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This paper summarizes the work on the manufacturing and testing of conductor samples for ITER poloidal field 1&6 coils. The work was carried out by the IO and multiple institutions in the EU, Switzerland and Russia. Five samples were manufactured and tested in the SULTAN test facility. Analyses of dc performances, ac, and mechanical losses test results have been performed. All measurements were carried out before and after cyclic mechanical load created by the electromagnetic forces. DC test results of all samples tested demonstrated that the conductor performances are not affected by cycling mechanical load and meet the required value of current-sharing temperature T cs ≥ 5.8 K at magnetic field B max = 6.4 T and operating current I op = 48 kA. Measurements of the ac and mechanical losses were performed by gas flow calorimetric method. AC loss measurements were carried out at different operating currents and background magnetic fields. AC loss of tested samples showed no change under cycling mechanical load, but showed a correlation of ac loss versus background magnetic field. Mechanical loss, as had been expected, decreased significantly after first ten cycles then reaches a stabile value. The value of mechanical loss even in virgin condition does not affect workability of a magnet.

Hysteresis losses of five Nb 3 Sn International Thermonuclear Experimental Reactor reference strands were investigated by means of magnetization loop measurements in a vibrating sample magnetometer in a perpendicularly applied magnetic field. The magnetization loops were recorded while continuously sweeping the applied field between the extreme values ±B m , covering a wide range of maximum applied fields (0.2-10 T). In this paper, we compare the directly determined hysteresis losses based on the area of the smaller measured loops and the losses calculated by the integration of the width ΔM of the B m = 10 T magnetization loop. A suitable fitting function is proposed to describe the ΔM(B) dependence, which leads, for each strand, to an excellent agreement with the experimentally determined hysteresis losses, magnetization, and pinning force. Transport critical current measurements in a perpendicularly applied magnetic field were also performed for all the strands, and on the basis of the comparison with the transport measurements, the critical current densities were determined from the magnetization data.

The ITER vertical stabilization coil system is designed to provide fast plasma stabilization in all scenarios with elongated plasma cross-section. It consists of two water-cooled coils located in upper and lower portions of the vacuum vessel (VV), with feedthroughs responsible for the transfer of electricity and cooling water across the vacuum boundary of VV. Depending on location, the feedthroughs can be divided into upper port feedthroughs (UPR feedthroughs) and lower port feedthroughs (LWR feedthroughs), each family sharing the same design. Both designs are subject to demanding operation scheme, a severe working environment and the presence of interfacing components, which impose important thermal, electromagnetic and seismic loads on the components. In particular, the LWR feedthrough, object of present paper, is more impacted by seismic loads due to the more flexible structure compared to the UPR feedthrough. During the worst seismic event, it is subjected to a vertical acceleration of 28.8 m s−2, acting on an effective mass of 249.9 kg at its first order modal mode. Furthermore, the current circulating in the conductor of the LWR feedthrough, impose significant linear forces, up to 27.7 kN m−1 during normal operation and 30.3 kN m−1 during categories I & II abnormal plasma events, respectively. Finally, the effect of differential thermal expansions of interfacing structures imposes a relative displacement between the supports up to 32.9 mm in X (radial) direction and up to 14.67 mm in Z (vertical) direction, during baking. Being part of the vacuum boundary, the feedthrough must assure confinement function under all normal and accidental loading conditions, with proper structural integrity assessment by applying analysis approach and design criteria foreseen by the applicable nuclear codes. The paper details the load types and categories, the analysis methodologies, the structural design criteria, and the assessment results in views of static stress and fatigue for the LWR feedthrough. This study could be a comprehensive reference for the structural integrity assessment of other similar components in fusion devices.

Direct current tests performed in the past on the conductor samples of the toroidal field (TF) ITER coils revealed degradation of current sharing temperature, T cs. The degradation progresses with repetitive electromagnetic loading, and also with thermal cycles between 4.5 K and room temperature. This feature was observed on short samples in SULTAN test facility (EPFL-SPC, Switzerland) as well as in TF Insert Coil tests in CSMC test facility (Naka, Japan). We present three independent observations suggesting that initiation of sample quench followed by a fast current discharge, which normally complements every I c and T cs test in both SULTAN and CSMC, enhances the T cs degradation rate. The exact mechanism of this contribution to the degradation remains unidentified.

