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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 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.

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.

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.

ENEA is currently involved in the manufacture of cable-in-conduit conductors (CICCs) for the magnets of the nuclear fusion facilities ITER and JT-60SA, as well as for laboratory superconducting magnets based on CICCs, as those of the NAFASSY and of the Nijmegen-High Field Magnetic Laboratory facilities. The acceptance tests of such industrial productions include the destructive examination of both cable and conductor samples. As widely experienced, during the manufacturing processes, deformations of both copper and superconducting wires may occur due to the large compaction forces involved to which the cabled structure is subjected. In order to evaluate the maximum extent to which typical deformations can be accepted at the level of the single wire within a cable, the characterization of wires artificially deformed has been carried out in terms of mechanical, structural, and electromagnetic properties. In addition, some relevant wire sections extracted from the destructive examination samples have been also characterized. Experimental results demonstrate the negligible influence on the overall conductor performances to be expected for the typical deformations that occur during cable and conductor manufacture. © 2014 IEEE.

In parallel with the programme to optimize the operation with a conventional divertor based on detached conditions to be tested on the ITER device, a project has been launched to investigate alternative power exhaust solutions for DEMO, aimed at the definition and the design of a divertor tokamak test facility (DTT). The DTT project proposal refers to a set of parameters selected so as to have edge conditions as close as possible to DEMO, while remaining compatible with DEMO bulk plasma performance in terms of dimensionless parameters and given constraints. The paper illustrates the DTT project proposal, referring to a 6 MA plasma with a major radius of 2.15 m, an aspect ratio of about 3, an elongation of 1.6–1.8, and a toroidal field of 6 T. This selection will guarantee sufficient flexibility to test a wide set of divertor concepts and techniques to cope with large heat loads, including conventional tungsten divertors; liquid metal divertors; both conventional and advanced magnetic configurations (including single null, snow flake, quasi snow flake, X divertor, double null); internal coils for strike point sweeping and control of the width of the scrape-off layer in the divertor region; and radiation control. The Poloidal Field system is planned to provide a total flux swing of more than 35 Vs, compatible with a pulse length of more than 100 s. This is compatible with the mission of studying the power exhaust problem and is obtained using superconducting coils. Particular attention is dedicated to diagnostics and control issues, especially those relevant for plasma control in the divertor region, designed to be as compatible as possible with a DEMO-like environment. The construction is expected to last about seven years, and the selection of an Italian site would be compatible with a budget of 500 M€.

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.

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.

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.

We review progress in the design of high field superconducting cable-in-conduit conductors (CICCs) for fusion applications, with special attention to the results of recent key experiments, leading to the state-of-the-art CICC technology: the ITER Toroidal Field and Central Solenoid programs, the EFDA Dipole conductor development program, the NHFML Hybrid Magnet project, the EU-TF Alt conductor demonstration, and the CRPP React & Wind flat cable test. For these projects, the main CICC design driver was the mitigation of Nb3Sn conductor performance degradation with electro-magnetic loading cycles. This was achieved by proper choice of cable layout and of conductor geometry, depending on the specific operating conditions and project requirements. In all cases, the necessity to limit cable movements inside the conductor jacket was identified to be of crucial importance. The main aspects of CICC manufacture are also discussed here, at least for what is the experience gained by the authors in both CICC jacketing and cabling processes. Finally, the state of the art of high-temperature superconducting (HTS) cables is discussed: at present, this technology is still in its infancy, but it is highly likely that major technological improvements could eventually lead to a widespread use of HTS CICCs. © 2015 IOP Publishing Ltd.

FAST (Fusion Advanced Studies Torus), the Italian proposal for a European satellite facility to ITER, is a compact tokamak ( R0= 1.82 m, a= 0.64 m, triangularity δ = 0.4) able to investigate non linear dynamics effects of alpha-particle behavior in burning plasmas and to test technical solutions for the first wall/divertor directly relevant for ITER and DEMO (e.g.: full-tungsten wall and divertor and advanced liquid metal divertor). The machine is designed to operate with Deuterium plasmas in a dimensionless parameter range close to that of ITER and to access advanced tokamak regimes with long pulse duration with respect to the current diffusion time. It foresees a maximum magnetic field on plasma axis of 8.5 T and a maximum plasma current of 8 MA. In the present design phase, the feasibility of a superconducting solution for the magnet system is being investigated by ENEA. It consists of 18 Toroidal Field, 6 Poloidal Field and 6 Central Solenoid module coils, all of which wound by Nb3Sn and NbTi Cable-In-Conduit Conductors. All the main aspects driving the magnets' design, from mechanical to neutronic and thermal analyses, are here presented and discussed.

