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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 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.

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 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.

Sedlak, K. et al., The European DEMO, i.e. the demonstration fusion power plant designed in the framework of the Roadmap to Fusion Electricity by the EUROfusion Consortium, is approaching the end of the pre-conceptual design phase, to be accomplished with a Gate Review in 2020, in which all DEMO subsystems will be reviewed by panels of independent experts. The latest 2018 DEMO baseline has major and minor radius of 9.1 m and 2.9 m, plasma current 17.9 MA, toroidal field on the plasma axis 5.2 T, and the peak field in the toroidal-field (TF) conductor 12.0 T. The 900 ton heavy TF coil is prepared in four low-temperature-superconductor (LTS) variants, some of them differing slightly, other significantly, from the ITER TF coil design. Two variants of the CS coils are investigated—a purely LTS one resembling the ITER CS, and a hybrid coil, in which the innermost layers made of HTS allow the designers either to increase the magnetic flux, and thus the duration of the fusion pulse, or to reduce the outer radius of the CS coil. An issue presently investigated by mechanical analyzes is the fatigue load. Two variants of the poloidal field coils are being investigated. The magnet and conductor design studies are accompanied by the experimental tests on both LTS and HTS prototype samples, covering a broad range of DC and AC tests. Testing of quench behavior of the 15 kA HTS cables, with size and layout relevant for the fusion magnets and cooled by forced flow helium, is in preparation., This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014–2018 and 2019–2020 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission. This work was supported in part by the Swiss National Science Foundation (SNF) under contract number 200021_179134.

The design of an interface device for interconnecting a SMES (Superconducting Magnetic Energy Storage) to a LV distribution grid is carried out, in order to demonstrate the feasibility of a SMES for achieving, with the same device, power quality and critical load protection functions. The work is developed inside a three-year research project, called DRYSMES4GRID, funded by the Italian Minister of Economic Development and sees the cooperation of five partners (companies, universities, and research institutes). The aim of the project is to demonstrate the feasibility of the SMES (500 kJ/200 kW) in the short/medium term at competitive cost based on Magnesium Diboride (MgB2). In particular, in this article, after a brief state of art of the SMES applications for the grid, the design and the control logics of the DC/DC chopper for the power modulation of the energy stored in the SMES and of the threephase AC/DC inverter are shown. The activities are mainly based on digital simulations to evaluate the performances of this device in a LV distribution grid in presence of linear and non-linear loads, considering in particular the peak shaving function, while the device is connected to the mains, and the islanding operation condition after the detection of a voltage dip.

A three-year research project, called DRYSMES4GRID, was recently funded by the Italian Minister of Economic Development, Italy. The project, which involves five Italian partners (companies, universities, and research institutes), is aimed to demonstrate the feasibility of cost-competitive SMES based on magnesium diboride (MgB2) with a cryogen-free cooling by means of the manufacturing and the testing of a demonstrator with an objective rating of 500 kJ/200 kW. This rating is deemed suitable for disclosing the critical technological aspects of all components in view of practical applications. A further goal of the project is the assessment of the technical and economic benefits that the SMES can bring to real-world electric grids. The general outline and technical objectives of the project are presented in this paper. Preliminary design assumptions and results concerning the MgB2 coil and the power condition system are also discussed. © 2018 IEEE.

The design of the superconducting magnet system of the European DEMO fusion reactor is currently being pursued in the framework of the EUROfusion Magnets Work Package. Three alternative winding pack (WP) options for the toroidal field coils (TFCs) are being proposed by different research units, each featuring a different conductor manufacturing technology (reactand-wind versus wind-and-react) or winding layout (layer versus pancake). One of the options (namely, WP#2), proposed by Italian ENEA, features a layer-wound WP design adopting a wind-andreact conductor with rectangular cross section with high aspect ratio, obtained squeezing an initially circular conductor. In order to assess the capability of all the TFC components to withstand the electromagnetic loads due to the huge Lorentz forces without any structural failure during the magnet lifetime, the mechanical analysis of the 2016 version of the WP#2 design option is performed here applying a hierarchical approach herein defined as the stress recovery tool: the finite element analysis (FEA) of a whole magnet (including the casing) is performed at a low computational cost adopting a coarse WP model with smeared (homogenized) properties. The displacements computed on the smeared WP are then used as boundary conditions for a refined FEA of some WP slices, located in selected (critical) poloidal positions, where all the conductors detailed features (jacket, insulations) are properly accounted for.

