Search

The design of a superconducting magnet system of a fusion reactor candidate is usually based on the Cable in Conduit Conductor (CICC) concept. CICC consist of complex structures with several hundreds of highly packed, multistage twisted superconductor and copper strands, cooling structures – all wrapped in thin steel foils and jacketed in relatively thick stainless-steel pipes. As recently demonstrated, such complex morphology can be captured by fully 3D X-ray tomography. One of the three alternative winding pack (WP) options for the toroidal field coils of the European DEMO superconducting magnet system as proposed by the Italian ENEA, features a layer-wound WP design adopting a wind-and-react conductor with high aspect ratio rectangular cross section. A multi-physics modelling requiring high accuracy from the strand to the full-size conductor scale was tested on two X-ray tomography systems. For the microstructural integrity analysis of the internal tin one mm diameter Nb3Sn strand, manufactured by Western Superconducting Technologies, a microtomograph with ˜1 μm feature recognition was used. For the full-size conductor (66 x 25 mm2 internal dimensions) a high energy (> 300 kVp) microtomograph with voxel resolution of ˜20 μm was employed. Voids and/or geometrical imperfections were detected, accurately pin-pointed and locally investigated at high resolution before and after heat treatment. Dimensional measurements on the 3D internal structure of the strand showed the small dimensional changes of the strand after the heat treatment. A significant result was the determination of the twist-pitch factor in a purely non-invasive way. The 3D reconstructions of the CICC cables were used to verify the cabling and compaction processes as well as the structural integrity of the strands, wrapping foils and spiral cooling structures. Also, 3D reconstruction of strand trajectories provided a strong input for a multiphysical modelling/interpretation of mechanical–electrical properties of CICC in cryogenic–electromagnetic environments.

The knowledge of the strain distribution of Nb3Sn filaments in a CICC cross-section is a key parameter in understanding the performance and its evolution when the cable undergoes electromagnetic cyclic loading. This work focuses on the Wind and React (WR1) rectangular CICC sample designed by ENEA for the European DEMO project, that has exhibited a current sharing temperature (Tcs) above 6.9 K at 13 T and 81.7 kA, during its qualification tests in the EDIPO facility at the Swiss Plasma Center. Compared to the critical current tests on the constituting wires, the observed CICC performance corresponded to an effective strain value of −0.55%. This work aims to interpret this result on the basis of the analysis of the strain distribution inside the cable itself. Susceptibility versus temperature of WR1 constituting free standing strands has been measured after etching of the stabilizing copper, and it has been compared with the susceptibility measured on the cable during the test campaign in EDIPO. A well-known and already assessed method has been used, starting from susceptibility measurements, to compute the Tc distribution in the cable cross-section and, subsequently, to infer from that the thermal strain distribution. The obtained results give an indication that the strain distribution in WR1 cable is narrow, compared to the strain distribution calculated with the same method for most ITER TF and CS conductors.

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.

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.

With the availability of High-Temperature Superconducting (HTS) prototype cables, based on high-performance REBCO Coated Conductor (CC) tapes, new designs can now be made for large bore high-field inserts in superconducting solenoids, thus extending the magnet operating point to higher magnetic fields. In this work, as an alternative approach to the standard trial-and-error design process, an optimization procedure for a HTS grading section design is proposed, including parametric electro-magnetic and structural analyses, using the ANSYS software coupled with a numerically-efficient optimization algorithm. This HTS grading section is designed to be inserted into a 12 T large bore Low-Temperature Superconducting (LTS) solenoid (diameter about 1 m) to increase the field up to a maximum value of at least 17 T. The optimization variables taken into consideration are the number of turns and layers and the circle-in-square jacket inner diameter in order to minimize the total needed conductor length to achieve a peak field of at least 17 T, while guaranteeing the structural integrity and manufacturing constraints. By means of the optimization, an optimal 360 m total conductor length was found, achieving 17.2 T with an operating current of 22.4 kA and a coil comprised of 18 × 12 turns, shortened of about 20% with respect to the best initial candidate architectural design. The optimal HTS insert has a bore compatible with manufacturing constraints (inner bore radius larger than 30 cm). A scaled HTS insert for validation purposes, with a reduced conductor length, to be tested in an advanced experimental facility currently under construction, is also mentioned.

