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The inter- and intra-granular critical current densities were determined for different types of coated conductors as well as mono- and multi-filamentary (BiPb)[sub 2]Sr[sub 2]Ca[sub 2]Cu[sub 3]Ox tapes. Either the inter- or the intra-granular critical current density was changed systematically by bending the tapes or by neutron irradiation, respectively. Thus, we are able to assess these two critical current densities, even in samples with very good grain connectivity. [ABSTRACT FROM AUTHOR]

ITER In-Vessel Coils (IVCs), resistive magnets to be installed in close proximity to ITER plasma to compensate fast perturbations of the plasma itself, have undergone a comprehensive revision of their requirements and operational loads, leading to design modifications and R&D activities. An update of the plasma operating scenarios has been done and the maximum currents during transient plasma events have been assessed considering actual operating currents and the surrounding vacuum vessel support structure leading to more representative load cases. The main IVC component modifications are the conductor material and the winding pack support structure. Phase 1 of the conductor procurement involving two suppliers has been completed and the supplier selection for the main production is ongoing. Additional modifications include the joints, where Cu brazing has been replaced by welding, and the conductors are not brazed to the winding pack bracket but clamped. Bracket mockups for the coils have been completed identifying welding parameters and corresponding non-destructive examination (NDE), while different designs of electrical insulating breaks (IB) using Alumina have been tested against pressures up to 80 bars on top of the electrical characterization. A summary of the key challenges experienced towards design finalization is presented. [ABSTRACT FROM AUTHOR]

The ITER IVCs are made of stainless steel mineral-insulated conductors composed by an OFE copper conductor and a SS316LN jacket. The baseline design requires the coils to endure all ITER machine life time. After a brief review of past and recent literatures, no results are available and of our interest. Moreover, cold working, aging treatments as well as the environment temperature affect the mechanical and fatigue properties. The present paper summarizes data from tensile and uniaxial tension load controlled fatigue tests on both the steel jacket and the copper conductor in the IVC representative operation conditions. Tests are performed on specimens sectioned from the tubes undergone cold working of conductor compaction, coil winding and so on as well as heat treatment at 240 °C for 24 h of baking, and finally tested at operation temperature of 120 °C. The fatigue tests were conducted at frequency of 15 Hz and the stress ratio of R = 0.1. Results from tensile tests show both strength and ductility decrease as temperature rises from RT to 120 °C. Results from fatigue tests are presented as S-N relationships with the maximum stress expressed as a percentage of the strength under static loading. It was found that the fatigue strength limit for OFE copper conductor and SS316LN jacket were close to 120 MPa and 275 MPa respectively. The influence from residual stresses on the static and fatigue performance of 316LN jacket were also investigated It was found that the ductility and fatigue life can be improved by residual stress releasing of heat treatment at 650 °C for 100 h. [ABSTRACT FROM AUTHOR]

NbTi strands to be used in four of the six ITER poloidal field (PF) coils, all the correction coils (CC) and all the superconducting feeder busbars are being produced in China. Short full-size qualification conductor (cabled and jacketed) samples have been developed at ASIPP and tested at CRPP. Single pinning mechanism parametrization for this Chinese strand (type S2) has been obtained using the Bottura scaling law. The determination of the scaling parameters using a Kramer-type regression method will be described. A comparison between the critical temperature at the operating current and field of a single strand as determined by the parametrization and the current sharing temperature (TCS) of a few conductor samples tested at the SULTAN facility will be made. The validity and limitation of the estimation will be discussed. The estimated TCS dependence on various (superconducting critical as well as geometric and volumetric) parameters will be assessed using the modelled critical surface. Errors propagated from critical current (Ic) measurements of the strands and parameter fitting, and other uncertainties, will be quantified. [ABSTRACT FROM AUTHOR]

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 Jc. Nb3 Sn strand with a critical current density 2400 A/mm2 (at 4.2 K and 12 T) and an effective filament diameter of 70 μm 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. [ABSTRACT FROM AUTHOR]

Abstract: Strands relevant for fusion with high critical current densities and moderate hysteresis losses were developed and already produced on industrial scale. Based on these achievements EFDA-CSU Garching has launched a Nb3Sn strand development and procurement action inside Europe in order to assess the current status of the Nb3Sn strand production capability. All six addressed companies have replied positively to the strand R&D programme which includes the three major Nb3Sn production techniques namely the bronze, internal-tin and powder-in-tube (PIT) route. According to the strand requirements for the ITER TF conductor a critical current density of 800A/mm2 (at 12T, 4.2K and 10μV/m) and overall strand hysteresis losses below 500kJ/m3 have been specified as the minimum guaranteed strand performance. The second major objective of this programme is to motivate the strand manufacturers to develop and design high performance Nb3Sn strands optimised for the ITER conductor. For this purpose, a target critical current density of 1100A/mm2 has been added to the specification. This paper describes the strategy behind the strand development programme, the actual status of the strand production as well as first preliminary results obtained from the strand suppliers. [Copyright &y& Elsevier]