The ITER toroidal field (TF) conductor is made up of superconducting Nb3Sn and copper strands assembled into a multistage, rope-type cable inserted into a conduit of butt-welded stainless steel jacket sections. For the ITER Project, the Korean Domestic Agency (KODA) took the responsibility of the procurement of 27 superconducting conductors for the ITER TF magnets. After concluding the Procurement Arrangement (PA) with the ITER International Organization in May 2008, KODA has been implementing the PA through four major industrial contracts: (1) Nb3Sn strand, (2) cable, (3) stainless steel jacket sections, and (4) jacketing. Prior to the production of conductors required for the TF coils, one 760-m-long copper dummy conductor and one 100-m-long superconducting conductor were fabricated for manufacturing process qualification. As of June 2013, 16 TF conductors were successfully manufactured. The full-size conductor performance tests in the SULTAN facility yielded very high performance. This article describes the technical requirements of the TF conductor and how KODA has been manufacturing the conductors with a high-level quality assurance/quality control system. It also presents the results of acceptance tests, including those of the SULTAN test.

The superconducting magnet system plays an important role in the framework of the design of the EU DEMO tokamak. In recent years, ENEA developed a prototype of cable-in-conduit conductor (CICC) with two low-impedance central channels to be used in the DEMO Toroidal Field (TF) coils with a graded winding pack (WP). In this paper, a model of a TF coil based on the thermal-hydraulic code 4C has been developed, including the WP, the steel casing with dedicated cooling channels (CCCs) and the two independent cryogenic circuits cooling the WP and the casing, respectively. The first part of the work analyzes the performance of the WP during a series of standard plasma pulses in normal operating conditions. In the second part different off-normal operating conditions during the plasma pulses are studied, namely the collapse of one or both central channel(s) in the most critical CICCs and the plugging of some CCCs at the most critical locations in the magnet.

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.

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 Nb 3 Sn 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 Nb 3 Sn TF coil is characterized by a peak field of 11.4 T on the conductor, operating at 46.3 kA; the Nb 3 Sn 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.

The design of a suitable quench protection system is fundamental for the safe operation of superconducting magnets and in turn requires the accurate simulation of the quench transient. The quench propagation in a toroidal field (TF) coil for the future European fusion reactor (EU DEMO) is analyzed here considering the latest, layer-wound winding pack (WP) design proposed by ENEA. The thermal-hydraulic model of a TF coil implemented in the 4C code is updated by including the external cryogenic circuits of the WP and of the casing cooling channels and proposing a preliminary layout of the quench lines. Three different locations are considered for the quench initiation: maximum temperature margin in the WP, and minimum and maximum temperature margin on the same turn of the innermost layer. The evolution of the main electrical and thermal-hydraulic parameters is simulated, such as voltage along each layer, quench front propagation both along and across the layers, hot spot temperature, pressurization of the coil and coolant mass flow rate at the coil boundaries, so that the 4C code provides a reliable (in view of its validation) and detailed virtual monitor of what happens inside the coil during the quench transient. In all cases considered, the ENEA design is predicted to satisfy the present (i.e., ITER) design criteria concerning the maximum allowed hot spot temperature. © American Nuclear Society.

In the framework of the Broader Approach Activities, the EU will deliver to Japan the 18 superconducting coils, which constitute the JT-60SA Toroidal field magnet. These 18 coils, manufactured by France and Italy, will be cold tested before shipping to Japan. For this purpose, the European Joint Undertaking for ITER, the Development of Fusion Energy (“Fusion for Energy”, F4E) and the European Voluntary Contributors are collaborating to design and set-up a coil test facility (CTF) and to perform the acceptance test of the 18 JT-60SA Toroidal Field (TF) coils. The test facility is designed to test one coil at a time at nominal current and cryogenic temperature. The test of the first coil of each manufacturer includes a quench triggered by increasing the temperature. The project is presently in the detailed design phase.

The ITER Conductor Procurement Arrangement requires that all the conductors have to be qualified before the production phase. The objects of the work here presented are the Poloidal Field Coils conductors, specifically the PF1/6 and the PF2 conductors. The realization of the samples comprised the jacketing, the compaction and the straightening of the CICC, together with the design and manufacturing of the bottom hairpin box and the upper termination. The instrumentation and the related jacket machining completed the preparation of the samples. The samples have been assembled according to the specification defined by ITER and have been shipped to SULTAN facility. This paper describes all the activities performed during the preparation, the features of the samples and all the related issues.

