Central Solenoid Systems Research Articles

Updates on Magnet Design For EU-DEMO Reactor: Optimization Studies on TF and CS Systems

In the framework of EU design activities for dimensioning the future fusion DEMOnstration reactor (DEMO), extensive analyses were conducted in EUROfusion context, aiming at ultimately defining the design of the DEMO magnets system. In this objective CEA (Commissariat à l'Énergie Atomique et aux Énergies Alternatives) proposes design for both Central Solenoid (CS) and Toroidal Field (TF) cryomagnetic systems. Considering the latest DEMO reactor baseline both systems were investigated using pre-dimensioning CEA tool MADMACS, with an outcome of several options for each of them. In a second step those designs were investigated with detailed analyses (electromagnetic, thermohydraulic, mechanic). In the paper a broad analysis of the CS design considering fatigue constraints is shown, together with recommendations considering flux target. The TF system considered is an ITER-like option with radial plates and the detailed electromagnetic analysis is exposed and discussed in comparison with former methods. The outcome is used for a thermohydraulic detailed static analysis and a mechanical analysis of TF winding pack is provided and discussed in terms of results but of method as well. Global considerations and recommendations will also be provided to give a first view on cryomagnetic system optimization with respect to cost merits related to both magnet systems.

CEA Broad Studies on EU DEMO CS and PF Magnet Systems

In the framework of EU design activities for dimensioning the future fusion DEMOnstration reactor (DEMO), extensive analyses were conducted in EUROfusion context, aiming at ultimately defining the design of the DEMO magnet system. In this objective CEA proposes designs for all cryomagnetic systems: Toroidal Field (TF) coils, Central Solenoid (CS) and Poloidal Field (PF) coils. In the previous DEMO reactor baselines, designs for CS and PF systems were investigated with pre-dimensioning macroscopic CEA tools, resulting in possible design options. Broad parametric investigations, conducted on the CS and PF systems will be presented. With considerations of constraints imposed by the new DEMO 2018 baseline important inputs (scenario current settings, flux at plasma center...) the PF and CS design features will be quantified, with considerations on design strategy. On PF system a parametric exploration will be led impacting on PFs aspect ratio and a tentative design proposal will be issued. On CS system conductor current range choice is investigated, (DEMO 2018 CS design proposal will be issued under those guidelines) and parametric studies on breakdown regime will be conducted on two CS designs. The outcomes of both exhaustive explorations will be exposed and discussion will be led with tentative recommendations.

Preliminarily Static Analysis of CFETR Central Solenoid Magnet System

Conceptual design of China Fusion Engineering Test Reactor (CFETR) Central Solenoid (CS) coil had been started in Institute of Plasma Physics, Chinese Academy of Sciences. The highest field of CS coil is 17.2 T when the running current is 60 kA. CS magnet system mainly consists of 8 Nb3Sn coils compressed with 8 sets of preload structure. The functions of the preload structure are to apply an enough axial compression to the CS coils and to have a mechanical rigidity against the repulsive force between 8 Nb3Sn coils. This paper describes structural design of CFETR CS magnet system. A global finite element model is created based on the design geometry data to investigate the mechanical property of CFETR CS preload structure and support structure under the different operating conditions. 2D finite element model under electromagnetic is created to calculate the stress on the conductor jacket and turn insulation.

Analysis of ITER Magnet in Safety-Related Fault Condition—Case Study for PF3

The ITER magnet system contains stored energy, 41 GJ in the toroidal field system and up to ~10 GJ in the poloidal field (PF)/central solenoid system during plasma operation. A quench in the ITER magnet is regarded as a normal event, and the ITER magnet system has a redundant quench detection system and a reliable fast discharge system in order to achieve low probability of unmitigated quench and its propagation in the ITER magnet system. An electrical circuit model is developed by using a circuit simulator in order to estimate the arc power during fault conditions. The ANSYS 3-D model of the magnet, including electrical, thermal, and arcing inside the coil, is being constructed to analyze the thermal and electrical behavior of the magnet in an unmitigated quench. In this paper, thermal analyses of the fault condition related to the PF coil and impacts to the vacuum vessel are reported.