One of the most important characteristics of Nb3 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 Nb3 Sn sfrands jacketed in stainless steel are presented as well. [ABSTRACT FROM AUTHOR]

The influence of neutron irradiation on the angular, field and temperature dependence of the critical transport current density Jc of YBa2Cu3O7−δ (Y-123) thick films, as well as on the irreversibility fields was investigated. The specimens were deposited by liquid phase epitaxy either directly on NdGaO3 or on seeded (1 0 0)MgO substrates with a final Y-123 layer thickness of 1 and 5–10 μm, respectively. XRD of the superconducting layers proves good in-plane and out-of-plane alignments with FWHM’s below 2°. The specimens showed excellent irreversible properties with critical current densities of up to 3 × 109 A/m2 at zero field and 77 K and irreversibility fields above 7 T at 77 K for H∥c prior to neutron irradiation, which was performed in fast neutron fluence steps of 1021 m−2, corresponding to 5 × 1021 defect cascades per m3 per step. Jc enhancements by 100% and an anisotropy reduction by a factor of 2 are achieved at the optimum neutron fluence. [Copyright &y& Elsevier]

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-δ] 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° and 9.4° for ω and ψ 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. [ABSTRACT FROM AUTHOR]

Studies the transport and anisotropic properties of coated conductors deposited by pulsed laser deposition (PLD) on steel substrates. Field and angular dependence of the critical transport current densities; Influence of neutron irradiation on the change in field dependence of critical current densities.

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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

The influence of the expected Lorentz loading and time dependent operating conditions of a magnet on the conductor AC loss is experimentally simulated by a cryogenic cable press that applies cyclic mechanical loading. A series of ITER conductor tests with the press have commenced and we report on the results from the first set of two TF conductors, which have the option-II cabling scheme but consist of Nb3Sn strands from different manufacturers. With the press, we apply a transverse load of 578 kN/m and the load cycle is repeated up to 30,000 times. As a function of load cycles, we measure the cable mechanical stiffness, interstrand contact resistances, and the coupling loss. When compared with a previously measured option-II type conductor, the present conductors have higher initial losses. However, they showed greater cable displacement and larger increase in contact resistance with load cycles. This is due to the lower cable stiffness thought to be related to the lower axial strand stiffness, resulting in greater cable displacement than the previous cable. Consequently, the two conductors tested here have lower losses already within the first few cycles. [ABSTRACT FROM PUBLISHER]

The ITER in-vessel Vertical Stabilization (VS) coils, composed by the upper and lower VS coils, are designed to provide fast vertical stabilization of the plasma through a fast-current noise profile to compensate small plasma oscillations and to recover vertical displacement events. The VS coils are made by 4 continuous turns of Mineral Insulated Conductor (MIC) joined by Inconel ® 625 welded brackets, installed on rails on the inner wall of ITER Vacuum Vessel (VV). Thus, the coil turns winding/forming and the entire coils assembly have to be performed inside ITER VV, forcing the assembly procedure and tooling design to cope with the very limited space and strict constraints of ITER VV environment. The proposed procedure foresees that MIC conductor is un-spooled in the Neutral Beam (NB) port area and inserted into the ITER VV through a NB equatorial port, then a set of tooling is installed on a reinforced staging platform to wind, form and seal the conductor turns and finally to weld brackets under pre-compression. For this purpose, the tooling design including the reinforcement staging has been developed to meet the required tooling function, configuration optimization, assembly procedure and environmental constraints. The dedicated tools to be used for the assembly procedure of the VS coils are categorized into ex-vessel tools for unspooling and cleaning, in-vessel tools for winding and bump forming, bracket assembly tools for pre-compression and welding, termination and sealing tools, ancillary lifting tools and reinforcement staging platforms. This paper provides the overall assembly procedure and tooling development for the in-situ winding of VS coils. [ABSTRACT FROM AUTHOR]

