B.A. Smith

The Levitated Dipole Experiment (LDX) magnet system

In the Levitated Dipole Experiment (LDX), a hot plasma is formed about a levitating superconducting dipole magnet in the center of a 5 m diameter vacuum vessel. The levitated magnet is suspended magnetically during an eight hour experimental run, then lowered and recooled overnight. The floating F-coil magnet consists of a layer-wound magnet with 4 sections, designed to wrap flux lines closely about the outside of the levitated cryostat. The conductor is a niobium-tin Rutherford cable, with enough stabilizer to permit passive quench protection. Lead strips are used as thermal capacitors to slow coil heating. An optimized system of bumpers and cold-mass supports reduces heat leak into the helium vessel. Airbags catch the floating coil on quenches and faults, preventing collision with the vacuum vessel.

Design, fabrication and test of the react and wind, Nb3Sn, LDX floating coil

The Levitated Dipole Experiment (LDX) is an innovative approach to explore the magnetic confinement of fusion plasma. A superconducting solenoid (floating coil) is magnetically levitated for up to 8 hours in the center of a 5-meter diameter vacuum vessel. The floating coil maximum field is 5.3 T, and a react-and-wind Nb/sub 3/Sn conductor was selected to enable continued field production as the coil warms from 5 K during the experiment up to a final temperature of about 10 K. The coil is wound using an 18-strand Rutherford cable soldered into a half-hard copper channel, and is self protected during quench. The coil is insulated during winding and then vacuum impregnated with epoxy. The impregnated coil is tested with 2 kA operating current at 4.2 K, and then a single, low resistance joint is formed at the outer diameter of the coil before the coil is enclosed in its toroidal helium vessel. This paper presents details of the coil design and manufacturing procedures, with special attention to the techniques used to protect the coil from excessive strain damage throughout the manufacturing process.

Design, fabrication and test of the react and wind, Nb/sub 3/Sn, LDX floating coil conductor

The Levitated Dipole Experiment (LDX) is a novel approach for studying magnetic confinement of a fusion plasma. In this approach, a superconducting ring coil is magnetically levitated for up to 8 hours a day in the center of a 5 meter diameter vacuum vessel. The levitated coil, with on-board helium supply, is called the floating coil (F-Coil). Although the maximum field at the coil is only 5.3 tesla, a react-and-wind Nb/sub 3/Sn conductor was selected because the relatively high critical temperature will enable the coil to remain levitated while it warms from 5 K to 10 K. Since prereacted Nb/sub 3/Sn tape is no longer commercially available, a composite conductor was designed that contains an 18 strand Nb/sub 3/Sn Rutherford cable. The cable was reacted and then soldered into a structural copper channel that completes the conductor and also provides quench protection. The strain fabrication steps such as: soldering into the copper channel, spooling, and coil winding, to prevent degradation of the critical current. Measurements of strand and cable critical during state of the cable was continuously controlled currents are reported, as well as estimates fabrication, winding and operating strains on critical current.

Charging magnet for the floating coil of LDX

The charging coil (C-coil) for the joint Columbia University/MIT Levitated Dipole Experiment (LDX) is under development jointly by MIT and the Efremov Institute. The NbTi superconducting C-coil serves to charge/discharge inductively the floating superconducting magnet to/from 2277 A when it is resting in the charging port at the bottom of the LDX vacuum vessel. The C-coil is designed for 3200 charge-discharge cycles. The solenoid magnet is installed in a low heat leak liquid helium cryostat with a warm bore of more than 1 m. The magnet protection system has an external dump resistor, which dissipates most of the 12 MJ stored during a quench.

High temperature superconducting levitation coil for the Levitated Dipole Experiment (LDX)

The Levitated Dipole Experiment (LDX) is an innovative approach to explore the magnetic confinement of fusion plasmas. A superconducting solenoid (floating coil) is magnetically levitated for up to 8 hours in the center of a 5-meter diameter vacuum vessel. This coil is supported by a levitating coil (L-Coil) on top of the vacuum vessel. In the initial machine design, this levitating coil was a water-cooled copper solenoid, and was the experiments single largest load on the available water system. The main benefit of using a high temperature superconducting coil is the ability to apply more auxiliary heating power to the plasma. However, this coil will also be the first high temperature superconducting coil to be used in a US fusion program experiment. The high temperature superconducting L-Coil is a solenoid, using a two-in-hand winding of a commercially available 0.17 mm/spl times/3.1 mm tape by American Superconductor Corporation with a critical current of 62 A at 77 K and self-field. The L-Coil will be operated at 0.9 T and 20 K. The L-Coil has a protection circuit that not only protects it against overheating in the event of quench, but also against F-Coil collision in the event of a control failure.

