Search

The low aspect ratio spherical tokamak (ST) MEDUSA-CR is currently being re-commissioned at Instituto Tecnologico de Costa Rica. One of the proposed first tasks is to simulate the magnetic confinement considering the particular coils arrangement of the device. This work presents the resulting shape of the plasma cross-section. MEDUSA-CR main specifications are: plasma major radius R0 approx. 0.14 m, plasma minor radius a approx. 0.10 m, toroidal field at the vessel geometrical center BT smaller than 0.5 T, plasma current Ip lesser than 40 kA, central electron temperature Te lesser than 140 eV, discharge duration lesser than 3 ms [1]. The free boundary equilibrium solver Fiesta has been used to obtain the plasma shaping and some equilibrium parameters. The code works the magnetic confinement for the static Magneto-Hydro-Dynamic (MHD) equilibrium case solving the Grad-Shafranov equation. A central solenoid was added to the original magnetic configuration aiming to create a limiter with an ergodic behavior in order to enhance the plasma confinement [2], [3]. The resulting cross-section (without divertor, i.e., natural divertor) for the plasma volume is a “bean” shape. The elongation is found to be 1.43 and the triangularity 0.547. A bean shape cross section is found to produce a higher triangularity, which have a significant effect on MHD stability [4]. Further simulations with FIESTA could be used as proof of principle for the effect of the plasma shape on transport and stability.

The EDT2 750V uses a micro pattern trench cell with a narrow mesa for reducing the on-state losses with a tailored channel width for short circuit robustness. To account for high system stray inductances (L stray ) and currents for Full or Hybrid Electric Vehicle inverter applications, it features a 750V voltage rating compared to the predecessor IGBT3 650V by an optimized vertical structure and proper plasma shaping. This plasma distribution not only determines the performance tradeoff between on-state and switching losses, but at the same time defines the surge voltage for a given L stray ∗I in the application as visualized in a switch-off loss vs. surge voltage trade-off diagram. Shaping of the feedback capacitance Cgc optimizes the tunability of the switching slopes by means of an external gate resistor for an easier adaption to a wider range of system inductances with low losses.

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 the first of a new class of stellarators known as "compact stellarators". Stellarators are characterized by three dimensional magnetic fields and plasma shapes and are the best-developed class of magnetic fusion devices after the tokamak. Stellarators are attractive because they solve critical problems of magnetic fusion energy: steady state operation without current drive and stable operation without feedback control or rotation drive. The differentiating feature of the compact stellarator is the use of a small plasma current in conjunction with external magnetic fields to provide the required plasma shaping and confinement. This permits a more compact stellarator design. Many of the components of NCSX are conventional in design and manufacture; however, two of the most critical components, the vacuum vessel and the modular coil winding forms, are also the most challenging components to manufacture. This paper describes design of these two critical elements of NCSX and the three-phase program that is being used in the manufacturing development of these components.

The superconducting tokamak project HT-7U, aiming at steady-state advanced operation mode, will make a contribution to future steady-state tokamak reactors. The scientific and the engineering missions of the project are to study physics issues of the steady-state tokamak operation and to establish technology basis of full superconducting tokamaks. It features: superconducting toroidal field system and poloidal field system, non-inductive current drive and plasma heating systems, flexibility and reliability of plasma shaping control, J(r) and P(r) control, replaceability of plasma facing components and divertors for power and particle handling study in steady-state operation and advanced diagnostic measurements. The physics design and the engineering design have been completed essentially. The key R&D programs of the tokamak device have been successful. The assembly of the device has begun. It is planned to obtain the first plasma in 2005.

The National Compact Stellarator Experiment (NCSX) is an experimental device whose design and construction is underway at the Department of Energy's Princeton Plasma Physics Laboratory (PPPL). The primary coil systems for the NCSX device consist of the twisted plasma shaping modular coils, the poloidal field coils, and the toroidal field (TF) coils. The TF coils are D shaped coils wound from hollow copper conductor and vacuum impregnated with a glass-epoxy resin system. There are 18 identical, equally spaced TF coils providing 1/R field at the plasma. They operate within a cryostat and are cooled by LN2 nominally to 80 K. Wedge shaped castings are assembled to the inboard face of these coils so that inward radial loads are reacted via the nesting of each of the coils against their adjacent partners. This paper outlines the TF coil design methodology, reviews the analysis results and summarizes how the design and analysis support the design requirements

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.

Recent experimental results in the fusion community have pointed to possible directions for improvement of the tokamak concept. A critical issue for many advanced tokamak scenarios is the optimization of the spatial distribution of plasma current density. Recent experiments on the Tokamak Fusion Test Reactor and Doublet III-D have demonstrated enhanced confinement and improved stability using transient techniques for the modification of the plasma current profile. The Princeton Beta Experiment-Modification (PBX-M) has begun experiments using profile modification techniques that are extrapolable to long-pulse or steady-state operation. PBX-M uses lower hybrid current drive for current profile modification, ion Bernstein wave heating for pressure profile variation, and plasma shaping to alter the magnetic geometry. This paper is an overview of the PBX-M program, its engineering facilities, and discussion of how this program is relevant to future advanced tokamak programs such as the Tokamak Physics Experiment.

