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• The flux, generated by a DEMO CS coil of fixed outer radius, is maximized. • Use of high- T c superconductors would allow a 32% increase of the generated flux. • Layer-wound CS coil design with superconductor and steel grading is proposed. • In the highest field sections, the use of high- T c superconductors is envisaged. The present work is performed within the framework of the EUROfusion DEMO project. Previously, it was demonstrated that for a maintained magnetic flux the use of HTS conductors at highest magnetic field in a layer-wound CS coil would allow the reduction of its outer radius by around 0.5 m as compared to the DEMO reference design using only Nb 3 Sn conductors. Alternatively, the superior high field performance of HTS conductors can be used to maximize the magnetic flux of a CS coil with a given outer diameter, which could significantly prolong the duration of plasma burn phase and thus the overall power plant efficiency. In the present study, the maximization of the magnetic flux, generated by a layer-wound CS coil of fixed outer radius, is considered. HTS conductors are envisaged to be used at highest field, while Nb 3 Sn and NbTi are foreseen to be used in intermediate and low field layers, respectively. The inner radius of the CS coil is optimized with respect to the generation of a maximum flux taking into consideration the superconductor properties, the hoop stress and the axial stress. In order to provide reasonable starting values for the finite element analysis, first a simplified model with layer-dependent current densities, however, without stainless steel grading will be considered. In the final outline design of the CS coil a superconductor and stainless steel grading will be implemented. [ABSTRACT FROM AUTHOR]

• Procedure to find change of TF radial build for varied conductor current. • Procedure to find change of TF radial build for varied discharge time constant. • Impact of changed TF coil designs on stresses in the steel jackets. • Relations between TF coil design and discharge voltage in case of quench. • Relations between TF design, number of feeders and discharge voltage. In addition to a reduced number of feeders, which would simplify their integration in the cryostat of the tokamak, the present study considers the possibility to reduce the voltage during safety discharge of the toroidal field (TF) coils. A lowered discharge voltage would ease the manufacture of the TF coils and reduce the operational risks. Both aspects closely related to the discharge time constant, determined by the ratio L TF / R d , where R d is the resistance of the dump resistor and L TF the inductance of the TF coil. The supply of current to two or more TF coils in series requires a reduced discharge voltage per TF coil. The present study considers a fixed total TF current of 14.9 MA-turn provided by the DEMO baseline 2018. The discharge voltage can be reduced by increasing the TF conductor current, and hence lowering the inductance, or by an enlarged discharge time constant. For an increased discharge time constant, the copper cross-section in the TF winding pack needs to be increased to limit the hot spot temperature to acceptable values. A parametric study of the impact of increased TF conductor currents and discharge time constants on the radial build of the TF winding pack has been performed. The advantages of a possible reduction of the TF discharge voltage need to be compared with the disadvantages of an increased radial build. [ABSTRACT FROM AUTHOR]

[Display omitted] • DEMO feeder system integration issues are evaluated at an early design stage. • Hybrid design based on Nb 3 Sn lead extensions, NbTi bus bars and HTS current leads. • A 2.5 K temperature margin is used in NbTi bus bar sizing and cable design. • HTS current leads outline design is based on a loss-of-flow accident. • Preliminary considerations on bus bar joints are presented. The integration of the feeder system in a tokamak involves complex engineering, manufacturing and logistics issues. Therefore, within the pre-conceptual design activities of the European DEMO fusion reactor, it is important to provide a preliminary sizing of the feeders at an early stage. Here we present the conceptual design of superconducting bus bars and current leads for the DEMO feeder system. The proposal is based on three main components: Nb 3 Sn lead extensions, NbTi bus bars and HTS current leads. We propose three different bus bar designs – for the TF, CS and PF coil feeders, respectively. Two additional values of the TF operating current are considered. The structural verification of the bus bars, including an S-bend geometry for reducing the mechanical loads due to cool down, is performed both analytically and by means of 3D finite-element simulations. In addition, we present an outline design of the HTS current leads, based on the minimisation of the cryogenic heat load and the simulation of a loss-of-flow accident. Finally, we briefly discuss the main characteristics of the joints, as well as some integration and assembly challenges. [ABSTRACT FROM AUTHOR]