Four ITER main busbar (MB) conductor samples were tested in the SULTAN test facility (Centre de Recherches en Physique des Plasma, Ecole Polytechnique Fédérale de Lausanne, Switzerland) between 2011 and 2013. The MB conductors are NbTi-based cable-in-conduit conductors (CICCs), and they will become part of the feeder system of the ITER magnets. The measured dc performance of the four MB samples varied significantly, supposedly depending on the design of the bottom terminations. Two out of three samples with a U-bend box, made of a continuous conductor section, exhibit approximately 0.5 K lower current-sharing temperature Tcs than the sample with a solder-filled bottom joint, consisting of two straight conductor sections. We assess the theoretically expected Tcs performance of the MB conductor based on the characterization of individual NbTi strands, on the ITER NbTi scaling law, and on the magnetic field distribution across the cable cross section. The magnetic field in a SULTAN sample consists of three components, namely, the background SULTAN field, the self-field generated by the current in the conductor under test, and the magnetic field generated by the current in the return conductor. Taking into account all the three components, we calculate the average electric field in the cable and determine Tcs as in the experiment, namely, as the temperature at which the electric field reaches the critical value of Ec=0.1 μV/cm. The theoretically assessed Tcs confirms that the MB sample with solder-filled joint behaves as expected.

Some of the bronze-route Nb3Sn wires produced for the ITER toroidal field magnets have values of the Cu residual resistivity ratio (RRR) lower than the specification (RRR > 100) when reacted with the longest ITER heat treatment cycle (cycle A: high-temperature plateau 200 h at 650 ◦C). As the low RRR value was suspected to be due to the Cr plating, CERN has carried out extensive investigations on two different Cr-plated Nb3Sn wires in order to assess how the RRR is influenced by the Cr plating. Each type of wire was reacted with three heat treatments differing for the duration of the high-temperature plateau (from 100 to 200 h). The presence of Cr in the Cu stabilizer was investigated using a transmission electron microscope together with energy-dispersive X-ray spectrometry. In the most critical wire reacted for 200 h, CrS nanoparticles and Cr in solid solution (0.02 at.%) with Cu were found up to 13 μm from the Cr–Cu interface. At larger distance, we could not quantify the Cr concentrations because of the detector sensitivity limit (about 0.016 at.% Cr in Cu). We provide a model that can explain the measured RRR degradation purely in terms of the Cr contamination that we observed.

Trepetlika

Abstracrt ITER is incorporating two types of In-Vessel Coils (IVCs): ELM Coils to mitigate Edge Localized Modes and VS Coils to provide a reliable Vertical Stabilization of the plasma. Strong coupling with the plasma is required in order that the ELM and VS Coils can meet their performance requirements. Accordingly, the IVCs are mounted on the Vacuum Vessel (VV) inner wall, in close proximity to the plasma, just behind the Blanket Shield Modules (BSM). Due to high radiation environment, mineral insulated copper conductors enclosed in a stainless steel jacket have been selected. The reference design and prototype work provided a good basis for the development of radiation resistant conductor capable of operating within the harsh conditions in ITER vacuum chamber. However, this effort identified shortcomings in achieving satisfactory manufacturing solution, and most significantly, difficulties in brazing the brackets onto the ELM coil conductor. Since this process has not proven successful, alternative designs are under development and prototyping. Prototype manufacturing on the alternative designs has been completed at ICAS, Italy and ASIPP, China. The aim was to eliminate the need for internal coil joints, to prove the principle of longer conductor length manufacturing, and to perform bending and welding trials on two different conductor cross-sections: circular and square. The procurement of the IVCs and their conductors will be done via direct call for tender from the ITER Organization. This paper will give an overview of the alternative design and prototype manufacturing of the ITER In-Vessel coils.