It is now a matter of fact that parameters such as cable twist pitch (TP) and void fraction (VF) have a strong impact on Cable-in-Conduit Conductor performances. A proper choice of their values in the direction of raising the former (TP) and lowering the latter (VF) has been proven to considerably enhance the transport properties, though increasing the AC losses, and to appreciably reduce the conductor degradation with electromagnetic and thermal loading cycles. It has been also demonstrated that a further route for CICCs performance improvement is represented by a suitable optimization of the conductor shape with respect to the electromagnetic force distribution. In this sense, CIC conductors with high aspect ratio rectangular geometry, if properly oriented, have shown a better response to high electromagnetic pressure, as proved by experimental evidences.

In the framework of the JT-60SA project, aiming at upgrading the present JT-60U tokamak toward a fully superconducting configuration, the detailed design phase led to adopt for the three main magnet systems a brand new design. Europe (EU) is expected to provide to Japan (JA) the totality of the toroidal field (TF) magnet system, while JA will provide both Equilibrium field (EF) and Central Solenoid (CS) systems. All magnet designs were optimized trough the past years and entered in parallel into extensive experimentally-based phases of concept validation, which came to maturation in the years 2009 and 2010. For this, all magnet systems were investigated by mean of dedicated samples, e.g. conductor and joint samples designed, manufactured and tested at full scale in ad hoc facilities either in EU or in JA. The present paper, after an overall description of magnet systems layouts, presents in a general approach the different experimental campaigns dedicated to qualification design and manufacture processes of either coils, conductors and electrical joints. The main results with the associated analyses are shown and the main conclusions presented, especially regarding their contribution to consolidate the triggering of magnet mass production. The status of respective manufacturing stages in EU and in JA are also evoked.

The primary issue facing Nb3Sn cable-in-conduit conductors (CICCs) has been degradation, often observed during electromagnetic cycling, in the current sharing temperature, n-index, and critical current, in respect to measured strands values, due to strain effects. In the last years, work performed mostly relating to fusion magnet technology has led to a better understanding of the parameters required to improve the constraints imposed by the brittle nature of the Nb3Sn filaments, such as cables low void fraction and long twist pitch sequence. On the other hand, experimental campaigns on bent Nb3Sn wires, pre-compressed into a stainless-steel jacket, have shown that an appreciable decrease in the n-index values already occurs at the strand level, well below the irreversible mechanical load regime for filament breakage. This result is explained, with the support of simulation results, taking into account the broadening of the critical current distribution on wires cross section due to the presence of the jacket and to bending strain. Considering the CICC's layout impact on their overall performances, the effective resemblance between pre-compressed bent wires and strands inside cables is emphasized, and an innovative interpretation of cabled conductors test results is given, suggesting a tool to predict their performances in terms of the n-index versus critical current relation of constituting strands, characterized under simultaneous pre-compression and bending strain.

In the framework of the final design activities related to the ITER Toroidal Field (TF) coils, following the very good results obtained during the TF Prototype Conductors measurement campaign, Fusion For Energy, the European Domestic Agency for ITER, has launched the conductor performance qualification phase in order to confirm the final ITER TF conductor design. Six conductor lengths have been cabled by different strand types coming from different producers: LUVATA Pori, Oxford Instruments Superconducting Technology (OST), ALSTOM (for two different strand layouts) and European Advanced Superconductors (EAS) (for two different strand layouts). ENEA has been in charge of performing QA and monitoring activity during the conductor production at LUVATA Fornaci di Barga and TRATOS Cavi, and to make visual/destructive tests over some spare lengths of the samples in order to have a detailed characterization of the produced conductors. The conductor unit lengths have been successfully manufactured and all the different work phases are here described, along with a discussion of the encountered problems and the adopted solutions. Five samples have been successfully tested in the SULTAN facility during 2009.

In the framework of the JT-60SA design activities, EU home team has defined a reference layout for the Toroidal Field conductor: it is a slightly rectangular Cable-In-Conduit NbTi conductor, operating at 25.7 kA with a peak field of 5.65 T. ENEA has assigned LUVATA Fornaci di Barga the task to produce the strands and to perform cabling, whereas jacketing and compaction have been carried out in its own labs. The sample, successfully tested at the CRPP SULTAN facility, has been assembled in such a way as to avoid the bottom joint between the two legs, thus using a single conductor length (about 7 m). An ad-hoc developed solution to restrain the U-bent conductor section (where jacket is not present), consisting in a stainless steel He-leak tight box with an inner structure designed in order to completely block the cable, has been also developed and manufactured by ENEA, where the sample has been also assembled. Instrumentation installation and final assembly of the sample have been performed by the SULTAN team. The main aspects of the sample manufacturing and characterization are here presented and discussed.

