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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.

The Divertor Tokamak Test facility (DTT) is a project of an experimental tokamak reactor developed in Italy, in the framework of the European Fusion Roadmap. This work presents the magnetic and the structural assessment of the performance of the DTT central solenoid. The CS is the core magnet of the poloidal system and generates the magnetic flux needed to induce the plasma current. This magnet is composed of a stack of six layer-wound independently energized modules, comprised of Nb3Sn Cable-in-Conduit Conductors. To optimize the amount of superconductive material, each module is divided into two submodules. The inner- most submodule operates in a range of about 8/13.5 T, while the outer one at 6/8.5 T. The objective of the design process is to obtain a coil that is capable of providing the required magnetic performance while being structurally compliant. To address this problem, an analytical assessment has been carried out and a thoroughly parametric Finite Element Model (FEM) has been implemented.

The Divertor Tokamak Test facility (DTT) is an experimental tokamak machine to be built in Frascati, Italy, at the ENEA research center. During its development, the DTT has gone through several design iterations. Developing a rigorous finite-element methodology to evaluate the performance of all its components has thus been a critical part of the verification phase of each new update. This work summarizes the assessment procedure that currently supports the design of the Toroidal Field magnet system, including the superconducting winding pack ( $\mathrm{\mathbf {Nb_3Sn}}$ ), the casing structure and all of the inter-coil structures. Given the high complexity of the 3D structure to reproduce, we implemented some modeling simplifications to solve the problems. Here we describe the finally adopted methodology next to all the motivations we examined to make sure the results were sound and reliable from an engineering point of view.

In a tokamak fusion reactor the confinement of the plasma is guaranteed primarily by a series of Toroidal Field Coils (TFC). The TFCs generate a toroidal magnetic field, and must typically withstand large electromagnetic forces. A constant-tension theoretical approach is commonly applied to define an ideal TFC profile, equilibrating in-plane load distributions without shear and bending effects. This paper focuses on the effectiveness of reduced-order models to analyse the TFC mechanical response when also non-ideal shapes are considered. Different beam models are introduced by referring to different treatments of shear and bending effects. A bottom-up homogenisation technique is employed in order to account for the subscale arrangement of TFC components, allowing also to deduce quantitative indications on average localisation effects. The considered beam theories have been specialised to quantify the discrepancies in the macro-micro TFC mechanical response when arc-based approximations are considered instead of the ideal TFC profile. The successful comparison with numerical benchmark solutions, defined by detailed three-dimensional Finite Element models, confirm that beam-like approaches may be considered reliable in a preliminary design stage.

The “Divertor Tokamak Test facility” (DTT) is an experimental fusion reactor being built in Frascati (IT) in the framework of the European Fusion Roadmap. The DTT Central Solenoid operates in a pulsed regime to generate the magnetic flux needed to induce the plasma current and is therefore subjected to burdensome load conditions. In DTT, this magnet will consist of a stack of six layer-wound independently energized modules, wound with Nb 3 Sn Cable-in-Conduit Conductors. Different layout solutions have been studied to improve the performance of the CS in terms of magnetic flux density and structural integrity. The design presented is the result of the “Design Explorer” algorithm that has been specifically developed for this purpose. This code investigates all possible configurations to find the optimum conductor geometry and winding solution by considering measurable physical quantities to maximize the magnetic flux density while maintaining mechanical stresses at an acceptable level. This manuscript includes the electromagnetic calculation related to the heterogeneous plasma scenarios of the machine, the static structural and the fatigue life assessment and the full set of in-depth analyses performed for the qualification of the main components of the CS structure.

In the context of the European fusion roadmap, the divertor tokamak test (DTT) experimental reactor is intended to investigate alternative divertor configurations in view of the EU-DEMO power exhaust handling necessities. The six poloidal field coils of this tokamak are responsible for the plasma shape and equilibrium, and numerous steps were taken to obtain a design that is magnetically consistent with plasma requirements and structurally compliant with the chosen failure criterion. This work presents the structural assessment that has been performed on the poloidal coil system, taking into account the cooldown process, the energisation to operating conditions and fatigue. Finite Element Analysis has been employed as the principal means of investigation.