A thermal–hydraulic model has been developed for the novel cable-in-conduit conductor (CICC) recently proposed by Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), which foresees the use of second-generation REBCO-based coated conductor tapes stacked into twisted grooves slotted into an Al core. A central hole in the Al core provides a low hydraulic impedance channel for the coolant, whereas direct cooling of the tape stacks is achieved through small channels at the bottom of the grooves and other small passages around the stack. The coolant could be either supercritical helium (SHe), for high-field/high-current applications, as in fusion magnets, or liquid nitrogen (LN2), for self-field/lower current requirements. The issues of the thermal–hydraulic performance and optimization of such a CICC are addressed for the first time in this paper. The 4C code is used to develop an adequate model for this complex 3D problem, taking advantage of the 1D nature of the coolant flow. The model, first calibrated and then validated using the results of thermal–hydraulic tests performed in cryogenic conditions at the Center for Research in Plasma Physics (CRPP), is used to evaluate the flow repartition among the different cooling paths in the CICC for SHe at 4.5 K and for LN2 at 77 K. The model is finally used to perform parametric analyses, varying, e.g., the central hole diameter, as well as the size of the grooves at the bottom of the tapes, within the limits imposed by structural constraints, and to assess the cooling capability of the cable.

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.

In 2013, the EUROfusion organization launched the preconceptual design studies of the fusion DEMO tokamak. Several options of superconducting magnet design are investigated within the scope of the project. The magnets based on Nb3Sn, namely the Toroidal Field (TF) and Central Solenoid (CS) coils, can be made either using the Wind&React technology, as in ITER, or React&Wind (RW) technology with potentially significant cost-saving benefits. Within the R&D program of the Swiss Plasma Center, the first full-size DEMO TF RW conductor prototype, RW1, was manufactured and tested in 2015–2016. The second, improved prototype, RW2, was built and tested in 2017–2018. The experimental results collected over several test campaigns performed on four assemblies of RW2 rated for 63 kA at 12.23 T are presented. The current sharing temperature $T_{{{cs}}}$ of the latest RW2 assembly in DEMO operating conditions reached 7.16 K after 1000 electromagnetic cycles and four thermal cycles. This $T_{{{cs}}}$ corresponds to the effective strain ϵeff = −0.27%, significantly exceeding the design requirement of $T_{{{cs}}}\,\geq \,\text{6.7}\,\text{K}$ . If RW2 have been used in ITER TF coils, the $T_{{{cs}}}$ at 68 kA and 10.9 T would be 7.42 K, i.e., approximately 1 K higher than most of the ITER TF samples, and that with only 55% of Nb3Sn used in ITER TF conductor. The tests of RW2 proved that the cost-saving potential of the React&Wind technology is enormous.

ENEA is currently involved in the design and manufacture of a fully high temperature superconductor (HTS) cable in the cable-in-conduit conductor (CICC) configuration exploiting commercial second generation ReBaCuO (Re: Rare Earth and Y) coated-conductors. The final cable will be composed of five slots obtained in a helically twisted aluminum central core and filled with 2G tape stacks. This conductor is designed to operate above 10 kA in 12 T background field at 4.2 K or at about 10 kA in self-field at 77 K. A first sample of about 1-m length with one fully superconductive slot has been manufactured using 15 tapes provided by Superpower, Inc. and 12 tapes from the SuNAM Company grouped in two sub-stacks divided by a Kapton foil. Each tape of the stack has been characterized individually by measuring critical current values Ic at 77 K (liquid N2 bath) in self-field and n-index. Results revealed that the tapes showed no degradation of critical current values when compared with suppliers specifications confirming that the proposed manufacturing process is fully compatible with commercial coated-conductors. Inter-tape resistance (Rinter) has also been measured and the observed dependence of Rinter on the tape position in the stack has been put in correlation with transverse stress distribution calculated by finite element models. A second sample with a full superconducting slot has been manufactured using 18 SuNAM tapes. Preliminary results on the stack transport measurements performed at 77 K in self-field will be presented and discussed. All the samples were manufactured by using already existing industrial equipments at Tratos Cavi SpA. © 2014 IEEE.

In 2013, the Superconductivity Laboratory at ENEA (the Italian National Agency for New Technologies, Energy, and Economica Developments) started a project aimed at using second generation (2G) High-Temperature Superconductors (HTS) tapes for the manufacture of a cable-in-conduit conductor. The considered layouts are based on a number of HTS tapes, which are arranged as five layered structures of twisted tapes wound on a helically slotted core with an external round jacket, and the conductor is designed to target 20 kA at operating conditions (T = 5 K, B = 15 T; or T = 77 K, self-field). With the commercialization of 2G HTS tapes, several options are now available on the market that can meet the ENEA cable specifications. Thus, for a sound design of the ENEA cable, a measurement campaign was launched aiming to determine the critical current, i.e., Ic, of commercial tapes as a function of an external applied field (up to 12 T) at different temperatures (in the 5-77 K range). Transport measurements were performed at 77 K in the high-field region, whereas the low-field and low-temperature critical currents have been determined from magnetization loops measured with a vibrating sample magnetometer system. Within a COMSOL Multiphysics environment, a 2-D finite-element method (FEM) magnetostatic model has been developed, in which a uniform Ic distribution is assumed in the HTS stacks. Both the experimental and FEM results are discussed in terms of cable performance, showing that current commercial tapes can already meet the ENEA cable specifications at T = 5 K, B = 15 T. © 2014 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.

