The DTT proposal not only considers the fundamental technical and scientific aspects, but also includes a careful financial and managerial analysis. It is organized taking into account that only a ...careful management of all human and financial resources can guarantee the success of the initiative. Regarding the financial aspects, the project analyzes the costs of both engineering design and construction phases, reporting a breakdown of the expenses and their time evolution. The study also considers the expected financial revenues in terms of consistency, funding channels, and time evolution. The organization and government layout (organs, functions, relationships) has been designed pursuing the best balance between autonomy and accountability, transparency and cooperation. The paper also discusses the capability of the ENEA Frascati site to host the device and, in addition, examines the socio-economic impact expected by the long, expansive presence of a high number of scientists and technicians in the area.
The main goal of the Divertor Tokamak Test facility (DTT) is to explore alternative power exhaust solutions for the next step after ITER, i.e., a demonstration power plant DEMO that will explore ...steady-state operation. The principal objective of DTT is to mitigate the risk of a difficult extrapolation to fusion reactor of the conventional divertor based on detached conditions under test on ITER. The task includes several issues, but with the main target to study the completely integrated (physics-technology and bulk-edge) power exhaust problems and to demonstrate how the possible implemented solutions (e.g., advanced divertor configurations or liquid metals) can be integrated in a DEMO device. This paper shows how the parameters for the design of a “flexible” facility, capable to perform this difficult task, can be worked out within the constraint of a fixed budget.
In the European Fusion Roadmap, one of the main challenges to be faced is the risk mitigation related to the impossibility of directly extrapolate to DEMO the divertor solution adopted in ITER, due ...to the very large loads expected. Thus, a satellite experimental facility oriented toward the exploration of robust divertor solutions for power and particles exhaust and to the study of plasma-material interaction scaled to long pulse operation, is currently being designed. Clearly, design requirements for this experiment are quite challenging, to account for the extreme operation conditions, which shall be as representative as possible of the DEMO ones, but in a much smaller device and at lower costs. A feasibility assessment has been carried out for the fully superconducting magnet system of the compact Divertor Tokamak Test (DTT) facility project. The overall magnet system is based on NbTi and Nb3Sn Cable-in-Conduit Conductors, and it adopts some of the most recent developments in this field. It consists of 18 Toroidal Field (TF), 6 Poloidal Field (PF) and 6 Central Solenoid (CS) module coils. In order to cope with the machine requirements such as plasma major and minor radii, magnetic field on plasma axis, plasma current, and inductive flux requirement, the Nb3Sn TF coil is characterized by a peak field of 11.4 T on the conductor, operating at 46.3 kA; the Nb3Sn CS modules are characterized by a peak field of about 13 T, with a conductor operating current of 23 kA; the PF coils are wound using NbTi conductors operating at a maximum peak field of 4.0 T, with operating currents in the range 21 kA to 25 kA, depending on the PF coil. Profiting of the compact machine size, and thus of relatively short conductor lengths, the TF coil winding pack is conceived as layer wound and made of two distinct sections, a low- and a high-field one, employing different superconductor cross-sections, and electrically connected through an embedded “ENEA-type” joint. The main features of the magnet system are described here; the results of mechanical, electrical and thermo-hydraulic analyses, which are discussed here, indicate that the proposed design fulfills all the required criteria. In addition, a brief description of the In-Vessel coils is given, though they are not superconducting, for the sake of completeness.
This paper firstly illustrates the objectives, the figure of merits and the specifications considered in the design of the equilibrium configurations in the Divertor Tokamak Test (DTT) facility. The ...reference single null scenario is detailed. The range of alternative plasma shapes and current capabilities are then discussed. A number of in-vessel coils can be used to locally modify the magnetic configuration in the divertor region. Finally, a comparison of costs and benefits of the various configurations is given, with particular reference to the power exhaust issues.
An upgrade to the lower divertor is currently being planned for EAST superconducting tokamak, aiming at reaching over 400 s long-pulse H-mode operations with a full metal wall and a divertor heat ...load of ˜10 MW/m2. A new divertor concept for EAST, “Tightly Baffled Divertor”, suited to water- cooled W/Cu plasma face components (PFC) with minimized divertor volume, has been proposed to achieve Te,target <5 eV across entire outer target at lower separatrix plasma density and optimized pumping by a simple closed divertor structure combining horizontal target with inclined baffle, dome and duct. This divertor should allow access to high-triangularity small Edge Localized Mode (ELM) H-mode regimes and also allow achieving advanced magnetic divertor configurations with the assistance of two water-cooled in-vessel divertor coils (Divertor coils). Preliminary engineering design of in-vessel Divertor coils indicates a maximum current of 8 kA for long-pulse discharges, and 20 kA for the shortest ones. However, flexibility on Divertor coils position optimization is limited to the water cooling system. Initial plasma equilibrium studies by EFIT code, used in combination with CREATE-NL and FIXFREE tools, show that the distance of the two nearby divertor poloidal field nulls, can be decreased up to ˜ 0.95 m with a plasma current IP ˜ 400 kA, leading to a configuration with the secondary X-point located close to the target, with a significant increase of magnetic poloidal flux expansion and connection length. This may provide a promising divertor solution compatible with advanced steady-state core scenarios.
