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The JT-60 Super Advanced (JT-60SA) tokamak aims to perform the ITER support and to demonstrate steady-state high-beta plasma project with the collaboration between Japan and EU. To attain these objectives, the negative-ion-based NBI (N-NBI) system is required to inject 10 MW for 100 s at the beam energy of 500 keV. On JT-60U, the present N-NBI ion source has injected 3.2 MW for 21 s at 320 keV; however, three key issues should be solved for the JT-60SA N-NBI ion source. One is to improve the voltage holding capability of the large negative ion source, where the available acceleration voltage has been limited to less than ~400 kV due to breakdowns. The accelerator of the JT-60U ion source is composed of large three-stage grids and three fiberglass reinforced plastic (FRP) insulators. Recent R&D tests suggested that the FRP insulators were not the main factor to trigger the breakdowns at the early conditioning stage. The accelerator with a large area of grids and their supporting structure may need a high margin in the design of electric field and a long time for conditioning. The second issue is to reduce the power loading of the acceleration grids. It was found that some beamlets were strongly deflected due to beamlet-beamlet interaction and strike on the grounded grid in the accelerator. Moreover, the electrons generated in the accelerator caused the grid loading and the overheating of the beamline components. The acceleration grids for JT-60SA are to be designed by taking account of the beamlet-beamlet interaction and the applied magnetic field in 3-D simulation. Third is to maintain the D- production for 100 s. Although a constant D- beam power was confirmed on JT-60U for 21 s, an active cooling system is required to keep the temperature of the plasma grid (PG) under optimum condition during 100-s operation. A simple cooling structure is proposed for the active cooled PG, where a key is the temperature gradient on the PG for uniform D- production. In the present schedule, design work, reflecting the latest R&D progress, will continue until ~2011. The modified N-NBI ion source will start on JT-60SA in 2015. Index Terms--Beam optics, injection power, JT-60 Super Advanced (JT-60SA), negative ion beam, negative-ion-based NBI (N-NBI), neutral beam, voltage holding capability.

The high-energy ionized particles collected by the divertor structures in a tokamak reactor cause a localized thermal load around the strike-points, which are the intersections of the separatrix with the divertor tiles. To spread on a larger region this thermal load, it is convenient to resort to a sweeping, i.e., a periodical movement of the strike-points. This paper introduces the new model-based sweeping algorithm recently implemented on the Joint European Torus tokamak, within the eXtreme Shape Controller architecture. Index Terms--Plasma-shape control, plasma strike-point sweeping, tokamaks control.

The RF behavior of a 42-GHz 200-kW continuous-wave (CW) gyrotron for plasma heating in a small experimental tokamak is presented in this paper. The operating mode, as a matter of technical convenience, is the [TE.sub.0,3] mode, as is required by the end user. Design constraints are carefully investigated, and starting currents are computed. Cold-cavity and self-consistent calculations are carried out; power and efficiencies are computed for a typical set of beam parameters. In addition, a triode-type magnetron injection gun (MIG) has been designed, and the realistic beam parameters obtained from these MIG calculations are used in time-dependent self-consistent calculations which are carried out before and after space-charge neutralization. These studies on RF behavior suggest that it is feasible to realize CW powers in excess of 200 kW at 42 GHz with [TE.sub.0,3] as the operating cavity mode. Index Terms--Electron cyclotron resonance heating (ECRH), gyrotron, tokamak.

The NSTAB and TRAN computer codes have been developed to study equilibrium, stability, and transport in fusion plasmas with three-dimensional (3D) geometry. The numerical method that is applied calculates islands in tokamaks like the Doublet III-D at General Atomic and the International Thermonuclear Experimental Reactor. When bifurcated 3D solutions are used in Monte Carlo computations of the energy confinement time, a realistic simulation of transport is obtained. The significance of finding many 3D magnetohydrodynamic equilibria in axially symmetric tokamaks needs attention because their cumulative effect may contribute to the prompt loss of [alpha] particles or to crashes and disruptions that are observed. The 3D theory predicts good performance for stellarators. magnetic fusion | magnetohydrodynamicsi plasma physics | stellarator | weak solution

This paper presents the hardware and software approach to replacing the current analogue linear feedback system on the TCV Tokamak with a DSP based system to allow for non-linear control. After preliminary combination, over 120 control signals are reduced to 32 parameters which enter an array of 32 DSP processors each of which is passed the complete parameter set. Apart from data transfer and storage, each DSP will run its separate control algorithm with a repetition rate up to 20 kHz. A separate DSP board is dedicated to the plasma vertical control which requires a 200 kHz feedback rate. This paper presents the hardware and the software that is being designed to put this control block in the hands of non DSP-expert physics users. Index Terms--Digital signal processor (DSP), multi-channel, non-expert user interface, non-linear, RT block programming, real time digital tokamak control.