We present the results of an extended measurement campaign carried out on available NbTi commercial strands to be used in the design of fusion reactor magnets, including candidate strands for the ITER high field Poloidal Coils, PF1/6, and for the Toroidal Field Coils of the JT-60SA Tokamak. Magnetic and transport measurements have been carried out at variable temperature and magnetic field. From magnetization cycles we were able to extract information about AC hysteretic losses, and to extend the current density determination to lower fields, thus enabling the optimization of numerical fits in a wider magnetic field range. It has been found that the normalized bulk pinning force of the material, though showing good temperature scaling throughout the explored range, cannot be described by a single function of the type balpha(1 - b)beta. On the contrary, the full summation of two contributions, each dominant in a different magnetic field range, returns a good fit of the data. Extending this 2-components description to an expression for the critical current density, a very good agreement with experimental measurements is obtained over the whole explored B, T range. Collecting a database of available strands performances, especially in the range of relatively high temperatures (T > 5.5 K) and magnetic fields (B ~ 6 T), typical of applications in large coils for fusion reactors, constitutes a sound basis for magnets design, which should be based on strand properties measured in the operation-relevant temperature and magnetic field range.

The present work illustrates the conceptual design of a new superconducting test facility, ENFASI (Enea FAcility for Superconducting Inserts), which has been mainly conceived to test long length wound conductors. It consists of a background 15 T Nb3Sn solenoid, with a warm bore of 90 cm diameter, divided into three concentric sections corresponding to high field (15 T), medium field (13 T) and low field (8 T), and globally composed of 983 turns. The three different CIC (cable-in-conduit) conductor layouts, one for each section and with rectangular cross-sections, operate at 20 kA and are made of strands with different Cu/nonCu ratios and segregated copper when needed. The facility permits to feed the sample with a current up to 70 kA and to vary the sample temperature in the range [4.5 K-77 K]. The main results of electro-magnetic, structural and thermo-hydraulic analysis which brought to the definition of the main aspects of the design, are presented.

The effect of bending strain on the transport properties of Nb3Sn strands has been the object of several investigations in the last years. We report on the performances of internal tin Nb3Sn strands with different filament twist pitches, pre-compressed by swaging into thin stainless steel tubes before the reaction heat treatment and subject to pure bending strain. We have previously reported on the comparison between the performances of the technological OST-Dipole strand (TW-strand) with those of the same strand, but with untwisted superconducting filaments (UNTW-strand). The critical current, as well as the n-index of the samples with twisted filaments gradually degraded with the applied bending strain, while an enhancement of the performances has been observed on UNTW-strands. Here, results on further experimental measurements carried out on OKSC internal tin strands with different filament twist pitches are discussed, in order to clarify some aspects of the current transfer process within the strands. Finally, we carried out an analysis through the second derivative of the V-I curve, which evidenced a peaked critical current distribution for the UNTW-strands, while TW-strands under bending showed a higher degree of non-homogeneity, proven by broader distributions.

EFDA is presently responsible for the fabrication of a large bore 12.5 T dipole magnet, for a second cable-in-conduit conductor test facility to be installed in CRPP (Switzerland). This dipole magnet is wound from square and rectangular type conductors using strands. Since the magnet is built from multiple, independently cooled double pancakes, a total of twelve joints are required to interconnect them. In order to reduce the footprint of each joint ENEA, in collaboration with EFDA, developed a special joint design that allows the two conductor-ends to interpenetrate, reducing to a minimum value the increase of the magnet cross-section at the joint position. Furthermore, this design maintains the electrical performances of the conductor and gives very low electrical resistance values, on the order of some tenths of . The properties of first prototypes of this joint were successfully tested at ENEA labs. This paper describes the novel joint design, in the dipole magnet version, and discusses some of the measured results.

The broader approach agreement between Europe and Japan includes the construction of a fully superconducting tokamak, the JT-60 Super Advanced (JT-60SA), as a satellite experiment to ITER. In particular, the whole Toroidal Field magnet system, described in this paper, will be provided to Japan by the EU. All the TF coil main constituents, i.e. conductor, winding pack, joints, casing, current leads, are here presented and discussed as well as the design criteria adopted to fulfil the machine requirements. The results of the analyses performed by the EU and JA to define and assess the TF magnet system conceptual design are reported and commented. Future work plan is also discussed.