The ITER magnet system uses cable-in-conduit conductor (CICC) technology with individual strands twisted in several stages resulting in a rope-type cable, which is inserted into a stainless steel conduit. The combination of high current (up to 68 kA) and background magnetic field (up to 13 T) results in large transverse Lorentz forces exerted on the conductors during magnet system operation. The high transverse forces, accompanied with the cyclic nature of the load, have a strong influence on the conductor properties. The Twente Cryogenic Cable Press is used to simulate the effect of the Lorentz forces on a conductor comparable to the ITER magnet operating conditions. An overview is presented of the AC coupling and hysteresis loss, mechanical deformation characteristics and inter-strand contact resistance measurement results obtained on full-size ITER CICCs measured in the Twente Cryogenic Cable Press. The aim of this work is to characterize conductors’ electromagnetic and mechanical properties during cycling of the load up to 30 000 cycles. The evolution of the magnetization (AC coupling loss time constant nτ), mechanical properties and inter-strand resistance Rc between selected strands is presented along with loading history. The Rc between first triplet strands is also measured as a function of applied load. It is shown that transverse load cycling has a strong influence on the CICC properties. An overview of the results for eight toroidal field conductors, two central solenoid conductors, three poloidal field conductors of different types (PF1&6, PF4, PF5), one main bus-bar and one correction coil conductor is presented. [ABSTRACT FROM AUTHOR]

• Feeders are the life lines of the ITER magnet system. • The PF4 CFT is the first ITER magnet component to be installed in the tokamak pit. • The Manufacture Inspection Plan (MIP) guarantees a good proficiency of the process of the Receiving, Inspection and Tests (RIT) phase. • Preliminary checks or trials are fundamentals avoiding potential future problems and saving time in the pit in the tokamak building. The ITER Magnets are in the procurement and assembly phases. During these phases, critical components need to be tested and assembly procedures need to be developed and qualified on mockups. To this end, ITER Organization (IO) and the Commissariat à l'Energie Atomique (CEA) have built up a support structure with space, expertise and equipment: MIFI – Magnet Infrastructure Facilities for ITER. In this framework, IO and CEA perform Receiving Inspection and Tests (RIT) on components delivered to IO by the Domestic Agencies (DA). RIT is a critical stage in the life process of components. These tests are meant to ensure the good quality of components and occur between the Factory Acceptance Test (performed before delivery to IO) and the Site Acceptance Tests that are performed after installation in the tokamak pit. The paper explains the RIT sequence managed on the Poloidal Field #4 Cryostat Feedthrough (PF4 CFT), the first ITER Magnets component to be delivered by a Domestic Agency to ITER Organization, and details leak tests, metrology tests and High Voltage tests performed. [ABSTRACT FROM AUTHOR]

The superconducting coils of the ITER magnet system have hundreds of electrical lap joints interconnecting superconducting cables. The joints operate in a magnetic field of up to 4 T, field derivatives of 0.5 T/s, and currents up to 70 kA. The acceptance tests for the toroidal field (TF), poloidal field (PF), and correction coil (CC) coils will be performed at 77 K, before they are assembled in the pit. Hence there will be no possibility to measure the resistance of the joints in the superconducting state before the whole magnet system is enclosed in the Tokamak cryostat. In addition, no reliable nondestructivemethod has been found to spot the joints with a failure at room temperature. Therefore, the production of the joints relies on the strict adhesion to established robust manufacturing procedures during the qualification phase. As additional quality monitoring, a periodic test of the joint samples manufactured in parallel with a coil fabrication is foreseen to control the reproducibility of the joint electrical performance. In order to qualify the manufacturing procedures, to establish a series production tools and worker teams, a comprehensive qualification program has been set up for manufacturers of the coils in Russia (Poloidal Coil 1, PF1), China (PF6, feeders, CC), Japan (TF), Europe (TF and PF), and USA (Central Solenoid, CS). This program includes a set of mockups manufactured according to the process to be used for the coils and submitted to different tests. They include mechanical testing of materials, electrical tests of full size joint samples, destructive microscopic examination of the joint mockups, and mechanical testing of the full size joint mockups. All tests are carried out in specialized laboratories qualified for this type of work. This paper describes the main items of the qualification program, the tests performed, and the acceptance criteria. The test results are reported and compared to the criteria. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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_{{\rm{cs}}}\geq $ 5.8 K at magnetic field $B_{{{\max}}}\,= $ 6.4 T and operating current $I_{{\rm{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. [ABSTRACT FROM PUBLISHER]

Novel eddy current probes were developed to detect sub-millimetre defects with any orientation on the inner surface of pipes. Five different probes were designed, produced and experimentally validated. These probes include arrays of planar trapezoidal coils in a flexible substrate used alone or together with different winded drive coils. Numerical simulations with Finite Element Method were used to predict the probe response to defects with any orientation. Experimental results in austenitic steel jackets used in ITER revealed that the new probes have an improved reliability compared to conventional toroidal bobbin probes, allowing a higher sensitivity to circumferential defects. [ABSTRACT FROM AUTHOR]

