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Using infrastructure from the old TPM-1 tokamak in Mexico, there is an ongoing project to bring it back into operation, but with important upgrades. One of the main planned improvements will be the...

The HL-2M tokamak, which is being built at the Southwestern Institute of Physics, is considered one of the most important short-pulse devices for future fusion research in China. The design criterion of the HL-2M specifies the demountable joints for the toroidal field coils (TFCs) due to the assembly requirements. A horizontal preload is needed to keep a good electrical conductance for the joint of each coil. In order to assure the appropriate preload to be exerted on the TFCs under different operation conditions and to provide the preload at 40 locations simultaneously, the horizontal preload system of the HL-2M tokomak has been developed. In this technical note, the design of this system is presented in detail. Finite element analysis is applied to evaluate the design. To validate the assembly and the performance, a prototype of the actuator has been manufactured and tested to finalize the design for fabrication.

The advanced physics and advanced technology tokamak power plant ARIES-ACT1 has a major radius of 6.25 m at an aspect ratio of 4.0, toroidal field of 6.0 T, strong shaping with elongation of 2.2, a...

China is charged with around 7.5% of the ITER toroidal field (TF) conductor fabrication. In total about 6000 km of Nb3Sn strands need to be supplied for Chinese Domestic Agency (CNDA), and now the ...

The toroidal field (TF) magnet of the Joint TEXT (J-TEXT) tokamak consists of 16 D-shaped copper coils interlinked in series. In this study, the finite element simulation software ANSYS is used to ...

KSTAR has a nonlinear magnetic material, INCOLOY® alloy 908 (Incoloy), in toroidal field and poloidal field (PF) coil systems. The effect of Incoloy on the magnetic configuration for the plasma ini...

The mechanisms of toroidal field generation in translated field-reversed configuration (FRC) plasmas are investigated analytically and numerically. Although the radial electric field in the moving ...

The ITER Tokamak will be the largest magnetic confinement fusion device ever built. Confinement will be achieved by a combination of magnetic fields generated by a plasma current of 15MA and extern...

An approach is developed to study the stability of mixed convection in the poloidal flows of the DCLL blanket. Modified Orr-Sommerfeld equations are derived and then solved using a numerical code b...

A new empirical scaling of the magnetic field ratio (the maximum magnetic field on the coil to the toroidal field on the center of helical windings averaged over one field period) for heliotron/tor...

The inductive energy of about 40GJ stored permanently inside the toroidal field (TF) coils of ITER provides a considerable potential of hazard in case of an accident. While for most accidents it could be proved that the damage is limited to the coils themselves, possible high current arcs at the busbars of the TF coils may propagate to and penetrate the cryostat wall. Model arc experiments were set up to understand the propagation and damage potential of such arcs to provide a database for the development and validation of a numerical model as the next step. This work reviews the basic arc propagation and burning modes found so far and introduces new experimental setups and findings.

The National Compact Stellarator Experiment (NCSX) was designed to test a compact, quasiaxisymmetric stellarator configuration. Flexibility and accurate realization of its complex 3D geometry were key requirements affecting the design and construction. While the project was terminated before completing construction, there were significant engineering accomplishments in design, fabrication, and assembly. The design of the stellarator core device was completed. All of the modular coils, toroidal field coils, and vacuum vessel sectors were fabricated. Critical assembly steps were demonstrated. Engineering advances were made in the application of CAD modeling, structural analysis, and accurate fabrication of complex-shaped components and subassemblies. The engineering accomplishments of the project are summarized

In the framework of optimization of confinement properties of Uragan-2M, neoclassical transport in the 1/v regime and the pertinent stored energy are studied. A rather wide range of device configurations is considered. For the computations the field line following code NEO for computation of transport measures is used. Computations are performed in real-space coordinates using the magnetic field produced by the device coils. The difference of currents in adjacent groups of toroidal field (TF) coils as well as the current in the TF coils and the additional vertical magnetic field are used as varying parameters. Some possibilities for improving the 1/v transport corresponding to optimum values of the varying parameters are presented.

