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A two-components model has been recently developed, for describing the normalized bulk pinning force curves and the critical current density of NbTi strands over a wider B-T range with respect to conventional single-component models. The model was previously successfully applied to data collected on several NbTi commercial strands, with different size, Cu:nonCu ratio, filament diameter and layout, thus confirming the presence of two different pinning mechanisms in conventionally processed NbTi wires. For a further validation, we have extensively tested a strand recently produced by the Chinese Company Western Superconducting Technologies for the ITER Poloidal Field (PF) magnets PF2 to PF5, and applied the model to these data. In order to take into account the observed non-scaling with temperature of the reduced pinning force curves, the model has been updated, including the observed difference in the temperature dependences of the two components contributing to the overall bulk pinning force. The importance of testing wires over very wide temperature ranges is evidenced, and the good agreement between experimental and fit results validate the proposed formulation, which can be regarded as a reliable tool for the description of NbTi performances, to be used in the design of superconducting magnets. From the phenomenological point of view, it is shown that at low temperatures, the two pinning mechanisms contribute to the bulk pinning force, resulting in a pinning force peaking at a reduced field B/Birr \cong 0.5. As the temperature increases, the pinning force peak moves to lower fields, indicating that the low field component pinning mechanism becomes dominant. [ABSTRACT FROM AUTHOR]

FAST (Fusion Advanced Studies Torus), the Italian proposal for a European satellite facility to ITER, is a compact tokamak (R0= 1.82 m, a= 0.64 m, triangularity \delta=0.4) able to investigate non linear dynamics effects of alpha-particle behavior in burning plasmas and to test technical solutions for the first wall/divertor directly relevant for ITER and DEMO (e.g.: full-tungsten wall and divertor and advanced liquid metal divertor). The machine is designed to operate with Deuterium plasmas in a dimensionless parameter range close to that of ITER and to access advanced tokamak regimes with long pulse duration with respect to the current diffusion time. It foresees a maximum magnetic field on plasma axis of 8.5 T and a maximum plasma current of 8 MA. In the present design phase, the feasibility of a superconducting solution for the magnet system is being investigated by ENEA. It consists of 18 Toroidal Field, 6 Poloidal Field and 6 Central Solenoid module coils, all of which wound by \ Nb3{\ {Sn}} and NbTi Cable-In-Conduit Conductors. All the main aspects driving the magnets' design, from mechanical to neutronic and thermal analyses, are here presented and discussed. [ABSTRACT FROM AUTHOR]

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 Nb3Sn 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. [ABSTRACT FROM AUTHOR]

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 Nb3 Sn 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. [ABSTRACT FROM AUTHOR]

Four Nb3Sn conductor lengths were prepared according to the ITER TF conductor design and assembled into two SULTAN samples. The four lengths are not fully identical, with variations of the strand supplier, void fraction and twist pitch. Lower void fractions improve the strand support and increased twist pitches also lower the strand contact pressure but both tend to increase the AC loss and the lower void fraction also increases the pressure drop so that the mass flow rate in the strand bundle area of the cable is reduced. The assembly procedure of the two samples is described including the destructive investigation on a short conductor section to assess a possible perturbation of the cable-to-jacket slippage during the termination preparation. Based on the DC performance and AC loss results from the test in SULTAN, the impact of the void fraction and twist pitch variations is discussed in view of freezing the ITER conductor design and large series manufacture. A comparison with the former generation of conductors, using similar strands but based on the ITER Model Coil layout, is also carried out. The ITER specifications, in terms of current sharing temperature, are fulfilled by both samples, with outstanding results for the conductor with longer twist pitches. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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 Nb3 Sn 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 nΩ. 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. [ABSTRACT FROM AUTHOR]

The effect of transverse loads on Nb3Sn strands has been pointed as a possible cause of the difference observed when scaling transport properties of single strands to those of cable-inconduit 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 Nb3Sn 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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Within the framework of ITER-related projects, new tasks have been recently launched by EFDA CSU Garching (European Fusion Development Agreement Close Support Unit Garching), for the definition and production on industrial scale of advanced Nb3Sn strands, to be used in the manufacturing of the ITER high field CS and TF magnets. We performed an extended characterization of the advanced Nb3Sn strands coming from different European companies, in terms of strand layout (diameter, thickness of Cr coating, Cu:non-Cu ratio), critical transport current, RRR, and hysteresis losses. The results of the measurement campaign show that the upgraded strands meet the latest ITER requirements, with an overall critical transport current of at least 200 A (at 12 T, 4.2 K), equivalent to a non-Cu Jc of about 800 A/mm², a Cu:non-Cu ratio of about 1, a strand diameter of 0.81 mm, and with non-Cu hysteresis losses limited to less than 1000 kJ/m³ on a ±3 T field cycle at 4.2 K. [ABSTRACT FROM AUTHOR]

Bus Bar HI (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. [ABSTRACT FROM AUTHOR]

As the test of the PFCI is foreseen in 2006 at JAERI Naka, Japan, it is essential to consider in detail the lessons learned from the short NbTi sample tests, as well as the issues left open after them, in order to develop a suitable test program of the PFCI aimed at bridging the extrapolation gap between measured strand and future PF coil performance. Here we consider in particular the following issues: 1) the actual possibility to quench the PFCI conductor in the Tcs tests before quenching the intermediate joint, 2) the question of the so-called sudden or premature quench, based on SULTAN sample results, applying a recently developed multi-solid and multi-channel extension of the Mithrandir code to a short sample analysis; 3) the feasibility of the AC losses calorimetry in the PFCI. [ABSTRACT FROM AUTHOR]

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 ∼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. [ABSTRACT FROM AUTHOR]

An evolution of the contact resistances (Rc) 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 Rc 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 Rc 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. [ABSTRACT FROM AUTHOR]

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 toad and AC loss, with a large number of voltage taps and Hail 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. [ABSTRACT FROM AUTHOR]

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, pickup 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. [ABSTRACT FROM AUTHOR]

Abstract: FAST (Fusion Advanced Studies Torus), the Italian proposal of a satellite facility to ITER, is a compact tokamak (R 0 =1.82m, a =0.64m, triangularity δ =0.4) able to investigate non linear dynamics effects of α-particle behavior in burning plasmas and to test technical solutions for the first wall/divertor directly relevant for ITER and DEMO. Currently, ENEA is investigating the feasibility of a superconducting solution for the magnet system. This paper focuses on the analysis of the TF magnets thermal behavior. In particular, utilizing only the room available in the resistive design and referring to one of the most severe scenario envisaged for FAST, the minimum temperature margin in the coil has been calculated for a thermal load distribution on winding and cable jacket due to nuclear heating only. [Copyright &y& Elsevier]