The ITER machine will require approximately 275 tons of Nb-Ti strands that will be used in poloidal field (PF) coils, correction coils (CC) and feeder busbars. The performance of all these conductors for the ITER machine is qualified by a short full-size sample (4 m) current sharing temperature (T cs) test in the SULTAN facility at CRPP in Villigen, Switzerland, at the design operating current and peak field. Three ITER domestic agencies participated in PF conductor fabrication (China, the European Union, Russia) while the conductors for feeder busbars and correction coils are entirely produced by the Chinese domestic agency. Each conductor type was qualified by the ITER International Organization after reaching T cs values in excess of ITER specifications. This qualification enabled the launch of procurement and industrial production of the Nb-Ti cable-in-conduit conductors in each domestic agency. In this paper, we summarize the performance of the qualified Nb-Ti samples of the ITER Project, comparing strand performance with conductor performance. The details of the test results will be discussed in terms of dc performance, ac losses and minimum quench energies of each conductor type.

The ITER toroidal field (TF) strand procurement initiated the largest Nb3Sn superconducting strand production hitherto. The industrial-scale production started in Japan in 2008 and finished in summer 2015. Six ITER partners (so-called Domestic Agencies, or DAs) are in charge of the procurement and involved eight different strand suppliers all over the world, of which four are using the bronze route (BR) process and four the internal-tin (IT) process. In total more than 500 tons have been produced including excess material covering losses during the conductor manufacturing process, in particular the cabling. The procurement is based on a functional specification where the main strand requirements like critical current, hysteresis losses, Cu ratio and residual resistance ratio are specified but not the strand production process or layout. This paper presents the analysis on the data acquired during the quality control (QC) process that was carried out to ensure the same conductor performance requirements are met by the different strand suppliers regardless of strand design. The strand QC is based on 100% billet testing and on applying statistical process control (SPC) limits. Throughout the production, samples adjacent to the strand pieces tested by the suppliers are cross-checked ('verified') by their respective DAs reference labs. The level of verification was lowered from 100% at the beginning of the procurement progressively to approximately 25% during the final phase of production. Based on the complete dataset of the TF strand production, an analysis of the SPC limits of the critical strand parameters is made and the related process capability indices are calculated. In view of the large-scale production and costs, key manufacturing parameters such as billet yield, number of breakages and piece-length distribution are also discussed. The results are compared among all the strand suppliers, focusing on the difference between BR and IT processes. Following the completion of the largest Nb3Sn strand production, our experience gained from monitoring the execution of the QC activities and from auditing the results from the measurements is summarised for future superconducting strand material procurement activities.

ITER cable-in-conduit conductor (CICC) used in the superconducting magnet system consists of a cable made of Nb3Sn or Nb-Ti strands inserted in a stainless steel tube (called jacket or conduit). Depending on the coil type, the jacket material is either made of a low carbon AISI 316LN (for toroidal field coil) and AISI 316L grade stainless steels (for poloidal field and correction coils) or a high Mn austenitic stainless steel developed for ITER called JK2LB (for the central solenoid coil). Mechanical properties of base material and weld joint need to be tested at room and/or cryogenic temperature under predefined mechanical deformation and heat treatment conditions. The amount of required steel requires a tight mechanical properties control to ensure homogeneity during production within the ITER defined requirements. This paper will compare test results such as the ultimate tensile strength (UTS), the yield strength (YS), and the elongation to failure (EL) measured by all ITER jacket supplier test laboratories both for seamless tubes and weld joints.

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

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.

The aim of this paper is to present an up to date review of the status of the European DIPOle (EDIPO) project whose objective is to build a new facility to perform both DC and AC tests of large superconductor samples in high magnetic background field (up to 12.5 T). EDIPO was designed by EFDA, is being built by BNG (Babcock-Noell) and, once completed, it shall be installed near the existing SULTAN facility at the Paul-Scherer-Institute, (CRPP, Villigen, Switzerland). The results presented focus on the manufacture of the dummy coil (which is a 1:1 model of one of the EDIPO poles) that was completed in spring 2009. Details are given about the dummy coil winding results, coil reaction heat treatment, impregnation and acceptance tests. Moreover, a progress status is provided about the manufacturing of the other parts of the EDIPO magnet assembly, in particular with respect to the manufacturing of the test-well, AC coils, yoke sheets cutting and assembly, the outer cylinder production and welding (316 LN steel, 35 mm thick) and the coils ends preparation. Some results regarding the R&D studies are also reported about the winding trials carried out to improve the winding accuracy and the numerical simulations made to support coil winding (e.g. conductor spring back).