Design and manufacture of the US-ITER pre prototype joint sample

The US-ITER pre prototype joint sample which has been fabricated at the MIT Plasma Fusion Center is the first attempt to fabricate an optimized full size joint which can be stably operated in ITER required AC background fields at reduced coupling losses. This paper presents an overview of the joints construction and fabrication, highlighting some of the procedural steps that have since been incorporated into fabrication of current terminations for the inner module ITER central solenoid (CS) model coil.

Design concept for the GEM detector magnet

The magnet has two symmetric and independent halves, each containing a cold mass assembly operating nominally at 4.5 K, a set of vapor cooled leads, a cold mass support system, a liquid nitrogen shield system, and a vacuum vessel. Also included in each half is a forward field shaper which provides a component of magnetic induction normal to the path of low angle muons in the forward region, thereby improving their resolution. The unique features of this magnet are the conductor design itself and the large coil diameter, which demands an on-site winding and assembly operation. The use of a natural convection thermosiphon loop for thermal radiation cooling eliminates plumbing complications. Locating the aluminium sheath outside the conduit for quench protection permits optimizing the copper-to-superconductor ratio inside the conduit for stability alone. The conceptual design for the magnet, including the design for the detector dependent magnetics, the superconducting coils and coil structure (cold mass), the coil winding process, the vacuum vessel and liquid nitrogen shields, the cold mass supports, and the magnet assembly procedure, are described.<<ETX>>

Superconducting magnets for Maglifter launch assist sleds

The Maglifter is an electromagnetic catapult being considered by NASA to reduce the cost of lifting a payload into space. The system would accelerate a vehicle of up to 590 tonnes to a final velocity of 268 m/s at an acceleration of 2 g. Superconducting coils are considered for levitation because they permit track-to-vehicle clearances of more than 95 mm. The high clearances reduce tolerances and maintenance costs, and allow a system with permanently deployed wheels for take- off and emergency landing. Cable-in-conduit conductors (CICC) were selected because of their high electrical and mechanical strength, as well as high energy margin for stability. The selected coil shape is a pair of racetrack coils forming a module with four modules on a sled. The superconducting levitation modules weigh about 4% of the gross lift off weight and are capable of achieving lift off at about 20 m/s. The maximum magnetic drag power is negligible compared to the power required for acceleration.

A Comparison of the Quench Analysis on an Impregnated Solenoid Magnet Wound on an Aluminum Mandrel Using Three Computer Codes

The magnet used for the quench protection comparison has an ID of 1.5 m. At a maximum current of ~ 210-A, the stored energy is ~13 MJ. The impregnated magnet coil is 281 mm long and about 105.6 mm thick. The coil is wound on a 6061-aluminum mandrel. The magnet quench protection system is passive. The magnet coil is subdivided with back-to-back diodes and resistors across each of the coil subdivision to reduce the magnet internal voltages. Conservative quench protection criteria were applied when the magnet was designed. These criteria are presented in this paper. Quench protection of the magnet was simulated using three computer codes from three different places. The results calculated using the three codes are compared to the original magnet quench protection criteria used to design the magnet. The three quench simulation codes assumptions are compared. The calculated hot-spot temperature and peak voltages are compared for the three quench simulation codes.

PTF, a new facility for pulse field testing of large scale superconducting cables and joints

A magnetic Pulse Test Facility (PTF), in which samples of CICC electrical joints from each ITER home team will be tested, has been fabricated at the MIT Plasma Fusion Center under an ITER task agreement. Construction of this facility has recently been completed, and an initial test phase on the first CICC joint sample has begun. PTF includes capabilities for sample currents up to 50 kA from a superconducting transformer developed by the University of Twente, magnetic fields up to 6.6 T with ramp rates to +1.5 T/s and -20 T/s, and a cryogenic interface, supplying supercritical helium with flow rates to 20 g/s through each CICC leg at controlled temperatures to 10 K and pressures to 10 atmospheres. A sophisticated, multiple-channel data acquisition system is provided to processed, digitally recorded sensor signals from both the sample and the facility. The facility is totally remote-controlled from a control room through a fiber optic link, and qualified users worldwide are afforded secured access to test data on a 24-hour basis via the Internet. The facility has successfully exercised the first joint sample over the ITER test spectrum with positive results.