrained to ensure kink stability. As the iterations continue, the coil geometry and the plasma simultaneously converge to an equilibrium in which the island content is negligible. The method is applied with success to a candidate plasma and coil design for the National Compact Stellarator Experiment [Phys. Plasmas 8, 2083 (2001)]].

Dynamic stress analysis of the International Thermonuclear Experimental Reactor vacuum vessel loaded by disruption forces was performed. The deformation and stress results showed strong inertial effects when compared to static analyses. Maximum stress predicted dynamically was 300 MPa, but stress shown by static analysis from loads at the same point in time reached only 80 MPa. The analysis also provided a reaction load history in the vessel`s supports which is essential in evaluating support design. The disruption forces were estimated by assuming a 25-MA plasma current decaying at 1 MA/ms while moving vertically. In addition to forces developed within the vessel, vertical loadings from the first wall/strong back assemblies and the divertor were applied to the vessel at their attachment points. The first 50 natural modes were also determined. The first mode`s frequency was 6.0 Hz, and its shape is characterized by vertical displacement of the vessel inner leg. The predicted deformation of the vessel appeared similar to its first mode shape combined with radial contraction. Kinetic energy history from the analysis also correlated with the first mode frequency.

The free-boundary equilibrium code VEQ provides equilibrium data that are used by the Tokamak Simulation Code (TSC) in design and analysis of the poloidal field (PF) system for the Compact Ignition Tokamak (CIT). VEQ serves as an important design tool for locating the PF coils and defining coil current trajectories and control systems for TSC. In this report, VEQ and its role in the TSC analysis of the CIT PF system are described. Equilibrium and coil current calculations are discussed, an overview of the CIT PF system is presented, a set of reference equilibria for modeling a complete discharge in CIT is described, and the concept of a plasma shape control matrix is introduced. 9 refs., 8 figs., 7 tabs.

Plasmas with volume-averaged beta equal to 3% have been produced in beam-heated ISX-B discharges. Previous modelling of such ISX-B plasmas, at somewhat lower beta and with limited detail as regards plasma shape, has shown that high-n ballooning instabilities should be excited, although none is observed. The authors now use data that allow them to specify free-boundary equilibria, which are influenced by the ISX-B poloidal coil system and iron core, and compare these equilibria with criteria for ideal magnetohydrodynamic (MHD) stability to both low- and high-n number ballooning and kink modes. Both analytic estimates and codes are used in the comparison. It is found that the equilibria studied are predicted to be unstable to both low- and high-n modes, in contrast to the experimental results. Several candidates for resolving these differences are discussed.

The Compact Ignition Tokamak (CIT) is a proposed modest-size ignition experiment designed to study the physics of alpha particle heating. The basic concept is to achieve ignition in a modest-size minimum cost experiment by using a high plasma density to achieve nT{sub E} {approx} 2 x 10{sup 20}s/m{sup 3} required for ignition. The high density requires a high toroidal field (10 T). The high toroidal field allows a large plasma current (10 MA) which provides a high level of ohmic heating, improves the energy confinement, and allows a relatively high beta ({approx} 6%). The present CIT design also has a high degree of elongation (k {approx} 1.8) to aid in producing the large plasma current. A double null poloidal divertor and pellet injection are part of the design to provide impurity and particle control, improve the confinement, and provide flexibility for improving the plasma profiles. Auxiliary heating is expected to be necessary to achieve ignition, and 10-20 MW of ICRF is to be provided.

Trade studies are performed to determine the optimum plasma elongation for a next-step tokamak such as the International Thermonuclear Experimental Reactor (ITER). Degradations of the plasma beta limit for high elongations and poloidal field coil scaling with elongation are included in the analysis. When plasma ignition is required using confinement scalings that include direct plasma current or power degradation terms, the optimum elongation is between 2.5 and 2.9, but generally the minimum-cost curve is relatively flat for elongations over 2.3. When confinement scalings that depend only on size are used or when only current drive performance is required, the optimum elongation is near 2.3. Also, when only a plasma current and neutron wall load are used as plasma performance limits, the optimum elongation is between 2.6 and 2.8, but with small cost benefits above elongations of 2.3.

The flux-conserving tokamak model suggested that rapid heating would yield equilibria with high relative energy density ($\ensuremath{\beta}=\frac{2p}{{B}^{2}}$) while nonetheless allowing control over $q$, the so-called safety factor for instability within the ideal magnetohydrodynamic plasma model. In this study, we show that this is adequate to provide stability to $\ensuremath{\beta}$ values of 10%, if there is a superconducting metal shell in the vicinity of the plasma.