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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.

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.

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.

The Divertor Tokamak Test (DTT) facility is currently under design in Italy. This fully superconductive compact tokamak will be the test bench of several DEMO-relevant divertor solutions. The 4C code model of a DTT toroidal field magnet (including its structures) is developed here and used to support some important design decisions related to the neutron shield to be adopted, the need of an active cooling of the casing and the static heat load reduction. The simulations confirm the need to actively cool the casing by suitable cooling channels, but also the need of a proper neutron shield to reduce the nuclear heat load on the superconducting coils. On the other hand, the proposed static heat load reduction measures do not appear to be effective enough to satisfy the design requirement of 1.4 K for the minimum temperature margin.

The Divertor Tokamak Test facility is an Italian experimental facility under design and construction at ENEA C.R. Frascati. The main goal of DTT is to provide an integrated environment, relevant to DEMO, where testing possible solution to the power exhaust problem in a tokamak (like for example: i) Plasma facing components technology ; ii) Plasma and divertor shape; iii) impurity seeding to increase radiation). In this respect, DTT has been designed to be flexible and adopting technologies relevant to DEMO. After its initial inception in 2015, concluded with the publication of the "DTT project proposal", a complete re-baseline has been provided oncluded with the publication of the "DTT Interim Design Report" in 2019, aimed at accommodating the request of flexibility coming from the international fusion community. During 2019, the engineering integration activity has started and the first construction contracts have been signed. This paper provides an overview of the integrated design activity towards the realization of the facility within 2025.

The first design of the NbTi poloidal field (PF) coils for the EU DEMO fusion reactor has been proposed by the Swiss Plasma Center and by the Italian National Agency for New Technologies, Energy and Sustainable Economic Development. The thermal-hydraulic (TH) performance analysis of the PF coil system presented in this work has been carried out using the state-of-The-Art 4C code. The minimum temperature margin and the TH response of the coils to the heat deposition due to AC losses are computed in two different plasma scenarios, using a single time constant ( nτ) whose value is currently unknown. Therefore, we apply our model to parametrically assess the sensitivity of the PF performance to a range of nτ values. The calculations are also performed taking into account that the high void fraction design of the conductor leads to the opening of a channel due to the Lorentz force. For all situations considered here, the 4C code predicts that the temperature margin never goes below the acceptable minimum of 1.5 K. © 2002-2011 IEEE.

In this paper, we report on the experimental study of the bending behavior, at 77 K, of a cable-in-conduit conductor composed of a stack of REBa 2 Cu 3 O 7-x (REBCO) tapes inserted into a helical duct formed in the extruded aluminum cylindrical core. The investigation was carried out by powering each single tape individually. By the analysis of the single tape I c dependence as a function of the bending radius, R b , the effect of bending strain as a function of tape position inside the stack was investigated for each tape. The results evidence the good bending strain tolerance of all the tapes, showing the onset of degradation at R b ≈ 0.25 m, ascribed to the slippage occurring among tapes within the stack. Interestingly, the decrease in the intertape contact resistance between neighboring tapes with respect to the straight cable condition reveals that the bending stress results in a beneficial additional transverse load on the stacks, which leads to a more uniform compaction of tape stack. The experimental I c behavior with R b has been explained considering that the current transfer mechanism among tapes might mitigate the degradation of the tape I c due to the bending strain. The identification of the mechanisms acting on individual tapes under cable bending conditions is a relevant achievement in perspective of the interpretation of the electromechanical behavior of the conductor with all tapes powered in parallel.

Abstracrt ITER is incorporating two types of In-Vessel Coils (IVCs): ELM Coils to mitigate Edge Localized Modes and VS Coils to provide a reliable Vertical Stabilization of the plasma. Strong coupling with the plasma is required in order that the ELM and VS Coils can meet their performance requirements. Accordingly, the IVCs are mounted on the Vacuum Vessel (VV) inner wall, in close proximity to the plasma, just behind the Blanket Shield Modules (BSM). Due to high radiation environment, mineral insulated copper conductors enclosed in a stainless steel jacket have been selected. The reference design and prototype work provided a good basis for the development of radiation resistant conductor capable of operating within the harsh conditions in ITER vacuum chamber. However, this effort identified shortcomings in achieving satisfactory manufacturing solution, and most significantly, difficulties in brazing the brackets onto the ELM coil conductor. Since this process has not proven successful, alternative designs are under development and prototyping. Prototype manufacturing on the alternative designs has been completed at ICAS, Italy and ASIPP, China. The aim was to eliminate the need for internal coil joints, to prove the principle of longer conductor length manufacturing, and to perform bending and welding trials on two different conductor cross-sections: circular and square. The procurement of the IVCs and their conductors will be done via direct call for tender from the ITER Organization. This paper will give an overview of the alternative design and prototype manufacturing of the ITER In-Vessel coils.

Within the frame of the R&D activities carried out in Europe for the toroidal field coils of the nuclear fusion device DEMO, a fundamental milestone was considered to be the demonstration of Nb3Sn cable-in-conduit conductors (CICC) performance in the demanding range of interest for DEMO. Among the different technological solutions envisaged, the present paper deals with a wind & react CICC solution, with rectangular geometry, thick steel jacket, and distributed pressure relief channel, designed to operate at 82 kA in a magnetic field of 13 T and with a current sharing temperature ${\rm{T}}_{{\rm{cs}}}\,{\rm{> \,6.5\,K}}$ . The main manufacturing steps of the prototype conductor are described in the present paper, carried out within industrial environment, partly using the facilities and procedures available for the manufacture of ITER conductors. A sample was designed for the EDIPO facility at the Swiss Plasma Center, Switzerland, in the configuration usually adopted for the test of ITER poloidal field conductors, where the two straight conductor legs are part of the same cable length, with a continuous transition through a bottom hairpin-type joint, thus avoiding any resistive connection. The conductor has been characterized in terms of dc performance at relevant operating conditions and the absence of any performance degradation with electro-magnetic load cycles has been verified, thus, qualifying the proposed technological solution. AC losses and thermo-hydraulic tests have also been carried out, providing relevant information for further coil design.

A 45-T hybrid magnet is being built at the High Field Magnet Laboratory of the Radboud University in Nijmegen, The Netherlands. The hybrid magnet consists of a 12-T cable-in- conduit-conductor (CICC) Nb 3Sn superconducting outsert and a 33-T resistive insert magnet. To verify the CICC design, a thorough testing has been completed in the SULTAN facility at Swiss Plasma Center, EPFL in Villigen (Switzerland) for the medium- grade conductor of the outsert. In two test campaigns, the dc cable performance (current-sharing temperature, critical current), the ac loss, and the conductor's performance stability during cyclic loading and after one warmup and cooldown cycle have been investigated. Two different cable layouts were tested - one with a very short twist pitch (STP) and the second one with a long twist pitch (LTP) cabling pattern. As both conductors were made of the same Nb 3Sn strand and underwent the same heat treatment and sample preparation procedure, the effect of the twist pitch on the ac loss and on the dc performance with respect to cyclic loading could be reliably evaluated. The test results show that both cable layouts are actually very robust. The cable could withstand 2000 load cycles and the warmup and cooldown cycle without any significant degradation of the dc performance, and even the overloading at BI product (field multiplied by current) approximately two times larger than those foreseen during magnet operation did not lead to a big performance change. Small differences between the STP and LTP options have been observed, indicating that the STP conductor withstands high electromagnetic loads better than the LTP one. © 2002-2011 IEEE.