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.

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.

The JT-60SA is a fusion experiment designed to contribute to the early realization of fusion energy, by providing support to the operation of ITER, by addressing key physics issues for ITER and DEMO and by investigating how best to optimize the operation of the next fusion power plants that will be built after ITER. It is a combined project of the JA-EU Satellite Tokamak Program under the Broader Approach (BA) Program and JAEA's Program for National Use, and it is to be built in Naka, Japan, using the infrastructure of the existing JT-60U experiment. This paper describes in detail the design of the JT-60SA Toroidal Field magnet and shows the strong points of each foreseen solution. Additional information about manufacturing procedures is given and technological issues are reported and critically analysed.

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.

In superconducting coils for fusion applications, the cable-in-conduit (CIC) conductor is the most adopted solution. ENEA reported last year about a newly developed compact joint, successfully tested for Nb3Sn conductors in a configuration suitable for the EFDA Dipole which is currently in construction phase. The main advantages of this new concept are: low room occupancy, low price and easiness of manufacturing process. In the present paper, new results coming from the characterization of NbTi joint samples are presented, suggesting new applications for this innovative junction. A joint layout optimization study by means of an electrical simulation code is included as well.

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.

A 12.5 T superconducting dipole magnet (European DIPOle, EDIPO) has been designed by EFDA and it is now being procured within the framework of the European Fusion Programme in order to be installed in CRPP-PSI. This saddle-shaped magnet is designed to reach 12.5 T in a 100 times 150 mm rectangular bore over a length of about 1.5 m in order to test full size conductor samples that shall be produced during the ITER magnets procurement. The magnet uses Cable In Conduit Conductor (CICC) technology and the cables are made of high Jc (about 2300 A/mm2 at 4.2 K, 12 T) superconducting strands. In this paper the main magnet parameters are given together with the key supporting electromagnetic, mechanical and thermal analyses. An update on the general status of the procurement of the strand, conductors, dipole magnet and facility is also given together with the key results of the on-going supporting R&D.

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.

The upgrade of JT-60U to JT-60 Super Advanced (JT-60SA), a fully superconducting tokamak, will be performed in the framework of the Broader Approach (BA) agreement between Europe (EU) and Japan. In particular, the Toroidal Field (TF) system, which includes 18 coils, is foreseen to be procured by France, Italy and Germany. This work covers activities from design and manufacturing to shipping to Japan. The present paper is mainly devoted to the analyses that lead to the conductor design and to the technical specifications of the joints for the JT-60SA TF coils. The conductor geometry is described, which is derived from Cable-In-Conduit concept and adapted to the actual JT-60SA tokamak operating conditions, principally the ITER-like scenario. The reported simulations and calculations are particularly dealing with the stability analysis and the power deposition during normal and off-normal conditions (AC losses, nuclear heating). The final conductor solution was selected through a trade-off between scientific approach and industrial technical orientation. Besides, the TF system connections layout is shown, derived from the industrially assessed twin-box concept, together with the associated thermo-hydraulic calculations ensuring a proper temperature margin.

The upgrade of JT-60U magnet system to superconducting coils (JT-60SA) has been decided by both parties of Japanese government (JA) and European commission (EU) in the framework of the Broader Approach (BA) agreement. The magnet system for JT-60SA consists of 18 toroidal field (TF) coils, a Central Solenoid (CS) with four modules, seven Equilibrium Field (EF) coils. The TF case encloses the winding pack and is the main structural component of the magnet system. The CS consists of independent winding pack modules, which is hung from the top of the TF coils through its pre-load structure. The seven EF coils are attached to the TF coil cases through supports which include flexible plates allowing radial displacements. The CS modules operate at high field and use Nb3 Sn type superconductor. The TF coils and EF coils use NbTi superconductor. The magnet system has a large heat load from nuclear heating from DD fusion and large AC loss. This paper describes the technical requirements, the operational interface and the outline of conceptual design of the superconducting magnet system for JT-60SA.