In the framework of the 'roadmap to fusion electricity by 2050,' the design of the European DEMO machine strongly relies on available technologies. The superconducting toroidal field (TF) magnets will be built using available low-temperature superconducting strands; the ENEA winding pack (WP) proposal will exploit graded layer-wound rectangular conductors, whose design is ongoing. The WP will be encapsulated in a stainless steel casing, whose cooling has not yet been designed. In this paper, we present the thermal-hydraulic (TH) model of an entire TF coil for the European demonstration power plant DEMO. Two cooling options are proposed and investigated for the casing, whereas for the WP, the ENEA design, with multiple low-impedance hydraulic channels, is considered. The thermal coupling between WP and casing is parametrically included in the model. The TH behavior of a TF coil (WP + casing) during a plasma burn is presented and discussed, comparing the two cooling options of the casing. © 2002-2011 IEEE.

The DEMO reactor is expected to be the first application of fusion for electricity generation in the near future. To this aim, conceptual design activities are progressing in Europe (EU) under the lead of the EUROfusion Consortium in order to drive on the development of the major tokamak systems. In 2014, the activities carried out by the magnet system project team were focused on the toroidal field (TF) magnet system design and demonstrated major achievements in terms of concept proposals and of consolidated evaluations against design criteria. Several magnet system RD activities were conducted in parallel, together with broad investigations on high temperature superconductor (HTS) technologies. In this paper, we present the outcomes of the work conducted in two areas in the 2014 magnet work program: 1) the EU inductive reactor (called DEMO1) 2014 configuration (power plant operating under inductive regime) was the basis of conceptual design activities, including further optimizations; and 2) the HTS RD activities building upon the consolidated knowledge acquired over the past years. © 2016 IEEE.

The effect of transverse loads on strands has been pointed as a possible cause of the difference observed when scaling transport properties of single strands to those of cable-in-conduit conductors. Single multifilamentary strands inside cables are in fact subject to bending strain due to the electromagnetic forces at operating conditions and to the geometrical layout. Here the influence of pure bending strain, applied in combination with a longitudinal strain, on the critical current of advanced strands for ITER has been studied. The tested samples are single strands inserted inside a thin stainless steel jacket and wound on stainless steel barrels. After the heat treatment, a pure bending strain has been applied transferring the wires on different diameter mandrels, using ad-hoc developed and qualified techniques. Transport critical current has been measured on the single strands before and after the steel jacketing, as well as after the additional application of two different values of maximum bending strain: 0.5% and 0.25%. This was the best choice in order to verify experimentally whether the so-called long twist pitch condition can be applied for the selected strands. The distribution of the bending strain over the strand cross-section has been calculated with finite element numerical codes, and the expected critical current degradation in the limiting cases of short and long twist pitch has been computed and compared with experimental data.

Demonstration fusion reactors (DEMO) are being envisaged to be able to produce commercial electrical power. The design of the DEMO magnets and of the constituting conductors is a crucial issue in the overall engineering design of such a large fusion machine. In the frame of the EU roadmap of the so-called fast track approach, mechanical studies of preliminary DEMO toroidal field (TF) coil system conceptual designs are being enforced. The magnetic field load acting on the DEMO TF coil conductor has to be evaluated as input in the FEM model mesh, in order to evaluate the stresses on the mechanical structure. To gain flexibility, a novel approach based on the meshless method of radial basis functions (RBF) has been implemented. The present paper describes this original and flexible approach for the generation and mapping of magnetic load on DEMO TF coil system. © 2015 Elsevier B.V. All rights reserved.

The comprehension of the current re-distribution phenomenon inside multi-filamentary conductors is a crucial point for the design of ITER-relevant coils, as it is by now assessed that current non-uniformity among cable sub-stages may strongly deteriorate Cable-in-Conduit Conductors (CICC) performances. The only feasible way to get information about the current flowing inside CICC sub-stages is an indirect evaluation by self-field measurements in regions very close to conductor surface. A 7m full-size NbTi conductor (Bus-Bar III) has been used as short-circuit during the test of an ITER Toroidal Field Coil HTS current lead at FzK. Its relatively simple shape and the absence of any other magnetic field source (background coils, etc.), made BBIII one of the most desirable candidate for a reliable measurement of the current distribution under controlled conditions. This is why it has been ad hoc instrumented with different arrangements of Hall-probes (rings and arrays), as well as with transverse and longitudinal voltage taps. This paper gives a qualitative interpretation of the current (re-)distribution events inside the conductor as derived from the analysis of the Hall sensors and the voltage taps signals, during Tcs measurements and as a function of different dI/dt. It has been shown that Hall probes represent a very reliable tool to investigate this issue. In fact, re-distribution phenomena have been clearly observed during transition, and even far before reaching Tcs, when voltage transverse signals had not yet showed any appreciable onset.

