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The calculations results of the maximal temperature heating of Tokamak-15 superconducting magnet (T-15 SM) under energy removal dependent on the current through the coil are given in paper. The calculations of SM thermomechanical strength have shown that the maximal coil heating temperature should not exceed 150-160 K. The range of the settings levels in SM protection system for currents I [less than or equal to] 4kA has been determined.

The 'GLOBUS-M' is a compact resistive tokamak with a central solenoid (CS) wound around the inner portion of the toroidal field coils. The magnetic field at the solenoid axis amounts to 8.3 T. The CS incorporates two layers of conductor (CuCr copper alloy) baked into insulation. The solenoid is designed to sustain 80.000 energizing. During each loading cycle the solenoid is subjected to the radial forces accompanied with the vertical compression. The most loaded region has been considered and modeled with the use of 2D axisymmetric finite element (FE) model. The model includes two conductor turns baked into insulation compound, copper cooling tubes and solder. The stress analysis shows that there is some plastic deformation in the copper tube and solder during loading and there is some back plastic deformation in the solder during unloading. The reloading does not cause any change in the solenoid stress-strain state in comparison with the case of loading. The number of cycles to failure has been simulated for all metallic components of the solenoid.

We have been developing a new type of [Nb.sub.3] Sn superconductor, a fiber-reinforced superconductor (FRS), for application in high-field pulsed superconducting magnets. This paper describes an apparatus for preparing [Nb.sub.3] Sn FRS. The apparatus has a pair of spools and opposed planar targets to produce thin superconducting layers on reinforcement fibers of meter class. Apparatus tests were performed to examine the cross section of sputtered layers and critical current in FRS of 1 m length. The tests showed that the opposed planar targets can produce uniform sputtered layers. The critical current density measurement for the FRS samples was also described.

A prototype tokamak poloidal field coil of 3 m diameter was successfully tested. The coil was designed to withstand typical field transients like fast ramping, plasma controlling and plasma disruptions. In order to achieve these requirements a low loss conductor was developed. The coil with its components was designed according to rules of high voltage technology. A realistic test of the magnet requires at 23 kV a current of 15 kA, that means a peak power of 345 MW. Therefore, a fast switching circuit was used combined with an additional current lead at the middle of the magnet winding. This allowed the performance of fast field changes with increasing or decreasing field. In one operation mode the coil was discharged into a dump resistor. In another operation mode, only one coil half is discharged which generated a field increase in the second short circuited half of the coil. Discharging time constants of 1 ms and local field sweep rates of about 240 T/s were obtained without quench. A reproducible and sharp quench boundary was found. The results showed clearly that ramp rate limitations are only given by the expected stability limit or the load line boundary.

This paper discusses advantages and disadvantages of different locations of co-wound voltage sensors for quench detection in tokamak magnets with a cable-in-conduit conductor. The voltage sensor locations are analyzed and estimates of the anticipated noise vs. dB/dt are derived for transverse, parallel, and self fields. The LLNL Noise Rejection Experiment, also described here, is designed to verify theoretical expectations on a copper cable exposed to these fields that will simulate the tokamak field environment.

The ITER-TF coil requires a generation of 12.5 T in a size of 12 m x 18 m. For such a high field-large coil, the applicability of [Nb.sub.3]Al conductors was considered with a react-and-wind technique which realizes high reliability and low cost for the coil fabrication. The maximum bending strain on the [Nb.sub.3]Al conductor in use of this technique is 0.39 %. Ic degradation due to the strain is expected to be below 5 %. The limiting current is estimated as 65 kA for 60 kA operation current. Ac loss in the TF coil with [Nb.sub.3]Al conductor is almost the same as with [Nb.sub.3]Sn conductor.

A poloidal field model coil of 3 m in diameter was designed and built according to the following specifications: Stability was required during an exponential field variation of [Delta]B = 0.4 T with a time constant of 5 ms, corresponding to a maximum field change of 80 T/s. The superconducting cable developed for this purpose was tested in a short sample experiment, yielding a stability limit beyond this requirement. The test results of the model coil are in good agreement with the stability limit expected from the cable experiment. They can be explained in terms of a simple stability model which compares the energy input due to coupling losses with the transient heat removal capability of supercritical helium. A more detailed stability calculation was performed with a computer code taking into account the temperature dependence of helium and conductor material properties. The coil reached the stability limit expected from the measured cable parameters and showed no unexpected ramp rate limitations.

A superconducting poloidal field model coil as needed for tokamaks has been constructed by GEC ALSTHOM within the Polo project in effective collaboration with Forschungszentrum Karlsruhe, Institut fur Technische Physik, Germany. The manufacturing procedures for the coil and its terminals are described.

The Engineering Design Activity (EDA) of International Thermonuclear Experimental Reactor (ITER) has been carried out since 1992 among the four Parties which are European Union, Japan, Russian Federation, and United States. The design and the fabrication of the Center Solenoid (CS) model coil is under way in the ITER-EDA. The CS model coil consists of four parts, which are the outer module, the inner module, insert, and the supporting structure. The design of the CS insert coil is reported and discussed. The inner diameter of the CS insert coil is 1.5m. The conductor of the CS insert coil is the same as that of the ITER-CS coil. The CS insert coil will be installed inside the CS model coil. The conductor performance will be demonstrated on the maximum magnetic flux density of 13 T at the test facility of Japan Atomic Energy Research Institute (JAERI).