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 ReactW –0.30%. After the baseline of the DEMO device was updated in 2015, the new requirements led to an updated conductor design, for 63 kA at 12.2 T. The manufacturing experience of the first prototype, named RW1, is exploited in a second short-length prototype conductor, named RW2, assembled and tested in 2017: The conductor aspect ratio is reduced, and the segregated copper wires are replaced by a solid block of mixed matrix stabilizer. Although designed for the TF coils, with dc operation, the moderate ac loss of the flat cable makes the RW2 a good candidate also for the central solenoid conductor. The Nb3Sn strand for RW2 is supplied by WST (PRC), and the flat cable is made at TRATOS (Italy). The rationale of the design, the conductor manufacture, the sample assembly, and the test results in SULTAN are reported.

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.

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 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 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.

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 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.

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.

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.

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.

Two short lengths of the full-size NbTi cable-in-conduit (CIC) conductor used to fabricate the ITER poloidal field coil insert were tested in the SULTAN facility. The two investigated conductors, otherwise identical, are distinguished only by the presence and absence of the last stage subcable wraps. The voltage-current (V-I) characteristics and the quench behaviour of the conductors with and without subcable wraps are compared. The dc performance of the conductors is limited by the occurrence of sudden take-offs above approximate to 35 kA in the leg with wraps and above approximate to 45 kA in the leg without wraps. Hall probes used to study the current distribution indicate that the better performance of the conductor without subcable wraps is closely related to the lower transverse resistance providing a more efficient current redistribution. (c) 2005 Elsevier B.V. All rights reserved.

An evolution of the contact resistances (R/sub c/) of two full-size ITER (International Thermonuclear Experimental Reactor) cable-in-conduit conductors (CICC) aimed for the Poloidal Field Conductor Insert (PFCI) was studied in the Twente Cryogenic Press. The conductors are distinguished by the presence of stainless steel wraps around the last stage sub-cables in one of the conductors. The R/sub c/ and AC loss were measured on the conductors being in the "virgin" state and after they have been loaded 40,000 times with a peak force of 315 kN/m. The test results are compared with those, obtained on sections from the same conductor lengths, in the SULTAN test facility. The consequences of the cyclic loading and the sub-cable wraps on the R/sub c/ and the coupling AC loss are discussed in view of the previously published results on full- and sub-size ITER NbTi CICCs. Special emphasis was paid to the measurements of the contact resistances between individual sub-cables and the conduit, since the knowledge of them is important for proper interpretation of the DC test results. As a final goal, the test outcome will be used as a reference for calibration of the PFCI AC loss performance.

One of the most important characteristics of Nb/sub 3/Sn strands is the strain dependence of the critical current density which has to be taken into account for every conductor design, especially for cable-in-conduit conductors (CICC) due to the additional influence of the jacket material. Therefore, a task has been launched in the framework of the European Fusion Technology Program to determine the residual thermal strain of single strands and different multi-stage conductors on absolute scale which is essential for the final optimization of the ITER TF conductor. For this purpose, an existing test rig for critical current measurements on single strands and multi-stage CICC as a function of strain and field, the FBI facility at the Forschungszentrum Karlsruhe, has been recently modified and modernized to improve the overall accuracy and resolution of the system. After a description of the facility, its capabilities and the recently performed upgrades, this paper explains the scheduled test program and the strategy to assess the impact of the cable pattern and size on the residual thermal strain. Results on Nb/sub 3/Sn strands jacketed in stainless steel are presented as well.

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.

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.

Taking advantage of the large experience of the ENEA Superconducting Laboratory in the manufacture and characterization of large scale superconducting systems, a project was launched, aimed at using High Temperature Superconductor (HTS) 2G wires for the manufacture of a cable-in-conduit conductor (CICC). In particular, the main aim was the definition of a conductor design fully compatible with existing cabling technologies, to be promptly transferred to an industrial scale production. The considered layout is based on 150 HTS tapes, arranged as five layered structures of twisted tapes wound on a helically slotted core with external round jacket. All manufacturing steps (slotted core production, tape stacking and insertion into the ducts, external wrapping and jacketing) are fulfilled by using equipments and technologies available at TRATOS Cavi S.p.A. These CICCs are intended for operation using forced flow of Helium. A 2D local thermal model has been developed for the optimization of the cooling configuration. This conductor is designed to target 20 kA at 4.2 K and 15 T (or, alternatively, higher temperature, in self-field and LN2 cooling) corresponding to a Je of about 55 A/mm2. The production of a short dummy sample revealed that the exploited industrial production process is very promising for the development of HTS CICC. © 2013 IEEE.