The large size of the magnets for DEMO calls for very large operating current in the forced flow conductor. A plain extrapolation from the superconductors in use for ITER is not adequate to fulfill the technical and cost requirements. The proposed DEMO TF magnets is a graded winding using both Nb3Sn and NbTi conductors, with operating current of 82 kA @ 13.6 T peak field. Two Nb3Sn prototypes are being built in 2014 reflecting the two approaches suggested by CRPP (react&wind method) and ENEA (wind&react method). The Nb3Sn strand (overall 200 kg) has been procured at technical specification similar to ITER. Both the Nb3Sn strand and the high RRR, Cr plated copper wire (400 kg) have been delivered. The cabling trials are carried out at TRATOS Cavi using equipment relevant for long length production. The completion of the manufacture of the two 20 m long prototypes is expected in the end of 2014 and their test is planned in 2015 at CRPP. In the scope of a long term technology development, high current HTS conductors are built at CRPP and ENEA. A DEMO-class prototype conductor is developed and assembled at CRPP: it is a flat cable composed of 20 twisted stacks of coated conductor tape soldered into copper shells. The 10 kA conductor developed at ENEA consists of stacks of coated conductor tape inserted into a slotted and twisted Al core, with a central cooling channel. Samples have been manufactured in industrial environment and the scalability of the process to long production lengths has been proven. © 2015 Elsevier B.V. All rights reserved.

Losses in multifilamentary superconducting conductors depend both on internal magnetic field amplitude and its changing rate, which determine the magnetization of the sample during a field cycle. Analytical computation of hysteresis and coupling losses is possible whenever the internal field can be assumed to be equal to the external one; otherwise, approximate solutions for the internal field time dependence have to be assumed when the conductor time constants are comparable with field variation periods. The internal field profile, the changing rate and the losses associated to an external field rapidly varying as compared to the coupling time constant of a cable-in-conduit conductor solenoid, have been calculated. The results have been compared with the losses measured for various field amplitudes and ramp rates. The calculations showed that a `saturated regime' is reached during ramping up and down. In this situation, no more coupling currents can be activated in part of the conductor, as the external strands are saturated at I c , and the external field variations are followed by the internal field. Experimental evidence of the saturated regime is reported.

The DEMO plant will demonstrate by mid-century the feasibility of electric power generation by nuclear fusion. In the scope of design studies coordinated by the European Commission, CRPP and ENEA are developing two force flow conductor layouts as candidates for the toroidal field (TF) coils of DEMO. The operating conditions include 82 kA operating current and 13.5 T peak field, placing the DEMO TF conductor at substantially higher performance compared to the ITER TF (68 kA/11.5 T). The innovative winding layout is graded, layer wound with Nb3Sn/NbTi hybridization for both conductor designs, aiming at minimizing the size and the cost of the superconductor: one of the conductors is a wind and react cable-in-conduit conductor with reduced void fraction and has a rectangular shape. The other conductor is a react and wind flat cable with copper segregation and thick conduit assembled by longitudinal weld. The conductor designs were first drafted in 2012 and reviewed in 2013 based on a first round of assessments and on an updated requirements catalog. The manufacture of full-size prototypes is planned in 2014. © 2013 IEEE.

JT-60SA is a fully superconducting coil tokamak to be built under the framework of the EU-JA Broader Approach Agreement, and it aims to contribute to the complement of the ITER experiments and to the DEMO reactor design by the study of steady-state high-beta plasma experiments. The conceptual design of the JT-60SA tokamak and the peripheral systems has been carried out in close collaboration of EU and Japan from middle of 2006, and the JT-60SA Conceptual Design Report which involves an analysis of plasma physics and engineering design is summarized in May 2007. This paper intends to present an overall view of the JT-60SA project, and the latest design status of the key engineering issues such as superconducting magnets, divertor target, plasma heating devices, and magnet power supplies.

Two short length samples have been prepared and tested in SULTAN to benchmark the performance of high current density, advanced Nb3Sn strands in the large cable-in-conduit conductors (CICC) for ITER. The cable pattern and jacket layout were identical to the toroidal field model coil conductor (TFMC), tested in 1999. The four conductor sections used strands from OST, EAS, OKSC and OCSI respectively. The Cu:non-Cu ratio was 1 for three of the new strands, compared to 1.5 in the TFMC strand. The conductors with OST and OKSC strands had one Cu wire for two Nb3Sn strands, as in TFMC. In the EAS and OCSI conductors, all the 1080 strands in the cable were Nb3Sn. A dc test under relevant load conditions and a thermal-hydraulic campaign was carried out in SULTAN. The CICC performance was strongly degraded compared to the strand for all the four conductors. The current sharing temperature at the ITER TF operating conditions (jop = 286 A/mm2, B = 11.15 T) was lower than requested by ITER.