Japan Atomic Energy Research Institute (JAERI) has been designing and developing the Central Solenoid (CS) model coil for International Thermonuclear Experimental Reactor (ITER). The CS model coil is composed of two modules and they will be manufactured by United States Home Team and Japanese Home Team. The coil winding method is two-conductors-in-hand and the coil which has an inner diameter of 1.6 m and an outer diameter of 3.6 m, produces a maximum field of 13 T. The coil will be tested in the JAERI facility. This paper describes the coil design, manufacturing procedure, stress analysis on the winding pack and AC loss analysis.

The [Nb.sub.3]Sn cable is being fabricated for the Central Solenoid (CS) Model Coil under the ITER Engineering-Design Activity. The cable consists of about 1000 strands whose diameter is 0.81 mm. The design current is 48 kA at a magnetic field of 13T. The 0.6-GJ CS model coil is operated in a pulse mode (0.5T/s). The first trial fabrication of a 100-m dummy cable and a 20-m superconducting cable was completed successfully. The second trial fabrication of a 1000-m dummy cable was performed to establish the stable manufacturing procedure in January, 1995. We measured the AC losses of the full-sized conductor and could determine the cable coupling time constant. We analyzed the heat generation of the CS model coil and calculated the temperature rise of the cable for the model coil.

The design and fabrication techniques for the ITER magnets are supported by the ITER Model Coil Program which focuses on component development and performance verification. The key component of the magnet system is the high current capacity (45 - 60 kA), [Nb.sub.3]Sn, cable-in-conduit conductor, which is now being fabricated under a staged development program. The integrated design features and performance of the 13 T Toroidal Field coils and Central Solenoid will be verified in the Model Coil Program. This paper describes the status of the structural and construction features of the model coils and the test facilities for the model coils at JAERI (Naka) and FZK (Karlsruhe).

The general design and the associated performance analysis of a joint proposed by the European Union for both ITER Central Solenoid and Toroidal Field coils are presented. Results of an extensive testing programme performed on subsize joints (scale 1/8), including DC resistance and pulsed field losses measurements are given and discussed, as well as their interaction on the final joint design. Results of feasibility tests carried out on a full-size piece are also given. This study has finally led to the design of a full-size joint sample to be fabricated in industry and tested in specific laboratory test facilities.

Tore Supra is the largest Tokamak with superconducting coils. This unique feature enables a large part of the physics programme to be dedicated to long pulse operation (30-60 s). After a brief description of some peculiar aspects of the Tore Supra magnets, the paper reports about the 6 year operating experience. Particular emphasis is put on the mechanical behaviour of the coils under repetitive loadings, on the thermal loads and cooling systems and on the protection and safety systems. Tore Supra has now been routinely operated with a current maintained constant on a whole day basis (total accumulated time at 1.8 K of 14000 Hours) for a programme of more than 16000 shots. While 31 safety discharges have been triggered for undue reasons, only one real quench has been detected for which the protection system reacted as expected.

The ITER poloidal field (PF) coil system consists of a central solenoid coil and seven ring coils. The PF coils must provide the magnetic fields to confine the plasma and control its position during the various phase of operation including plasma initiation, ramp-up, burn and shut-down. They contribute the magnetic flux change to ramp up the plasma current and a part of the flux change to maintain it. This paper describes the requirements and the conceptual design for the seven outer PF coils.

Design of the GLOBUS-M tokamak inductor (8.3 T) supported with some mechanical studies is presented. Mechanical characteristics of the conductor candidate materials have been considered. The conductor and insulation materials have been selected on the basis of mechanical and electrical tests.

Many tokamak devices require a preionization system in order to start up both the discharge cleaning and the main plasma discharge. A stray vertical magnetic field profile of 100 G at the center of the plasma column has been measured on the NOVILLO tokamak by magnetic probes especially made to study the breakdown process. A new vertical field profile, superimposed to the stray field component parallel to the equilibrium field, is created by two additional compensation poloidal coils. Such arrangement of the vertical field profiles, with a 1/R-dependence, has been found optimum to achieve the plasma breakdown without any preionization system. Index Terms--Assisted start-up, magnetic confinement, preionization, stray magnetic fields, tokamak engineering.

Fast visible imaging of Mega-Ampere Spherical Tokamak plasmas during injection of refueling pellets of frozen deuterium can sometimes vividly reveal their internal magnetic structure. Concentrated deposition of particles around the flux surface with unit value of field-line helicity results in locally increased bremsstrahlung emission which illuminates this whole surface or, occasionally, a single flux tube within it. Analysis of such images could be exploited to check equilibrium reconstruction models. Index Terms--Ablation, imaging, magnetic confinement, magnetic surfaces, pellet fuelling, tokamaks.

The Compact Toroid (CT) Injection Experiment (CTIX) is a coaxial plasma rail gun originally designed for tokomak refueling. CTIX can form and accelerate CT plasmas of controllable densities and velocities. For a given chord through a plasma, changes in line-integrated transverse electron density gradient can be resolved by measuring the angle of deflection of a laser beam traversing the plasma. A vertical array of four temperature-controlled laser deflectometers was installed in the viewing section of CTIX and used to profile the line-averaged density of an unconstrained CT drifting at high velocity. Index Terms--Compact toroid, deflectometer, deflectometry, electron density.