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.

The AC loss measurements of the high temperature superconductor (HTS) cable prototype in the EDIPO test facility motivated detailed investigations of the loss contributions from the tape, strand and cable stages of the HTS fusion conductor design proposed at the Swiss Plasma Center. As an initial step of the task, magnetization tests of soldered stacks of HTS tapes were carried out at temperatures of 5 and 77 K and magnetic fields up to 12 T using the vibrating sample magnetometer technique. The influence on the magnetization loss of the number of tapes, width of the tape, field's orientation and tape's manufacturer is studied experimentally performing both the major and minor magnetization loops with different ramp rates of the applied magnetic field. In order to validate the test results, a numerical model is developed and presented in this work. From the numerical model we also deduced an analytical approach for the magnetization loss in the stacks of tapes with arbitrary number of tapes in the critical state model. Comparison between the measured and estimated magnetization loss of the cable prototypes is reported as well.

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.

Current reconstruction in cable-in-conduit conductors (CICC) cables is a crucial issue to determine cables performance in working conditions, and must be performed using inverse problem approaches as direct measurement is not feasible. The current distribution has been studied for the ITER Poloidal Field Insert Sample (PFIS) conductor using annular arrays of Hall probes placed in three different locations along the sample during the test campaign at the SULTAN facility. The measurement apparatus is also described in the paper, together with the approach to current reconstruction. (c) 2005 Elsevier B.V. All rights reserved.

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 Poloidal Field Conductor Insert (PFCI) of the International Thermonuclear Experimental Reactor (ITER) has been designed in the EU and is being manufactured at Tesla Engineering, UK, in the frame of a Task Agreement with the ITER International Team. Completion of the PFCI is expected at the beginning of 2005. Then, the coil shall be shipped to JAERI Naka, Japan, and inserted into the bore of the ITER Central Solenoid Model Coil, where it should be tested in 2005 to 2006. The PFCI consists of a NbTi dual-channel conductor, almost identical to the ITER PF1 and PF6 design, 45 m long, with a 50 mm thick square stainless steel jacket, wound in a single-layer solenoid. It should carry up to 50 kA in a field of 6 T, and it will be cooled by supercritical He at 4.5 K and 0.6 MPa. An intermediate joint, representative of the ITER PF joints and located at relatively high field, will be an important new item in the test configuration with respect to the previous ITER Insert Coils. The PFCI will be fully instrumented with inductive and resistive heaters, as well as with voltage taps, Hall probes, pick-up coils, temperature sensors, pressure gauges, strain and displacement sensors. The test program will be aimed at DC and pulsed performance assessment of conductor and intermediate joint, AC loss measurement, stability and quench propagation, thermal-hydraulic characterization. Here we give an overview of the preparatory work toward the test, including a review of the coil manufacturing and of the available instrumentation, a discussion of the most likely test program items, and a presentation of the supporting modeling and characterization work performed so far.

A short sample of the NbTi cable-in-conduit conductor (CICC) manufactured for the ITER PF insert coil has been tested in the SULTAN facility at CRPP. The short sample consists of two paired conductor sections, identical except for the sub-cable and outer wraps, which have been removed from one of the sections before jacketing. The test program for conductor and joint includes DC performance, cyclic load and AC loss, with a large number of voltage taps and Hall sensors for current distribution. At high operating current, the DC behavior is well below expectations, with temperature margin lower than specified in the ITER design criteria. The conductor without wraps has higher tolerance to current unbalance. The joint resistance is by far higher than targeted.

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 crucial issue for a superconducting coil in order to be safely used in the magnetic system of a fusion reactor is stability against all foreseen disturbances. To simulate the fusion machine conditions, including off-normal events, e.g. plasma disruptions, the energy deposition has to be spread over a "long length" cable in conduit conductor (CICC) and a background magnetic field is needed. We have therefore designed and built an experiment consisting of an instrumented NbTi test module inserted in a pair of co-axial pulsed copper coils. A 0.6 m diameter superconducting coil provides a background magnetic field up to 3 T. Calibration of the energy inductively coupled between the pulsed coils and the module has been obtained measuring the system temperature increase just after the pulse by means of thermometers positioned along the conductor. Stability vs. operating current I/sub op/ has been examined for different helium temperatures and different background magnetic fields. The finite element code Gandalf for the stability and quenching transients analysis in forced flow cooled superconducting coils has been run to check the matching with the experimental results. (3 refs).