The Burning Plasma Experiment (BPX) poloidal field (PF) system is required to support a maximum plasma current of 10.6 MA while providing plasma position and shape control. The plasma shaping control scenarios of double null, asymmetrical single null, and limiter operations must be accommodated. The coil power is to be supplied by existing power supplies which are currently in use on the Tokamak Fusion Test Reactor (TFTR). The power supplies derive their power from two existing variable frequency TFTR motor generator sets. An additional requirement of the system is the ability to accommodate an upgrade of the maximum plasma current pulse from 10.6 MA to 11.8 MA with minimum impact on tokamak operation. The authors describe and analyze the various PF coil power system requirements, the switching circuits used to accomplish the required coil scenarios, and the optimal allocation of the TFTR power supplies to meet the base requirements and anticipated upgrade requirements. The results from the simulation studies which verify the operation of the PF power system design are also presented. >

The highly flexible plasma shaping and diverter configuration of Alcator C-MOD necessitates the use of many separate magnet power supply systems. Eleven separate systems are used to power the 14 magnets in the tokamak. The severe budgetary constraints prevented the purchase of an entirely new converter set for each of these 11 systems. The solution to this dilemma was the long-term loan of the TMX Upgrade magnet power supplies from Lawrence Livermore National Laboratory to MIT. The converters operate in groups of two or three to power four of the 10 poloidal field magnet systems and to provide excitation to the alternator. The 'TMX supplies' have been completely overhauled and redesigned for their use in the Alcator program. All have new control logic, PLC interface, and gate drive circuits. Several have been modified to provide two-quadrant operation. The AC and DC bus has been redesigned and simplified. Twelve of the units are mounted in a double-high rack to preserve floor space. The AC unit substations are located near the converters to minimize resistive and reactive losses. Some test data from rebuilt supplies are presented along with details of the interface to the Alcator control system. All power supplies have been pulse tested to 300 V at 5 kA into an inductive load. >

During the last two years four different Poloidal Field configurations have been analyzed in detail and compared. All configurations studied were more or less equivalent from the point of view of plasma performance, two were ruled out for mechanical problems. The PF system, selected as reference, has nine PF coils, all using NbTi superconductors.

The KSTAR (Korea Superconducting Tokamak Advanced Research) project is the major effort of the Korean National Fusion Program to design, construct, and operate a steady-state-capable superconducting tokamak. The project is led by Korea Basic Science Institute and shared by national laboratories, universities, and industry along with international collaboration. It is in the conceptual design phase and aims for the first plasma by mid 2002. The key design features of KSTAR are: major radius 1.8 m, minor radius 0.5 m, toroidal field 3.5 T, plasma current 2 MA, and flexible plasma shaping (elongation 2.0; triangularity 0.8; double-null poloidal divertor). Both the toroidal and the poloidal field magnets are superconducting coils. The device is configured to be initially capable of 20 s pulse operation and then to be upgraded for operation up to 300 s with non-inductive current drive. The auxiliary heating and current drive system consists of neutral beam, ICRF, lower hybrid, and ECRF. Deuterium operation is planned with a full radiation shielding.

The SST-1 tokamak is being designed for steady state operation with plasma durations in the range of 100-1000 s. Both the toroidal field coils (TFC) and the poloidal field coils (PFC) in the SST-1 tokamak are superconducting. TF coils are required to produce a magnetic field of 3 T at 1.05 m from machine axis. The maximum field seen by the TF coil conductor will be /spl les/4.3 T. The cable for the TF coils has, therefore, been designed for a field of 5 T at the conductor. The PFC are used for plasma shaping and equilibrium and the magnetic field on the PFC conductor is estimated to be /spl les/3.2 T. Hence a cable designed for 5 T operation will be suitable for PFC also. Subjected to the stability against the disturbances generated by the current ramping in the PFC. We have designed two Cable-in-conduit-conductor (CICC) type cables, one with copper conduit (to be preferably used with the TFC) and other with Stainless Steel conduit to be used with PFC. In the following we describe some design aspects' of these cables and discuss the stability of these cables against disturbances.

Summary form only given. Plasmas become strongly coupled (SCP) when the coulomb interaction energy between particles is comparable or greater than the thermal kinetic energy. This occurs at low temperatures (T/spl sim/1 eV) and high densities (n/sub c//spl sim/10/sup 21/ cm/sup -3/). SCPs are found in the interior of gas planets and white dwarf stars, as well as in high energy density physics experiments. We present a pulsed power and plasma shaping system designed to generate an aluminum SCP with a rectangular column profile. A 3.5 kJ, 4-stage Marx has been constructed to explode 200-micron aluminum wires fused in lead glass (/spl rho//spl sim/5.4 g/cm/sup 3/). The glass tamper restricts the exploding wires radial expansion, while axial expansion is permitted. The central region of the expanding plasma passes through a 100-micron square aperture assembly creating a plasma column. Above the aperture, the plasma plume is expected to have a density of one-tenth solid (/spl rho//spl sim/0.27 g/cm/sup 3/) and a temperature of the order I eV. Absorption of magnesium K-shell x-rays from a laser-produced backlighter is used to measure the density of the plasma. An optical pyrometer is used to estimate the plasma surface temperature based on a blackbody approximation with constant emissivity. Raven and Crunch were used to simulate the behavior inside the tamped region using conductivity tables generated by Desjarlais. The simulations predict fairly uniform density and temperature conditions in the central 100-microns of the tamped exploding wire. The plasma plume above the aperture will be used for equation of state studies of SCP matter.

Summary form only given. The PBX-M physics program exploits strong plasma shaping, control of the pressure and current profiles, and a close-fitting conducting wall for the understanding and optimization of plasma stability and confinement. Transport studies and ITER and CIT RaD-relevant studies of high-performance plasmas have focused on: (1) high-beta (lbr~6.8%) plasma with good energy confinement and high-temperature; (2) high-beta relative to the Troyon limit: (3) highly-indented ( i ~30%) plasmas. using plasma current ramps to drive the necessary edge current; (4) H-mode phases in both limited and diverted discharges; (5) current profile control by current ramping, counterneutral beam injection, and pellet injection: (6) development of j ( r ) diagnostics and initial studies of q ( r ) effects on sawteeth and fishbones: (7) modifications of disruption precursor growth rates and mode structure; and (8) disruptive vertical displacement events