Abstract--This paper is mainly concerned with the application of a novel technique of data interpretation for classifying measurements of plasma columns in Tokamak reactors for nuclear fusion applications. The proposed [...]

Nowadays, every tokamak has a plasma control system to monitor and drive discharge parameters like position, density, current, and shape. Several different solutions have been adopted to cope with the real-time constraints, ranging from the shared memory to the transputers technologies. At present, a VME/PPC604r embedded controller running a LynxOS operating system is used on Frascati Tokamak Upgrade (FTU) for the plasma position control, with a response time of about 150 [micro]s. In this paper, a report of the tests carried out using the open source Linux real-time operating system RTAI on a VME/INTELx86 embedded controller will be presented, in order to evaluate the replacement of the existing expensive commercial solution for the plasma control system. To this aim, the plasma position control software has been ported on the open source platform. Moreover, the performance of the two real-time operating systems will be compared, both at system level and at application level. The comparison will show that the open source real-time operating system is a good alternative to the commercial product and, because of its hard real-time features, it is a well-guaranteed platform for our plasma control system software. Index Terms--Open source systems, plasma control, real time systems.

Testing of blip and cancellation coils for the Samsung Superconductor Test Facility (SSTF) for KSTAR (Korea Tokamak Advanced Research) is described and discussed.

Design of toroidal field (TF) coil for the JT-60 superconducting tokamak is described and discussed.

Development of a Cable-In-Conduit Conductor ( CICC) for the Korea Superconducting Tokamak Advanced Research (KSTAR) TF (Toroidal Field) coil system is described and discussed.

Research involving simulations of dynamic stress in PF (Polodial Field) superconducting magnets for the KSTAR ( Korea Superconducting Tokamak Advanced Reactor) is described and discussed.

To provide plasma-physics scientists with worldwide access to historical and current measurement data from the Textor-94 Tokamak experiment, we developed a new architecture: the 'WebUmbrella.' It uses HTTP to transfer data in (X)HTML, plain text, or other MIME-type documents. We provide a single point of access to our servers via HTTP redirection. Server implementations for several data stores are already present, and the first user tests are positive. In particular, we present one server implementation that provides access to already existing common object request broker architecture (CORBA) servers. It uses the Apache Web server and the Tomcat Servlets engine, and consists of Java Servlets that convert HTTP requests into CORBA requests for the existing interface. These developments are the aftermath of an earlier experience with CORBA interface definition language (IDL) interfaces between user clients and several data servers, which provide access to the different data stores. The interfaces were developed in a telematics application program project, 'Dynacore.' After the evaluation of this demonstrator, we designed a new system to improve the service. Because of the positive experience, we inherited the data representation, the data-browsing features, and the single access point for clients to several data stores. We tried to eliminate the obligatory use of CORBA, as its added value did not weigh up against the problems we faced using it: CORBA is difficult to understand for scientists inexperienced in object-oriented design and difficult to access with simple analysis tools like IDL or MATLAB. We also did not find a suitable CORBA implementation for Virtual Address eXtension/Virtual Memory System (VAX/VMS) with C/Fortran bindings. Finally, WebUmbrella's transport protocol, contrary to CORBA's, does not suffer from firewall and access problems. Index Terms--Data communication, distributed database systems, internetworking, tree data structures.

The critical current of mass-produced Nb[subscript 3]Al strand developed for the International Thermonuclear Experimental Reactor (ITER) project was measured at various temperatures and magnetic fields. An empirical equation for the critical current density was obtained from these data.

Electrode biasing experiments are carried out on the KT-5C tokamak to investigate the effects of the radial electric field [E.sub.r] on turbulence. A modestly enhanced [E.sub.r] gradient layer is formed at the plasma edge by the electrode biasing. In the [E.sub.r] gradient layer, reductions in the fluctuation amplitude, radial correlation length and turbulent particle flux are observed. These results support the theoretical models on the turbulence suppression by decorrelation due to E x B flow shear. Index Terms--Biasing, radial electric field, turbulence.

A relation between travel of the bending roller and spring back radius is described. The relation was obtained by means of virtual manufacturing with a 3-roll bending process.

MFEM1D, a numerical method based on the finite-element method, is described. It is used to investigate the helium expansion and quench propagation in cable-in-conduit conductors (CICC).