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.

FAST (Fusion Advanced Studies Torus), the Italian proposal of a Satellite Facility to ITER, has been the object of an intense re-design activity aimed at verifying the feasibility of a superconducting solution for its magnet system. One of the most difficult and critical obstacles is represented by the central solenoid (CS) magnet, due to the rapid variation of the current and magnetic field (up to 40 T/s) foreseen so far in the reference scenario. The associated losses induced in the CS magnets should be dissipated by the flow of helium that cools the conductors. Nonethe- less, even the losses depend on the conductor design, therefore, not only the thermal-hydraulics of the Cable-in-Conduit Conduc- tors (CICCs) but also the strand properties, should be considered to identify the best design capable to sustain the transient in the CS magnets, keeping the scenario originally foreseen for the resistive configuration of the machine unaltered. The present paper discusses the electromagnetic and thermal-hydraulic behavior of the CICC design envisaged for the FAST CS magnet in the H-mode scenario, and identifies the possible strategies to make a superconducting solution sustainable.

The ENEA conductor for the EU DEMO Nb3Sn toroidal field (TF) magnets, cooled by supercritical He, features a rectangular cross section with two small pressure relief channels ('holes'), separated from the cable bundle by means of a flat spiral, twisted together with the last cabling stage. A well-instrumented short sample of the ENEA TF conductor has been tested in SULTAN at SPC, Villigen (Switzerland) in 2016, aimed at its thermal-hydraulic characterization, and the test results are presented here. A correlation for the friction factor in the small holes is derived, best fitting the results of a set of computational fluid dynamics simulations. The new correlation (combined with existing correlations for the He friction factor in the bundle region) is shown to allow a proper reproduction of the measured hydraulic characteristic of the conductor. The heat slug propagation tests are used to calibrate the hole-to-bundle heat transfer coefficient in the 4C thermal-hydraulic code and to estimate the characteristic length for the homogenization of the He temperature on the conductor cross section, following a localized thermal perturbation. © 2002-2011 IEEE.

The Toroidal field coils (TFC) of the EUROFusion DEMO reactor call for Nb3Sn conductor with high field and high current. Another major requirement is cost-effectiveness, to keep the ratio of investment to electric power in the same range of the competing energy sources (fission, hydro, coal, gas, etc.). The TFC proposed by the Swiss Plasma Center (SPC) is based on a double-layer Nb3Sn/NbTi winding. A react-and-wind flat cable is the core of the Nb3Sn conductor, with six grades to minimize the cost and maintain a roughly constant temperature margin of 1.5 K over the winding cross section. A short length section of the high grade Nb3Sn conductor has been manufactured using relevant industrial cabling equipment. One hundred kilograms of 1.5-mm Nb3Sn strand has been procured at WST with average Jc up to 15% higher than specified (Jc≥ 1000A/mm2 at 12 T/4.2 K). A dedicated cabling line has been set up at TRATOS cavi (Italy), producing over 350 m of dummy cable and about 13 m of superconducting cable. The assembly of the cable into a conduit by longitudinal laser welding of two steel profiles is demonstrated, including the QA procedures. A test sample has been prepared at SPC by heat treating straight sections of the cable and encasing it into a steel jacket after the heat treatment to minimize the thermal strain. The test was carried out in three test campaigns at the EDIPO facility at SPC. The test program includes Dc performance at the relevant operating conditions. An assessment of the conductor test results in terms of strand performance suggests that the applicable thermal strain is less than -0.33%. The performance is stable upon load cycles. © 2002-2011 IEEE.

The experience gained in the past for the ITER toroidal field model coil conductor and the results obtained so far have led to the definition of an upgraded full size prototype conductor, based on advanced Nb3Sn strand, and entirely manufactured in the European Union (EU). Samples for the characterization in the Sultan facility have been prepared by Luvata (Italy) following the conductor layout defined by ITER. ENEA was responsible for conductor fabrication. Since the conductor layout was new, a full size copper dummy conductor has been preventively produced for the setting of the cabling and jacketing tools. Then, a total of four full size superconducting cables have been prepared by using Nb3Sn advanced strands produced by Oxford Instruments (OST) and European advanced superconductors (EAS), by internal tin and bronze technology, respectively. The details of manufacturing procedures will be described in this paper.