The performance degradation under electro-magnetic, transverse load has grown to a key issue for the design of Nb3Sn cable-in-conduit conductors (CICC). Beside the tolerance to bending strain of the basic Nb3Sn strand and the void fraction of the CICC, a relevant parameter is thought to be the cable pattern. A sequence of ldquolongrdquo twist pitches in the early stages of a multi-stage cable is credited to mitigate the performance degradation compared to ldquoshortrdquo pitches. To assess quantitatively the effect of long/short pitches maintaining constant all other conductor parameters, a short length of four stages CICC is prepared, where the first half length has long pitches (83/140/192 mm) in the first three cable stages and the second half length has short pitches (34/95/139 mm). The last stage pitch is 213 mm for both lengths. The cable is made of Cr plated copper and Nb3Sn strands with a diameter of 0.81 mm. The conductor is assembled into a SULTAN hairpin sample where the two branches have respectively long and short pitches. The DC performance, AC loss and pressure drop are measured in both conductor sections and compared to former conductors with the same design. The results are reported and the balance of advantages and drawbacks of long vs. short pitches is discussed.

In the ITER TF Coils reference design the circular conductor is enclosed in the grooves of a stainless steel plate, the so called radial plate. In this paper, we present an alternative solution using a relatively thin walled rectangular conductor without an additional reinforcing structure. We describe the conductor and the layout of the winding pack and analyze the magnetic, thermo-hy-draulic and mechanical behavior of the proposed coil. Then a cost analysis is carried out to estimate the significant saving that can be achieved by the proposed solution.

Following the outcome of the conceptual design phase the EFDA dipole magnet will be made of rectangular cable-in-conduit conductors (CICC) jacketed in 316LN. In order to optimize the required amount of superconductor two different conductor types are used: a high-field (HF) conductor consisting of 144 strands and a low-field (LF) conductor with 108 strands. A high strand with a critical current density (at 4.2 K and 12 T) and an effective filament diameter of was selected. The first series of conductor prototype specimens was tested in summer 2006 but the conductor performances were lower than expected from the pre-prototype tests of 2005 and not fulfilling the design criteria. The conductor layouts were modified to increase the strand support inside the cable and the revised HF conductor design was qualified successfully end of 2006. A current sharing temperature 6 K was found at the dipole operating conditions (12.8 T, 17 kA) confirming the required temperature margin of more than 1 K. The HF conductor qualification process including the design modifications, analysis of the test results and comparison to the expectations are discussed.

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

Recent studies have shown that the presently considered conductor for the poloidal field coil 6 (PF6) of the ITER machine is being designed at the limit of what the considered NbTi superconductor can deliver. Furthermore tests have shown that the performance of conductor prototypes remained under the specified 1.5 K temperature margin. As a result of these issues, ITER is considering a PF6 coil design using Nb3Sn superconductor. We have investigated a PF6 coil design based on Nb3Sn superconductor, which at the same time delivers the required temperature margin as well as the higher performance capability at a similar cost as the reference NbTi design. In addition the design analysis presented here discusses performance parameters such as hot spot temperature and quench propagation. Finally we also present possible coil design and manufacturing procedures.

The ITER superconducting magnet system will require approximately 650 tons of toroidal field and central solenoid Nb3Sn strands with different designs, and more than 250 tons of Nb–Ti strands. This called for a significant scale up of the worldwide production of Nb3Sn and Nb–Ti strands. Over the worldwide and mass productions, it is essential to accurately analyze and characterize the properties of ITER superconducting strands in terms of critical current $(I_{c})$ and ac loss, and thus determine the operational limits of the conductors, ultimately optimizing the operating scenarios of the Tokamak. In this paper, the proper scaling parametrization $I_{c}(B,T)$ and $n$ -value as a function of $I_{c}\ (n(I_{c}))$ are investigated for Nb–Ti strands throughout massive production. Despite the differences in ITER Nb–Ti strand's architecture and their composition, optimized Nb–Ti strand scaling parameterizations throughout production for each supplier have been reached with the minimal deviation ( $B, T$ window and even