Europe has started since the '80s with the preparatory work done on NET the Next European Torus,, the successor of JET, to prepare for the construction of the next generation experiment on the road to the fusion reactor. In 2000 the European Fusion Development Agreement (EFDA) has been signed by sixteen countries, including Switzerland, not a member of the Union. Now the signatory countries have increased to twenty-five. A vigorous programme of design and R&D in support of ITER construction has been conducted by EFDA through the coordinated effort of the national institutes and laboratories supported financially, in the framework of the VI European Framework Research Programme (2002-2006), by contracts of association with EURATOM. In the last three years, with the expenditure of 160 M?, the accent has been particularly put on the preparation of the industrial manufacturing activities of components and systems for ITER. Prototypes and manufacturing methods have been developed in all the main critical areas of machine construction with the objective of providing sound and effective solutions: vacuum vessel, toroidal field coils, poloidal field coils, remote handling equipment, plasma facing components and divertor components, electrical power supplies, generators and power supplies for the Heating and Current Drive Systems and other minor subsystems. Europe feels to be ready to host the ITER site and to provide adequate support and guidance for the success of construction to our partners in the ITER collaboration, wherever needed.

In order to generate a toroidal field in field-reversed configuration (FRC) plasma, a poloidal current drive by axial neutral beam injection (NBI) is investigated analytically and numerically. Spec...

As an R&D Task of shielding neutronics experiment under the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER), streaming experiments with simulating a gap configuration formed by two neighboring blanket modules of ITER were carried out at the FNS facility. In this work, prediction capability of various nuclear design parameters was investigated through analysis of the experiments. The Monte Carlo transport calculation code MCNP-4A and the FENDL/E-1.0 and JENDL Fusion File cross section data libraries were used for the analysis with detailed modeling of the experimental conditions. As a result, all the measured quantities were reproduced within about ± 30 % by the calculations. It was concluded that these calculation tools were capable of predicting nuclear design parameters, such as helium production rates at connection legs of blanket modules to the back plate and nuclear responses in toroidal field coils, with uncertainty of ± 30% for the geometry where gap str...

The design and operation of a spheromak-like compact toroid (SCT) plasma accelerator is described. As an example application, some principles are presented for using the device as a plasma injector...

Experiments in Tokamak-15 were performed to study the STFC heating under the quench. The quench was specially caused by current introduction into STFC at the unchanged input helium temperature. The experimental results and simulation data on temperature heating and amount of heat realized in the pancakes under the quench are given. In the experiments was shown that quench occurs in the internal turns of pancakes and estimations of maximal temperature heating corresponds to calculated ones.

The basic problems of Tokamak-15 electromagnetic system (EMS) elaboration are discussed. The EMS design is described, its main component parameters and features are presented. The R and D program for the EMS design including its most complicated component, the superconducting toroidal field system (TF system), is considered. Post-assembly EMS tests comprising the achievement of a toroidal field nominal value and plasma discharges are described.

A boundary α-effect based on the reflection of hydromagnetic waves from the core-mantle boundary is studied in a simple plane model. The main features of this α-effect, namely: the concentration near the boundary, the change of sign of α in the layer, enhanced values of α and geometrical behavior, are in good agreement with those assumed by Braginsky (1978) in his model-Z numerical analysis. According to the results obtained in this paper the α-layer widens if the toroidal field decreases. Taking this into account a new mechanism of negative feedback is proposed which limits the growth of the toroidal field.

The CIT design has evolved recently from a system in which the central solenoid outer boundary interacts with the inner legs of the toroidal field coils for mutual radial support during operation, ...

Phaedrus-T is a moderate sized tokamak with low toroidal field and substantial radio frequency power capabilities. This machine has only recently become operational and full RF power is expected in...

Magnetic divertors have been proven successful in minimizing plasma/wall interactions and in leading to high-confinement regimes in tokamaks. This suggests that similar benefits may occur i...

The analysis of an l = 2 torsatron with a relatively small pitch angle of the helical winding, which requires an additional toroidal field (TF), is presented. The main advantage of this torsatron d...