MAGS (MAGnet System), a computer code system for simulating accident scenarios of magnet systems of fusion plants, is used to perform a safety analysis for a Toroidal Field (TF) coil of the International Thermonuclear Experimental Reactor-Fusion Energy Advanced Tokamak (ITER-FEAT) design.

Thermal analysis of the Korea Superconducting Tokamak Advanced Research (KSTAR) device's poloidal field (PF) was accomplished using the code SAITOKPF

The relation between the toroidal magnetic field and stress is obtained, using the virial theorem. A new toroidal field (TF) coil for tokamaks is proposed.

Research is presented describing the eddy current and thermal analysis of the Toroidal Field Model Coil that is being manufactured in association with the International Thermonuclear Experimental Reactor project. The coil is an example of a large tokamak's toroidal field system.

A method to calculate out-of-plane shear stresses and their distributions is proposed for system codes to support Toroidal Field coils in recent tokamaks.

HT-7U tokamak magnet coils are investigated and the winding technique is examined with regard to a new winding facility with the capability for continuous winding of long length cable-in-conduit conductor.

A free electron maser (FEM) has been built as a pilot mm-wave source for applications on future fusion research devices such as international tokamak experimental reactor (ITER). A unique feature of the Dutch FEM is the possibility to tune the frequency over the entire range from 130 to 260 GHz at an output power exceeding one MW. In the first phase of the project, the so-called inverse setup is used, where the electron gun is mounted inside the high-voltage terminal. The entire beam line was tested successfully with an extremely low loss current, lower than 0.05%. First generation of mm waves was achieved in October 1997. Up to now a peak power exceeding 730 kW was measured at 200 GHz and 350 kW at 167 GHz. Output power, start-up time and frequency correspond well with simulation results. Parameter scans for the longitudinal undulator gap, accelerator voltage, and reflection coefficient have given a wide range of interesting data of which a few highlights are given. Index Terms - Free electron lasers, masers, microwave generation, microwave power transmission, microwave tubes, millimeter wave generation, millimeter wave tubes.

Research was conducted to test and compare two candidate resin systems with the ITER standard resin system. Results show that the modified resin system can be successfully applied to impregnate the KSTAR CS model coil. A sufficiently low viscosity and long useable life were also demonstrated. Findings for the mechanical properties were comparable to those of the ITER resin system particularly with the increased toughness and excellent bond to stainless steel relative to the ITER resin system.

The development of the Korea Superconducting Tokamak Advanced Research (KSTAR) Central Solenoid model coil is described. All the fabrication tools were developed and equipped at Seoul National University to satisfy the main characteristics of the KSTAR coils. A full account of the detailed manufacturing processes such as jacket removal, termination, insulation, winding, welding, heat treatment and resin impregnation are discussed.

Practical considerations on conductor design criteria in superfluid helium are presented, discussed and compared to design criteria in supercritical helium. Results show that the operating temperature has a critical influence on the cost of the system and can be determinant in the choice of the superconducting material. Some economical analysis are given based on a reference toroidal field magnet concept.

We focus on the estimation of the minimum poloidal field coil voltages for the plasma shape control in a tokamak. For given coil locations and for a class of disturbances we propose a procedure for the evaluation of the minimum voltages needed to guarantee that the plasma boundary displacement remains within specified limits. The impact of this problem on the minimization of the power supply cost is discussed for the case of the International Thermonuclear Experimental Reactor (ITER). Index terms - Optimization, plasma control, power supplies, tokamaks, electromagnetic fields, finite element methods.

The maximum safe operating limits of the DIII-D tokamak are determined by the force produced in the ohmic heating coil and the toroidal field coil during a plasma pulse. This force is directly proportional to the product of the current in the coils. Historically, the current limits for each coil were set statically before each pulse without regard for the time varying nature of the currents. In order to allow the full time-dependent capability of the ohmic coil to be used, a system was developed for monitoring the product of the currents dynamically and making appropriate adjustments in real time. This paper discusses the purpose, implementation, and results of this work.

The TCV hybrid analogue-digital plasma control system has been superseded by a high performance Digital Plasma Control System, DPCS, made possible by recent advances in off the shelf technology. We discuss the basic requirements for such a control system and present the design and specifications which were laid down. The nominal and final performances are presented and the complete design is given in detail. The integration of the new system into the current operation of the TCV tokamak is described. The procurement of this system has required close collaboration between the end-users and two commercial suppliers with one of the latter taking full responsibility for the system integration. The impact of this approach on the design and commissioning costs for the TCV project is presented. New possibilities offered by this new system are discussed, including possible work relevant to ITER plasma control development.