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The 'Divertor Tokamak Test' (DTT) is an experimental fusion reactor being built in Frascati (IT) in the framework of the European Fusion Roadmap. The DTT Central Solenoid, used to drive the current in the magnetically coupled plasma, comprises six Nb3Sn layer-wound independently energized modules. Each module is made of three sub-modules: High Field (HF), Medium Field (MF) and Low Field (LF) grades. To meet all goals of the DTT scientific program, a variety of plasma scenarios have been designed. These cause intense and heterogeneous loading conditions for the CS stack; from the mechanical point of view, each module is subjected to a vertical expansion or compression and to a huge radial action, whereas the current variations cause relevant heat loads due to AC losses, with impact on the coil temperature margin. In this work, the engineering study on the current design and the analyses on the module windings are presented.

The solution of the problem of heat exhaust has been pointed out as one of the main challenge towards the realization of magnetic confinement fusion. In the last years, two concepts have been proposed in alternative to the conventional divertor solution adopted for ITER: modification of the magnetic topology in the divertor region and liquid metal as plasma facing component. The role of the Divertor Tokamak Test facility (DTT) in the power exhaust implementation strategy is discussed. The evolution of the project, since the original proposal in 2015 to the present design, is shown. The DTT facility is well integrated in the European strategy and the final decision on the divertor configuration will be made, within 2022-23, on the basis of the indication of the Power Exhaust Group constituted by the EUROfusion Consortium. Finally, the main milestones and the timeline of the project are illustrated. © 2019 Elsevier B.V.

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.

The paper is focused on the electromagnetic analysis of the Toroidal Field (TF) superconducting coils of the Divertor Tokamak Test facility (DTT) when electrical transients occur in the TF coils system: for example, during the operations of the Fast Discharge Units (FDUs) and considering also, the simultaneous occurrence of a fault condition. During the FDU intervention, a transient voltage excitation lasting few microseconds occurs at the TF coil terminals and it electrically stresses the insulations of TF coils itsef. To investigate the voltage distribution across, inside and between different Double Pancakes (DPs) of each TF coil, a lumped parameters circuital model has been developed and implemented in Ansys Simplorer simulation environment. This model includes both the detailed sub-model of each TF coil and FDU. The transient analyses have been carried out for two different scenarios: a reference one and a failure scenario, considering three different fault resistance values and also two different values of the resistance connecting the TF case to ground. In order to verify the correct sizing of the coil insulation and the TF case-to-ground resistance value inserted in the circuit of each TF coil, the voltages of each TF coil (terminal-to-terminal, terminal-to-ground, across of adjacent DPs and so) were computed in the time domain (in the range of milliseconds) for both scenarios. An overview of calculations and simulation results is presented and discussed, pointing out a high sensitivity of fault conditions and of overvoltage values and addressing mitigation strategies.

The EU DEMO fusion reactor, now entering in its conceptual design phase, will have to tackle several challenges, such as the long plasma pulse duration and the exhaust of very high heat fluxes from the plasma. The Divertor Tokamak Test (DTT) facility, a compact superconducting tokamak under construction at ENEA Frascati, will address the power exhaust problem in DEMO by testing several DEMO-relevant divertor solutions and operation scenarios. The DTT superconducting coils are the subject of the analyses presented here: as the tokamak must be very flexible to face different plasma scenarios, the design of the magnet system must be proven to be robust. In particular, the thermal-hydraulic performance of the coils operated in pulsed mode, namely the Central Solenoid (CS) and the Poloidal Field (PF) coils, will be analyzed. The CS is composed of 6 modules, layer-wound with Nb3Sn Cable-in-Conduit Conductors (CICCs), while the PF coils are pancake-wound. The two PF coils closest to the machine axis are wound using Nb3Sn CICCs, while the others adopt NbTi CICCs. All the coils are cooled by supercritical He (SHe) in forced flow. The pulsed operation of these coils induces AC losses, eroding their temperature margin: a detailed thermal-hydraulic model of the DTT pulsed coils is developed here using the 4C code. The model is then applied to the simulation of the two reference plasma scenarios, a single null and a double null, computing the minimum temperature margin. Different cooling options (single- or double-pancake) are investigated for the PF coils, while the sensitivity to the coupling time constant value is assessed for the CS.

In the frame of the conceptual design studies for the Toroidal Field (TF) coils of DEMO, a solution based on a layer-wound magnet, rectangular-shaped Cable-in-Conduit conductors and W&R manufacturing approach, is being developed. The feasibility and performance of a large-size Nb3Sn conductor operating at about 82 kA in a 13 T field has been proven in the past. Another key technology to be demonstrated for a layer-wound TF coil, is that of a joint between two different conductor grades, to be possibly manufactured in-line during winding. The proposed joint solution would provide the minimum room occupancy, with the joint embedded within the winding pack volume. A joint between two low-field conductor grades, i.e., constituted of a small number of superconducting strands and a high number of stabilization copper wires, has been designed and manufactured. The two conductor lengths are characterized by a different number of superconducting wires and different outer dimensions. A joint sample has been assembled and instrumented, in the configuration allowing testing at the SULTAN facility of the Swiss Plasma Center. Both DC and AC performance of the joint has been characterized at the DEMO TF operating conditions. The present paper reports the main manufacturing steps for the joint and its main test results. The implications on the performance and design approach of the TF coil are also discussed, based on the outcome of such tests.

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.