The Texas Experimental Tokamak is a medium-scale tokamak operated as a national user facility. Now in operation, it provides a plasma with a 1-m major radius, 28-cm minor radius, and 400-kA nominal plasma current at up to 3-T toroidal field for pulse lengths of 300 to 500 ms. The facility includes all standard tokamak diagnostics and an integrated data system that makes all data available after each shot, as often as once every 2 min. The design is generally conventional and conservative, the vacuum vessel provides numerous large-aperture radial and vertical ports for complete views of the plasma. 5 refs.

The Advanced Toroidal Facility (ATF-1) is a torsatron being designed at ORNL as a replacement for the ISX-B tokamak. The concept for the device includes two continuous helical field coils and a set of twelve toroidal field coils. The vacuum vessel is placed inside the helical field coils and an external toroidal shell support structure is provided to react all coil and vacuum vessel loads. The present design status for all systems is discussed.

Phase-space grouping techniques have been applied to two distinct problems in fusion product physics: (a) slowing down drift motion of highly energetic alpha particles in a symmetric toroidal field...

The present status of modeling toroidal field (TF) ripple loss of fast alphas from tokamaks is summarized. The modeling issues are discussed, and several new aspect of this problem are described, including gyromotion, radial electric field, and sawtoothing. Existing models predict that TF ripple loss of fast alphas will have a low to moderate impact on the design of a tokamak engineering test reactor.

Several preliminary designs have been generated to compare poloidal field (PF) coil set characteristics for comparable limiter and divertor configurations for the Ignition Studies Project. Different coil set locations relative to the toroidal field (TF) coils have been considered. Specifically, configurations have been evaluated with the PF coils entirely external to the TF coils, entirely internal, and combined (hybrid) with all PF coils internal except one.

The TFCX conceptual designs call for the construction of the reactor torus through the use of ''pie-shaped'' segments for mechanical and maintenance considerations. The use of this concept results in hot spots in the inboard section of the torodial field (TF) coils due to neutron and photon streaming through the slots between the segments. This work studies these effects on the nuclear responses in the TF coils and introduces design solutions to reduce the impact on the reactor design.

A study was conducted to establish the tasks and special equipment required to remove and replace a Toroidal Field (T.F.) coil from the FED baseline configuration and to identify critical design fe...

A toroidal field coil for a small aspect ratio, high field tokamak Ignition Test Reactor was designed based on compound constant tension coil shapes. Finite element analyses revealed the existence of bending stresses in this coil due to incompatible deformations of the different constant tension shapes at their common joining points. To correct this situation, the deformation equations of constant tension coils were derived and solved analytically. This permitted selection of cross-sectional dimensions for each coil segment that gave compatible deformations and no bending moments at the joining points. Zero moment theory for compound coils was then extended considering nonuniform continuous elastic coil support effects in the governing equations. Analytical and numerical results were obtained for this case as well.

In this paper we describe a design of an ignition test reactor based on the LITE concept. The main objective of the device is to study key physics issues of ignited plasma operation. LITE-type devices are characterized by toroidal field coils of plate magnet construction. The plates form a continuous structure in the throat of the magnet. The inplane loads are supported by the conductor, while the out-of-plane loads are supported by external frames. The magnet is designed for operation at relatively high stresses and magnetic fields. There is little or no shielding between the vacuum vessel and the magnet. This type of design results in a compact machine.

The PRETEXT tokamak was designed to investigate the initial stages of a tokamak discharge. The machine has a major radius of 53 cm, limiter radius of 15 cm, and a toroidal field (TF) of 8 kG. Plasm...

A multi-year development and manufacturing effort was recently completed with the shipment of the last toroidal field (TF) coil for the Tokamak Fusion Test Reactor (TFTR) project. This manufacturing effort required the development and implementation of complex manufacturing processes and controls to assure TF coil integrity. This paper describes the manufacturing processes, tests and controls which were necessary to produce the 4.26 m OD, 22680 Kg copper TF coils for the TFTR.