•Parallel plasma equilibrium reconstruction using GPU for real-time control on EAST.•Vertical control using Bang-bang+PID method to improve the response and minimize the oscillation caused by the ...latency.•Quasi-snow flake divertor plasma configuration has been demonstrated on EAST.
In order to improve the plasma control performance and enhance the capability for advanced plasma control, new algorithms such as PEFIT/ISOFLUX plasma shape feedback control, quasi-snowflake plasma shape development and vertical control under new vertical control power supply, have been implemented and experimentally tested and verified in EAST 2014 campaign. P-EFIT is a rewritten version of EFIT aiming at fast real-time equilibrium reconstruction by using GPU for parallelized computation. Successful control using PEFIT/ISOFLUX was established in dedicated experiment. Snowfldivertor plasma shape has the advantage of spreading heat over the divertor target and a quasi-snowflake (QSF) configuration was achieved in discharges with Ip=0.25 MA and Bt=1.8T, κ∼1.9, by plasma position feedback control. The shape feedback control to achieve QSF shape has been preliminary implemented by using PEFIT and the initial experimental test has been done. For more robust vertical instability control, the inner coil (IC) and its power supply have been upgraded. A new control algorithm with the combination of Bang-bang and PID controllers has been developed. It is shown that new vertical control power supply together with the new control algorithms results in higher vertical controllability.
For long pulse or steady-state advanced plasma discharges, it is necessary to keep plasma parameters such as plasma shape, loop voltage, plasma pressure in a steady state way. In particular, the heat ...load on the divertor target must be effectively reduced. By using Low Hybrid Wave (LHW) for the current drive, loop voltage has been feedback controlled while plasma current was controlled by the PF coil current. The control of the plasma pressure has been demonstrated by using LHW. Heat load reduction has been done by radiation and advanced shape configuration. By using impurity seeding with gas puff for the feedforward and Super-Sonic Molecular Beam Injection (SMBI) to feedback the total radiation, radiation can be effectively controlled with slight influence to the core confinement. The quasi-snowflake (QSF) discharge in H-mode has been achieved. It is verified again in H-mode operation, heat load can be effectively reduced under the QSF shape. All these new control algorithms give rise to more assistances to the EAST long pulse operation.
The paper presents a concept design of a remote handling (RH) system oriented to maintenance operations on the divertor second cassette in FAST, a satellite of ITER tokamak. Starting from ITER ...configuration, a suitably scaled system, composed by a cassette multifunctional mover (CMM) connected to a second cassette end-effector (SCEE), can represent a very efficient solution for FAST machine. The presence of a further system able to open the divertor port, used for RH aims, and remove the first cassette, already aligned with the radial direction of the port, is presumed. Although an ITER-like system maintains essentially shape and proportions of its reference configuration, an appropriate arrangement with FAST environment is needed, taking into account new requirements due to different dimensions, weights and geometries. The use of virtual prototyping and the possibility to involve a great number of persons, not only mechanical designers but also physicist, plasma experts and personnel assigned to remote handling operations, made them to share the multiphysics design experience, according to a concurrent engineering approach. Nevertheless, according to the main objective of any satellite tokamak, such an approach benefits the study of enhancements to ITER RH system and the exploration of alternative solutions.
•A MIMO decoupling controller has been designed for EAST shape control.•Tokamak simulation code (TSC) has been used to verify this MIMO controller.•This method is used in EAST experiment.•Preliminary ...results show the potential of this approach for the EAST plasma shape control systems.
The efficient and safe operation of large fusion devices relies on plasma configuration inside the vacuum chamber. Due to EAST PF coils distribution, it is difficult to physical decouple plasma control parameters and poloidal field (PF) coils current. In this paper, we present a MIMO decoupling controller aimed to overcome the intrinsic limitations of simpler SISO controller, this new controller has been validated by means of the Tokamak Simulation Code (TSC) and implemented on the EAST tokamak. Preliminary results show the potential of this approach for the EAST plasma shape control systems.
PDF