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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.

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 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.

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 performance of ITER toroidal field (TF) conductors still have a significant margin for improvement because the effective strain between -0.62% and -0.95% limits the strands' critical current between 15% and 45% of the maximum achievable. Prototype Nb3Sn cable-in-conduit conductors have been designed, manufactured and tested in the frame of the EUROfusion DEMO activities. In these conductors the effective strain has shown a clear improvement with respect to the ITER conductors, reaching values between -0.55% and -0.28%, resulting in a strand critical current which is two to three times higher than in ITER conductors. In terms of the amount of Nb3Sn strand required for the construction of the DEMO TF magnet system, such improvement may lead to a reduction of at least a factor of two with respect to a similar magnet built with ITER type conductors; a further saving of Nb3Sn is possible if graded conductors/windings are employed. In the best case the DEMO TF magnet could require fewer Nb3Sn strands than the ITER one, despite the larger size of DEMO. Moreover high performance conductors could be operated at higher fields than ITER TF conductors, enabling the construction of low cost, compact, high field tokamaks. © 2018 IOP Publishing Ltd.

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.

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

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

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

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

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 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.

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

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.

The superconducting toroidal field (TF) coils composed of a winding pack (WP) without radial plates, encapsulated in a steel casing, are currently being studied in the framework of the EUROfusion work package magnets (WPMAG). The ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development) 2014 TF WP design consists of graded (Nb3Sn + NbTi), double-layer wound rectangular cable-in-conduit conductors, for which the operational as well as accidental transients must be carefully investigated. This paper presents the analysis of the quench propagation inside the TF WP proposed in 2014 by ENEA using the state-of-the-art thermal-hydraulic code 4C. Different locations are considered for the quench initiation: 1) at the maximum temperature margin $\Delta T_{\mathrm {mar}}^{\mathrm {max}}$ , resulting in the most severe hotspot condition and 2) at the minimum temperature margin $\Delta T_{\mathrm {mar}}^{\mathrm {min}}$ , to analyze the most probable event. The evolution of voltage, normal zone, hotspot temperature, and maximum pressure in the TF WP is computed comparing the results from the two cases, highlighting the role of heat diffusion inside the TF WP.

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

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

In the frame of the European DEMO reactor activities, Work Package MAG, superconducting Toroidal Field (TF) coils composed by a graded (Nb3Sn + NbTi) winding pack (WP) without radial plates, encapsulated in a steel casing, are under analysis. The ENEA WP design consists of double-layer wound rectangular cable-in-conduit conductors (CICC), for which operational as well as accidental transients must be carefully investigated. The paper presents the application of the state-of-The-Art thermal-hydraulic code 4C to the analysis of the quench propagation inside the WP proposed in 2014 by ENEA. The quench is conservatively initiated at the location of the maximum temperature margin and the voltage, normal zone, hot spot temperature and maximum pressure evolutions in the WP are computed, highlighting the role of thermal coupling inside the WP. © 2015 IEEE.

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

The tests analysis of the International Thermonuclear Experimental Reactor (ITER) model coils showed lower performances than expected. For that reason, a specific program has been launched in Europe for investigating the influence of bending strain on the transport properties of high-performances Nb3Sn strands inside cable-in-conduit conductors when experiencing significant Lorentz forces. To this purpose, an experimental campaign has been initiated on single Nb3Sn strands jacketed inside a 0.2 mm thick stainless-steel (SS) tube in order to be tested in mechanical ITER Toroidal Field (TF) Coils relevant conditions. In this article we describe in a first part the procedure investigated at CEA to impose a controlled pure bending strain upon jacketed stands and we present the results of the qualification tests carried out in the ENEA facility to assess the reliability of the chosen method. In a second part we propose a new model to simulate the strands behaviour experiencing bending strain together with an initial compressive strain. To understand our results, a key parameter considered is a specific weighting function relying on strand interfilament current redistribution capacity. At the end, this model enables us to explain more reliably the experimental results but also shows discrepancies which requires further improvements and extra refinements

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.

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.

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.

Bus Bar III (BBIII) is a curvilinear NbTi full-size ITER-type multistrand superconducting cable 7 m long, tested in 2004 at Forschungszentrum Karlsruhe, Germany. The sample was cooled with a forced helium flow and was equipped with many electromagnetic and thermal-hydraulic sensors, to acquire as wide as possible an experimental database, aimed at getting a benchmark for the validation of the THELMA code, a computer tool for the analysis of CICC prototypes and magnets, recently developed and presently under validation. This code can compute the current distribution among the macrostrands used to represent the cable, taking into account their superconducting behavior, as well as the temperature and the magnetic field distribution. The paper deals with the application of THELMA to some tests on BBIII useful for the electromagnetic model validation and presents in detail the analysis setting-up, the results, its critical aspects and limits. A comparison is shown between the measured and the computed waveforms of the electromagnetic data.