The behaviour of Ag films grown on Ni–Cr substrate subjected to high-temperature annealing was studied to determine their microstructural, mechanical and chemical stability. Several buffer layers, including native oxide, MgO, CeO2 and Pd, were used to hinder Ag film degradation. Ag films, as well as the employed buffer layers, were deposited by e-beam evaporation. Ag film thickness ranged from tenths up to several μm. The main degradation of the Ag film after annealing in air is the formation of micrometric holes by film dewetting. The behaviour of films grown on oxidized Ni–Cr is in broad agreement with results reported in literature, whereas the use of either MgO or CeO2 layers strongly suppressed hole formation. Both oxides, and particularly the latter, were demonstrated to be effective barriers against Ni and Cr diffusion towards the Ag layer. However, these architectures showed poor adhesion and blister formation occurred. Moreover, severe delamination in long samples was observed. The use of an addi...

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.

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.

Within the framework of ITER-related projects, new tasks have been recently launched by EFDA CSU Garching (European Fusion Development Agreement Close Support Unit Garching), for the definition and production on industrial scale of advanced Nb3Sn strands, to be used in the manufacturing of the ITER high field CS and TF magnets. We performed an extended characterization of the advanced Nb3Sn strands coming from different European companies, in terms of strand layout (diameter, thickness of Cr coating, Cu:non-Cu ratio), critical transport current, RRR, and hysteresis losses. The results of the measurement campaign show that the upgraded strands meet the latest ITER requirements, with an overall critical transport current of at least 200 A (at 12 T, 4.2 K), equivalent to a non-Cu Jc of about 800 A/mm2, a Cu:non-Cu ratio of about 1, a strand diameter of 0.81 mm, and with non-Cu hysteresis losses limited to less than 1000 kJ/m3 on a plusmn3 T field cycle at 4.2 K

In the framework of the International Thermonuclear Experimental Reactor (ITER) R&D programme part concerning the Nb3Sn cable-in-conduit conductor of the Toroidal Field coils, a dedicated programme of action was launched within the European Fusion Technology Programme, designed to improve the investigation of the impact of bending strain on recently developed industrially advanced Nb3Sn strands. In order to ensure that the ITER TF coils experience relevant mechanical conditions, the Nb3Sn strands were jacketed inside a 0.2 mm thick stainless steel tube, simulating the mechanical influence of the jacket on the cable strands. In this paper we will describe in detail the four candidate methods that were investigated for imposing a pure controlled bending strain on the jacketed strands. A common feature of these methods is that they study the reaction of the strand on a heat treatment mandrel and transfer it onto a test mandrel with a different diameter. In practice, the imposed bending strain is limited to 0.5%. Two solutions are considered: a reduction or an increase in diameter of the mandrel. Also, two options are possible for the jacket removal at strand ends for connection to the current leads: before or after the heat treatment. The four options are tested to provide an area of investigation that is as large as possible. Specific support together with tool design and manufacture will be presented. A general comparison of all options as regards specified criteria is also carried out in order to precisely define an action process of the bending application method developed. We find finally that the best approach is the method in which the radius is increased and the stainless steel jacket is removed before heat treatment.

A NbTi 36 strands cable-in-conduit conductor will be wound in a coil (SEx-Up Experiment) in the framework of a study of the performance of a subsize cable for the Poloidal Field Coils of ITER. The basic multifilamentary NbTi strand has been extensively characterized. The electrical characterization enabled us to get a database, essential for the future analysis of the experimental data, and to find the strand characteristic parameters for the I/sub c/(H, T) fit. We measured critical temperature vs. applied magnetic field, transport critical current vs. applied magnetic field, and magnetization vs. applied magnetic field and temperature. The last measurement gives the hysteresis losses, and implicitly the critical current values; magnetization data have been therefore correlated with direct J/sub c/ transport measurements, taking into account the self-field effect. A good match has been found, so magnetization data allowed us to enlarge the range of field and temperature toward regions in which the transport measurements are more difficult, namely low fields or high temperatures.

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 paper presents the results of a numerical analysis campaign which studies the steady-state behavior and the typical tests (current-sharing temperature and critical current measurements) foreseen for the SULTAN samples of the ITER TF reference conductor, with special emphasis to the current and electric field distribution among and along the sub-cables. In this analysis, the sample geometrical parameters (twist pitch, joint/termination length) and some electrical parameters (joint, termination and inter-bundle resistance) are supposed to range in their design or their measured boundaries, in order to understand their individual effect on the conductor performances and the test conditions.

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.