Conceptual integrated approach for the magnet system of a tokamak reactor

In the framework of the reflexion about DEMO, a conceptual integrated approach for the magnet system of a tokamak reactor is presented. This objective is reached using analytical formulas which are presented in this paper, coupled to a Fortran code ESCORT (Electromagnetic Superconducting System for the Computation of Research Tokamaks), to be integrated into SYCOMORE, a code for reactor modelling presently in development at CEA/IRFM in Cadarache, using the tools of the EFDA Integrated Tokamak Modelling task force. The analytical formulas deal with all aspects of the magnet system, starting from the derivation of the TF system general geometry, from the plasma main characteristics. The design criteria for the cable current density and the structural design of the toroidal field and central solenoid systems are presented, enabling to deliver the radial thicknesses of the magnets and enabling also to estimate the plasma duration of the plateau. As a matter of fact, a pulsed version DEMO is presently actively considered in the European programmes. Considerations regarding the cryogenics and the protection are given, affecting the general design. An application of the conceptual approach is presented, allowing a comparison between ESCORT output data and actual ITER parameters and giving the main characteristics of a possible version for DEMO.

Conceptual design of spherical torus with TF–CS hybrid coils based on virial theorem

A conceptual design of a spherical torus (ST) device with a new type of toroidal field (TF) coils and a central solenoid (CS) whose stress is reduced to a theoretical limit determined by the virial theorem is proposed. A virial-limit coil (VLC), which is a helical-type hybrid coil of TF coils and CS, and its winding is modulated so that its winding direction is nearly vertical and more horizontal in the outer and inner sides of torus, respectively. Thus, the configuration of VLC is similar to that of CS and TF coil systems in conventional tokamaks. Recently, we have developed a VLC concept with a non-circular cross-section, and reduce the maximum stress to 30% of that of TF coils in a two-dimensional analysis. Moreover, the VLC configuration is remarkably similar to that of a ST with a low aspect ratio and a strongly elongated cross-section with a triangularity. In this work, we analyze three dimensional stress distributions, and propose a design of VLC ST.

A tokamak with nearly uniform coil stress based on the virial theorem

A novel tokamak design with a new type of toroidal field (TF) coil and a central solenoid (CS) whose stress is reduced greatly to the theoretical limit determined by the virial theorem, has been devised, and plasma production and confinement experiments in a new small tokamak based on this design are presented. According to the virial theorem, the best TF coil for producing the strongest magnetic field under the weakest stress requires equal averaged principal stresses in all directions. Applying this condition to a helical coil, its pitch number is determined as a function of the aspect ratio. The helical winding according to this condition is modulated in such a way that the poloidal field exists only outside of the torus, which reduces the torsional force on the helical coil and makes plasma breakdown possible. Moreover, a helical coil with this modulation and a low aspect ratio is similar to CS and TF coil systems in conventional tokamaks since its helical winding is nearly vertical in the outer part of the torus. With the aspect ratio A = 2, our optimal coil theoretically reduces the working stress in the coil to about one-third less than that of conventional TF coils. On the basis of the design, a small prototype, ‘Todoroki-II’, with a major radius of 0.3 m, a minor radius of 0.08 m, toroidal magnetic field strengths of BT < 1.5 T and plasma currents of IP < 40 kA, was made. An external vertical field increased the plasma pulse length and the current to 1 ms and 11 kA, respectively, while they were restricted with no vertical field control. Using a Cauchy-condition surface method, the shape and displacement of the plasma boundary were reconstructed, and position control using the vetical field was confirmed.

The ATLAS magnet test facility at CERN

The ATLAS magnet system consists of a barrel toroid (BT), two end-cap toroids (ECT) and a central solenoid (CS). The BT with overall dimensions of 20 m diameter by 26 m length is made of 8 coils placed in their individual cryostats and symmetrically assembled around the central axis. The ATLAS magnet test facility is being completed at CERN and will host in the next years to come pairs of BT coils at the same time as well as the ECT and the CS systems already fully assembled. Two model coils will be tested in the years 2000-2001, the B00 model coil and the B0 model coil, a 9 m short version of the BT coil built to verify the functionality and the construction concepts of the BT.