It is well assessed that the cable in conduit conductors (CICC) performances are affected by the current distribution among their strands. In many experiments, during Tcs measurements, it has been evidenced that a current re-distribution among the cable sub-stage bundles appears just before the conductor transition. This is deduced mainly from the change of the magnetic field measured in the conductor proximity, even with a constant cable transport current. This paper presents the experimental evidence that the current begins to re-distribute well before any detectable voltage development. Such a phenomenon, repeatable and depending on the overall transport current, has been observed by means Hall probe arrays. These probes have been designed with suitable sensitivity and distribution, to permit also the reconstruction of the current distribution inside the CICC by means of the solution of the inverse magnetostatic problem. Some results from the experimental data taken from the NbTi BusBar III and the Poloidal Field Insert samples, tested respectively in the TOSKA facility at FZK, and in SULTAN at PSI in 2004, are discussed here.

The 'NAFASSY' (NAtional FAcility for Superconducting SYstems) facility is designed to test wound conductor samples under high-field conditions at variable temperatures. Due to its unique features, it is reasonable to assume that in the near future NAFASSY will have a preeminent role at the international level in the qualification of long coiled cables in operative conditions. The magnetic system consists of a large warm bore background solenoid, made up of three series-connected grading sections obtained by winding three different Nb3Sn Cable-in-Conduit Conductors. Thanks to the financial support of the Italian Ministry for University and Research the low-field coil is currently under production. The design has been properly modified to allow the system to operate also as a stand-alone facility, with an inner bore diameter of 1144 mm. This magnet is able to provide about 7 T on its axis and about 8 T close to the insert inner radius, giving the possibility of performing a test relevant for large-sized NbTi or medium-field Nb3Sn conductors. The detailed design of the 8 T magnet, including the electro-magnetic, structural and thermo-hydraulic analysis, is here reported, as well as the production status. © 2015 IOP Publishing Ltd.

Publisher Summary This chapter illustrates the design and manufacturing of the insert coil and also gives a description of the test facility layout. A magnet wound with a sub-size NbTi conductor has been realized to study its characteristics in view of the realization of the full-size conductor for the ITER Poloidal Field Coils. The coil has been designed and constructed by ENEA and Ansaldo CRIS, with a conductor provided by Europa Metalli Superconductor and consisting of 36 Ni-coated strands twisted in a three stage cable, jacketed and compacted in a SS round tube. A self consistent unit has been manufactured assembling the insert module and its mechanical supporting structure, to allow easy installation inside the background magnet of the ENEA test facility where the coil will be tested under representative conditions for the ITER PF magnets. After having determined stability conditions with uniform transport current, the same rims will be repeated parametrically varying non-uniform current distribution inside the cable. The same kind of measurements will then be repeated after reversing the helium flow. This will cause the transition region to move along the conductor length, owing to the interplay between the self-field profile and the steep temperature profile along the conductor length.

The Bus Bar III (BBIII), fabricated within the Toroidal Field Model Coil Task of the International Thermonuclear Experimental Reactor (ITER), was tested at the Forschungszentrum Karlsruhe, Germany, in the spring of 2004. The BBIII consists of an approximately 7 m long NbTi dual-channel conductor with a thick square stainless steel jacket, cooled by forced flow supercritical He. It was energized with currents up to 80 kA and operates in its self magnetic field (up to /spl sim/0.8 T). The BBIII was instrumented with Hall-probe heads and arrays, voltage rings and longitudinal voltage taps for electro-magnetic measurements, in order to get experimental data to be used for the validation of a recently developed hybrid thermal-hydraulic electro-magnetic code (THELMA), as well as for the assessment of the possibility of performing a reliable reconstruction of the current distribution in the conductor cross section under controlled conditions. In the tests, current ramps at different rates were applied to characterize the conductor time constants, while two different resistive heaters (one upstream of the BBIII inlet, another one directly on the BBIII jacket) were separately operated in order to approach current sharing in the conductor and to observe the related current re-distribution. In this paper, a summary of the collected experimental results is presented, with particular emphasis on those aspects more relevant for the forthcoming THELMA analysis.

A dedicated facility prepared by ENEA (Italian Agency for Energy and Environment) for the cryogenic testing of by-pass diodes for the protection of the CERN Large Hadron Collider main magnets will be described. This experimental activity is in the frame of a contract awarded to OCEM, an Italian firm active in the field of electronic devices and power supplies, in collaboration with ENEA, for the manufacture and testing of all the diode stacks. In particular, CERN requests the measurement of the reverse and forward voltage diode characteristics at 300 K and 77 K, and endurance test cycles at liquid helium temperature. The experimental set-up at ENEA and data acquisition system developed for the scope will be described and the test results reported.