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.

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

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

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.

A facility for the characterization of superconducting strands at variable temperatures has been recently upgraded at ENEA. Measurement of transport properties of superconducting strands can be now carried out in the temperature range from 3.5 K to 15 K, with stability less than 10 mK, and in magnetic fields up to 12.5 T, covering a range of critical currents between a few Amperes, up to about 300 A. In this study we present the results of a measurement campaign carried out on NbTi wires, aimed also at comparing the performances of some available candidate strands, to be used in the design of the Toroidal Field Coils of the JT60-SA Tokamak, as well as in the design of the ITER high field Poloidal Field Coils, PF1 & 6. We measured strands with different diameters, and with Cu-nonCu ratios ranging between about 1.5, up to about 7. The measurements showed the critical behavior of the NbTi in the range of relatively high temperatures (T > 5.5 K) and high magnetic fields (B > 6 T), typical for large coils in fusion reactors, where Cable-in-Conduit Conductors are used and where temperature margins of at least 1K are required for a sound design. The experimental results highlighted the need for an optimization of the numerical fits normally used to predict NbTi critical current densities, starting from measurements performed at 4.2 K. Based on our results, we conclude that fit parameters for each NbTi strand should be inferred from experimental characterizations in the temperature and magnetic field range of interest for the design.

In the framework of the DEMOnstration fusion power plant (DEMO) design coordinated by the EUROfusion consortium, a pre-conceptual design of the superconducting magnet system has been developed. For the toroidal field coils (TFCs), three winding pack (WP) options have been proposed; exploring different winding approaches (pancakes vs. layers), and manufacturing techniques (react & wind vs. wind & react Nb3Sn). Thermal-hydraulic and mechanical analyses on the three WPs have produced encouraging results, with some critical issues to be solved in future studies and optimizations. The experimental tests on TF prototype short sample conductors have demonstrated a limited performance degradation with electro-magnetic cycles and significantly lower effective strains than most of the large-size Nb3Sn conductors reported in literature. The toroidal field quench protection circuit has been studied, starting from different topologies and focusing on the most promising one. Two designs are also presented for the central solenoid magnet, with preliminary evaluations on the AC losses during the plasma breakdown. Finally, the design of a TF winding pack based on HTS conductors and the experimental tests on “fusion-relevant” HTS cables are illustrated.

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.

This paper reports the main outcomes of the 2013 design activities for the toroidal field magnet of the European nuclear fusion reactor DEMOnstration Power Plant (DEMO). Within the current project plan, the manufacture of full-size conductor prototypes during 2014 and their test during 2015 are foreseen, aimed at the assessment of the proposed cable-in-conduit-conductor technologies. Industrial partners have been thus already involved at this early stage, and the preliminary cabling trials performed are illustrated here. As far as the coil design assessment is concerned, one of the main aspects to be investigated is the mechanical performance of the magnet, considering its large size and the high electromagnetic forces involved. An overall 3-D finite-element model of the coil has been set up, and mechanical stress analysis has been carried out for a selected instant of the plasma scenario after the computation of the electromagnetic loads within a full 3-D system model, including central solenoid and poloidal field coil contributions. The resulting loads have been mapped on the finite-element-method model, making use of a novel mesh-independent interpolator based on radial basis functions. The results reported here highlight a quite high level of stress, which is close to the acceptance limits, in some regions of the steel casing. Further investigations and assessments are required, but the most critical areas are clearly identified.

Thermo-hydraulic modeling of cable in conduit conductors (CICC) contains one main uncertainty in the heat exchange coefficient used in the different possible fluid regimes. Nowadays validation of codes, by comparison with data from experiments involving long length instrumented conductors, is one of the main technique for assessing the design tools for the large magnets foreseen in the next generation fusion machines. To this purpose, a broad disturbance scenario has to be investigated to confirm the magnet's stability versus normal and off-normal working conditions. In this paper a comparison between the predictions of the code Gandalf and the data obtained from an instrumented NbTi conductor module is carried out.

In this paper, we discuss the predictive analysis performed in support of the test program of the International Thermonuclear Experimental Reactor (ITER) poloidal field conductor insert (PFCI). A subset of the test program items was considered, with particular emphasis on DC performance and AC losses. The results and implications of the comparison of selected predictions from different laboratories will be presented.