In the cable-in-conduit-conductor (CICC) design of the toroidal field system for the international thermonuclear reactor (ITER) Nb3Sn is used as superconductor material. Considering the single strand performance, the crucial characteristic is the strain dependence of the critical current. Within this context, the performance of the CICC under strain is determined by the behaviour of the single strands and additional effects related to the manufacturing process. In the framework of the European fusion technology program a task has been started to investigate single strands as well as sub-size CICC performance using different cable layouts (9, 45 and 180 strands). For this systematic approach, parameters such as the void fraction, the number of pure copper strands, the void fraction or the cabling pattern have been varied. To examine the critical properties in detail, the available test facility, consisting of two experimental setups, is capable to measure the strain dependence in magnetic fields up to 14 T at 4.2 K, by applying an axial load to the samples. Measurements on such sub-size CICC samples are presented and compared to the expected performance.

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

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

In the scope of the design activities for a 12.5 T dipole magnet, a short length of cable-in-conduit conductor (CICC) with steel jacket was prepared in spring 2005 as a pre-prototype of the high field dipole conductor. The small, rectangular conductor uses for the first time "advanced" Nb3Sn strands, with Jc=1100 A/mm2 at 12 T, 4.2 K, also planned in the qualification samples for the ITER conductors. At CRPP, a hairpin sample for SULTAN has been prepared from the short conductor length, heat treated and instrumented. The test program included pressure drop, AC losses and full DC characterization before and after cyclic loading. At large, the test results indicate that the high performance of the advanced strands is preserved in the small size, steel jacketed CICC

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

The latest achievements in the development of high performance Nb3Sn strands are reported. In the field of high energy physics (HEP) critical current densities up to 3000 A mm−2 (at 12 T, 4.2 K and 10 µV m−1) can be achieved. Current actions are aimed at reducing the filament size to about 50 µm and decreasing production costs by using less expensive base materials. In order to take advantage of the progress made by the HEP R&D program, the fusion community has initiated several strand procurement actions worldwide to produce advanced Nb3Sn strand types fulfilling the updated strand specification of the International Thermonuclear Experimental Reactor project. Recently developed internal-tin strands reach critical current densities of more than 1100 A mm−2 and acceptable non-Cu hysteresis losses of about 700 kJ m−3. An important milestone towards higher production rate and yield was reached: a billet of about 100 kg was successfully drawn to the final strand diameter without problems. In powder-in-tube conductors, the composition and number of filaments have been modified to meet the European target specification. Although the bronze technique itself sets certain limitations on Jc, strands with critical current densities around 800 A mm−2 have been produced on an industrial scale.

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 tests of the toroidal field model coil (TFMC) were completed in 2002 in the TOSKA facility of Forschungszentrum Karlsruhe, Germany. Operation reached a combined 80 kA in the TFMC and 16 kA in the LCT coil, resulting in a peak electromechanical load very close to that expected in the full-size ITER TF coils (800 kN/m). Here we concentrate on the measurements of the current sharing temperature T/sub cs/ of the TFMC conductor, possibly the highlight of the whole test campaign. These tests were performed by increasing in steps the helium inlet temperature T/sub in/ in double pancake DP1, resulting in an increasing normal voltage V across the DP1.1 and DP1.2 conductors, and were repeated for several combinations of currents in the TFMC and in the LCT coil. The analysis of the V - T/sub in/ characteristic by means of the M&M code allows to self-consistently deriving an estimate of T/sub cs/, as well as an indirect assessment of the "average" strain state in the conductor. The TFMC isolated strand has also been very recently characterized at different applied uniaxial strain, and preliminary results indicate a stronger reduction of carrying capacity compared to the extrapolation from Summers scaling used in the analysis so far. As a consequence, the performance of the TFMC conductor, as preliminarily re-evaluated here, appears more in line with the strand performance than in previous analysis, although a BI-dependent "degradation" is still present.