This paper describes the results of the measurements performed in June 2011 on a NbTi cable-in-conduit conductor curved joint in the ENEA Superconductivity Laboratories (C.R. Frascati). The effectiveness of the joint design, patented by ENEA, had been already verified for the Nb3Sn EDIPO conductor; in this case, it has been applied to a NbTi cable-in-conduit conductor. The joint was manufactured and instrumented by ENEA, using part of a JT-60SA toroidal field conductor sample (namely the TFCRS) produced by Luvata Italy and successfully tested in SULTAN. For testing purposes, the NbTi 'shaking hands'-type joint was hosted in a variable temperature cryostat with a background superconducting magnet. The sample was used as the secondary coil of a transformer having the background magnet as primary coil. Measurements were preceded by preliminary code simulation studies computed in the Matlab workspace, and by 3-D finite element method analyses, to estimate the test requirements and its feasibility. The joint resistance value, found to be less than 2 nΩ has been derived both by the sample current decay time constant evaluation and by calorimetric calculations. After a short description of the joint design and fabrication, as well as of the test instrumentation and geometry, the joint resistance evaluation will be presented, together with the data analysis procedures and results. © 2012 IEEE.

The European and the Korean contributions to the ITER Project in terms of cable-in-conduit conductors have been assigned to the Italian Consortium for Applied Superconductivity (ICAS). ICAS' charter members are: ENEA, with the role of general supervision and monitoring, CRIOTEC Impianti S.r.l., in charge of jacketing activities, and TRATOS Cavi S.p.A., with the duty of cable manufacturing. The ITER production consists of about 40 km of Nb3Sn toroidal field (TF) conductor and 22 km of NbTi poloidal field (PF) conductor, for coils 1 and 6. As part of the EU contribution to the construction of the JT-60SA Japanese Tokamak Reactor, ICAS is also in charge of producing about 28 km of NbTi TF conductor, required for the whole TF magnet system. Thanks to the high level and unique expertise available at ICAS, the consortium has been also assigned to manufacture the ITER Central Solenoid conductor samples to be tested in SULTAN for the final, positive, assessment of the updated design. In the present paper, the steps leading to the successful qualification of the main production processes will be described. What is more, the manufacturing of the first conductor unit lengths of the 'production phase,' i.e., the ones that will be wound to operate in the actual ITER TF and JT-60SA TF superconducting magnets, will be reported. The fabrication of the ITER CS conductor samples will be illustrated as well. © 2012 IEEE.

The on-going construction of the ITER and JT-60SA tokamak reactors requires the production of a large quantity of Cable-In-Conduit Conductors (CICCs) to be employed in the superconducting magnets for plasma confinement and control. A new company, named ICAS (Italian Consortium for Applied Superconductivity), was established thanks to the large expertise in this field that has been present in Italy for nearly three decades. ICAS is a consortium among ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development) and two Italian private companies: CRIOTEC Impianti, which owns an 850m long jacketing line for CICCs, and TRATOS Cavi, which has a large experience in cable manufacturing for various applications. ICAS' current activities are aimed at supplying to Fusion for Energy the European quote of ITER TF and PF conductors and the whole JT-60SA TF conductor quantity, as well as at supplying to the Korean Domestic Agency about 19 km of ITER TF conductor. This paper offers an overview of ICAS' technical capabilities and facilities, and outlines the first outcomes from the first productions.

Superconductors for the ITER Poloidal Field Coils are large cable-in-conduit conductors (CICC) made of NbTi strands encased in a round-in-square stainless steel jacket. Three prototype conductor sections for poloidal field coils PF1/6, PF2/3/4 and PF5 have been fabricated in collaboration of the domestic RF, CN and EU agencies and tested in SULTAN Test Facility at the nominal operating conditions. The test aimed to characterize the DC and AC behavior of the conductors. The DC test was focused on the current sharing temperature (T-cs) at the nominal operating current and nominal operating background field. The take-off electric field at the nominal Helium mass flow rate was investigated versus the cable current density over a broad range of field and temperature. The AC loss measurement was performed before any electromagnetic loading and after a number of load cycles in order to define the impact of cyclic loads on the coupling currents constant of the cable. From the test results in SULTAN test facility, the margins in normal operation and the limits of the operation range of the ITER PF conductors are assessed.