The European Roadmap to Fusion Electricity (Federici et al., 2018) [1] details the path to complete within the next three decades the DEMOnstration power plant, DEMO, aiming to a net gain of Energy Q=40. The 2018 DEMO baseline considers a 2 GW tokamak device, a target performance of 6 T on the plasma, with a major radius of 9 m. Considering the plasma aspect ratio and elongation, and the radial build of the thermal shield, the outer radius of the TFC inner leg is about 4.3 m. In the current designs, the TFC winding pack operates in a range of B = 12−13 T and is based on low temperature superconductors (LTS). As an alternative, High Temperature Superconductors (HTS) could be employed to extend the coil operation beyond the typical J/T/B range of Nb3Sn. The rationale behind the exploration of a fully HTS or Hybrid HTS/LTS design of the TFC Winding Pack (WP) is twofold: on the one hand, it is possible to maintain the target performance of the machine redesigning the radial build of each magnet as a function of a more efficient cable in terms of critical current density; on the other hand, it is possible to keep the machine power constant, increasing the field on the plasma axis, and consequently define a new radial build. This work presents and describes a combined algorithm implemented to enhance the DEMO magnet systems and optimize the DTT central solenoid, capable of accounting for the main electromagnetic and structural constraints, and consistently reshape the machine. • This paper describes the MAgnet Design Explorer algorithm (MADE). • The aim of this tool is improving the capability of the existing design codes. • Characterize the magnet radial-builds inside the reactor. • Intertwine physical, mechanical and manufacturing aspects of the fusion magnets. • Compare different superconducting technologies highlighting advantage and disadvantage. [ABSTRACT FROM AUTHOR]

A hybrid design of the Central Solenoid is under study within the conceptual design phase of the European DEMO fusion reactor. It is based on HTS conductors in the innermost high-field layers, and Nb 3 Sn and NbTi conductors for field values below 15 T. Since the resistance of the stainless steel jacket to mechanical fatigue drives the whole magnet design, alternative layouts of the cable-in-conduit conductors are aimed at decoupling the mechanical and hydraulic functions of the jacket itself. Thus, a reduced steel thickness, a smaller radial build of the coil and a larger magnetic flux are achievable. In this work, we analyse a new cable design based on filled conductors and tight He channels, both for the HTS and the LTS layers, and study their thermal-hydraulic performance by means of numerical simulations. After having introduced the geometrical and material details of the considered conductors, we describe the model and assumptions made for estimating the AC losses, assumed as the only heating source. Finally, we discuss the resulting minimum temperature margin of the conductors and provide useful insights for future activities on conductor development. [ABSTRACT FROM AUTHOR]

Abstract: In the neutral beam cell of ITER all the maintenance operations on the neutral beam components (BLC''s) must be performed by an overhead crane of large payload capability (30–50tonnes). A crane system is presented consisting of a monorail, a carriage, and a lifting mechanism. The monorail must give access to the BLC''s in the beam line vessel, the front components connecting the NB vessel with the Tokamak, and a storage area at the north end of the NB cell. Rail switching points are required at the intersections between radial and toroidal branches. A translational switching mechanism is proposed. The crane carriage consists of two independent sub-carriages, each composed of four wheels. A set of four secondary wheels attached to the main carriage prevents the crane tilting due to the CoG misalignment of some loads. The elevation system proposed consists of an electromechanical crane of four independent drums and 50tonnes payload. In parallel with the crane design, a logistics and space availability study has been carried out, leading to the detection of clearance or transport problems that could be taken into account in the final crane design. [Copyright &y& Elsevier]

Abstract: The maintenance operations of ITER NB components inside the vessel - Beam Line Components (BLC''s) involve the removal of the faulty component, its transport to the hot cell as well as the reverse operations of transport of the repaired/new component and its reinstallation inside the vessel. Prior to the removal of the BLC''s the cooling pipes must be detached from the component following a procedure that applies to the cutting of the pipes and subsequent welding when the component is re-installed. The purpose of this study, conducted in the framework of EFDA, is to demonstrate the feasibility of the cut and weld operations on the water pipes of the BLC''s using fully remote handling techniques. Viable technologies for the cut and weld operations have been identified within the study; in particular the following aspects will be presented in the paper: [•] Different strategies can be pursued in the detachment of the components depending on the number of cut and weld operations to be performed on the pipes. The selected strategy will impact on the procedure to be followed likewise on important aspects as the requirements of the flexible joints assembled on the pipes. [•] The existing cutting techniques have been examined in the light of the remotely performed pipe cutting at the NB cell. Modifications of commercial tools have been proposed in order to adapt them to the BLC''s pipes requirements. The debris produced during the cutting process must be controlled and collected, therefore a cleaning system has been integrated in the adapted cutting tool referred above. [•] The existing welding techniques have been also examined and compared based on different criteria such as complexity, reliability, alignment tolerances, etc. TIG welding is the preferred technique as it stands out for its superior performance. The commercial tools identified need to be adapted to the NB environment. [•] The alignment of the pipes is a critical issue concerning the remote welding. A proper alignment system has been proposed taking into account the pre-selected welding technique. [Copyright &y& Elsevier]