The influence of neutron irradiation on the angular, field and temperature dependence of the critical transportcurrent density J c of YBa 2 Cu 3 O 7 d (Y-123) thick films, as well as on the irreversibility fields was investigated. Thespecimens were deposited by liquid phase epitaxy either directly on NdGaO 3 or on seeded (100)MgO substrates with afinal Y-123 layer thickness of 1 and 5–10 lm, respectively. XRD of the superconducting layers proves good in-planeand out-of-plane alignments with FWHMs below 2 . The specimens showed excellent irreversible properties withcritical current densities of up to 3 10 9 A/m 2 at zero field and 77 K and irreversibility fields above 7 T at 77 K for Hkcprior to neutron irradiation, which was performed in fast neutron fluence steps of 10 21 m 2 , corresponding to 5 10 21 defect cascades per m 3 per step. J c enhancements by 100% and an anisotropy reduction by a factor of 2 are achieved atthe optimum neutron fluence. 2003 Elsevier B.V. All rights reserved.

We report on the field and temperature dependence of the critical transport current density Jc, the angular dependence of the transport current at various external magnetic fields and the irreversibility fields in YBa2Cu3O7?? (Y-123) thick films prepared by liquid phase epitaxy (LPE). A comparison of the irreversible properties between specimens produced with and without silver additions to the melt is also presented. Transmission electron microscopy (TEM) was employed to obtain information on the correlation between the transport properties and the microstructure. The samples were deposited either directly on NdGaO3 (NGO) or on seeded (100) MgO substrates, where a 200 nm thin YBCO film deposited by pulsed laser deposition (PLD) acts as seed layer for the LPE process. The final thickness of the Y-123 layer is of the order of 1 ?m for the NGO and between 2 and 10 ?m for the MgO samples. The critical current densities reach 3 ? 109 A m?2 at zero field and 77 K in the best case.

Results on the angular and the field dependence of the critical transport current density as well as on the temperature dependence of the irreversibility fields in YBa/sub 2/Cu/sub 3/O/sub 7-/spl delta// coated conductors are presented. The films were deposited by pulsed laser deposition on (Y/sub 2/O/sub 3//YSZ/CeO/sub 2/) triple buffer structures. Either textured magnetic Ni-Fe alloys or recently developed nonmagnetic Ni-Cr-W alloys with good biaxial texture (8.4/spl deg/ and 9.4/spl deg/ for /spl omega/ and /spl phi/ scans) were used as substrates. The influence of artificial pinning centers, introduced by fast neutron irradiation, on the irreversible properties of the YBCO films is discussed.

High critical current densities (J/sub c/) are required for applications, but grain boundaries suppress the attainable J/sub c/ values. Because of the presence of 90/spl deg/ grain boundaries and because of poor structural and electrical properties, a-axis thin films are of great interest to investigate the role played by Y/sub 0.8/Ca/sub 0.2/Ba/sub 2/Cu/sub 3/O/sub 7-/spl delta// (YCaBCO) cap layers in the transport properties. The transport critical current densities of the films under various magnetic fields and the angular dependence of J/sub c/ at fixed temperature were measured before and after irradiation. A thin YCaBCO cap layer deposited on top of a-axis oriented YBa/sub 2/Cu/sub 3/O/sub 7-/spl delta// (YBCO) films, is found to improve the electrical transport properties. Ca-doping decreases the resistivity and strongly increases J/sub c/ of a-axis oriented YBCO films. Furthermore, strong pinning by 90/spl deg/ grain boundaries was observed.

Coated conductors deposited by a novel high-rate pulsed laser deposition (PLD) technique on ion-beam-assisted-deposition (IBAD) layers of yttria-stabilized zirconia (YSZ) on non-magnetic steel ribbons were investigated. An XRD analysis of the superconducting layers showed in-plane alignments with a FWHM from 10 to 25/spl deg/ and an out-of-plane alignment of better than 2/spl deg/. We report on the field and temperature dependence of the critical transport current density in these YBa/sub 2/Cu/sub 3/O/sub 7-/spl delta// films of various thickness. In addition, the angular dependence of the transport current at various external magnetic fields is presented. Thirdly, artificial pinning centers were introduced into the superconductors by neutron irradiation. The influence of these defects on the critical current densities and on the anisotropy ratio are discussed.