Taking the relay of the large Hadron collider (LHC) at CERN, ITER has become the largest project in applied superconductivity. In addition to its technical complexity, ITER is also a management challenge as it relies on an unprecedented collaboration of seven partners, representing more than half of the world population, who provide 90% of the components as in-kind contributions. The ITER magnet system is one of the most sophisticated superconducting magnet systems ever designed, with an enormous stored energy of 51 GJ. It involves six of the ITER partners. The coils are wound from cable-in-conduit conductors (CICCs) made up of superconducting and copper strands assembled into a multistage cable, inserted into a conduit of butt-welded austenitic steel tubes. The conductors for the toroidal field (TF) and central solenoid (CS) coils require about 600 t of Nb3Sn strands while the poloidal field (PF) and correction coil (CC) and busbar conductors need around 275 t of Nb–Ti strands. The required amount of Nb3Sn strands far exceeds pre-existing industrial capacity and has called for a significant worldwide production scale up. The TF conductors are the first ITER components to be mass produced and are more than 50% complete. During its life time, the CS coil will have to sustain several tens of thousands of electromagnetic (EM) cycles to high current and field conditions, way beyond anything a large Nb3Sn coil has ever experienced. Following a comprehensive R&D program, a technical solution has been found for the CS conductor, which ensures stable performance versus EM and thermal cycling. Productions of PF, CC and busbar conductors are also underway. After an introduction to the ITER project and magnet system, we describe the ITER conductor procurements and the quality assurance/quality control programs that have been implemented to ensure production uniformity across numerous suppliers. Then, we provide examples of technical challenges that have been encountered and we present the status of ITER conductor production worldwide. [ABSTRACT FROM AUTHOR]

Construction of the ITER magnet systems has been started at the end of 2007 following the signature of the first procurement arrangements (PA) for the toroidal field (TF) conductors. Six ITER members are involved in the share of the ITER magnet components and, to date, eighteen PA between the ITER Organization and six domestic agencies have been signed. Substantial progress towards full-scale construction has been achieved with the placement of the first large manufacturing contracts, the production of several tens of tons of advanced Nb3Sn and NbTi strand, and the set-up of large cabling and jacketing facilities. The detailed design of the coils and support structures has also been finalized. The qualification of the fabrication processes for the TF coils and poloidal field (PF) coils has been initiated. The detailed design of the central solenoid (CS) coils is being developed. The design of the correction coils (CCs) with their support structures has been finalized, as well as for the TF gravity supports and clamps of the PF coils. The manufacture of prototypes of the feeder lines and current leads has been started, while ITER is in charge of the procurement of the required magnet instrumentation. This paper provides a progress report on the ITER magnet construction as per December 2010. [ABSTRACT FROM AUTHOR]

The differences in thermal contraction of the composite materials in a cable in conduit conductor (CICC) for the International Thermonuclear Experimental Reactor (ITER), in combination with electromagnetic charging, cause axial, transverse contact and bending strains in the Nb3Sn filaments. These local loads cause distributed strain alterations, reducing the superconducting transport properties. The sensitivity of ITER strands to different strain loads is experimentally explored with dedicated probes. The starting point of the characterization is measurement of the critical current under axial compressive and tensile strain, determining the strain sensitivity and the irreversibility limit corresponding to the initiation of cracks in the Nb3Sn filaments for axial strain. The influence of spatial periodic bending and contact load is evaluated by using a wavelength of 5 mm. The strand axial tensile stress-strain characteristic is measured for comparison of the axial stiffness of the strands. Cyclic loading is applied for transverse loads following the evolution of the critical current, n-value and deformation. This involves a component representing a permanent (plastic) change and as well as a factor revealing reversible (elastic) behavior as a function of the applied load. The experimental results enable discrimination in performance reduction per specific load type and per strand type, which is in general different for each manufacturer involved. Metallographic filament fracture studies are compared to electromagnetic and mechanical load test results. A detailed multifilament strand model is applied to analyze the quantitative impact of strain sensitivity, intrastrand resistances and filament crack density on the performance reduction of strands and full-size ITER CICCs. Although a full-size conductor test is used for qualification of a strand manufacturer, the results presented here are part of the ITER strand verification program. In this paper, we present an overview of the results and comparisons. [ABSTRACT FROM AUTHOR]

The performance of the toroidal field (TF) magnet conductors for the ITER machine are qualified by a short full-size sample (4 m) current sharing temperature (Tcs) test in the SULTAN facility at CRPP in Villigen, Switzerland, using the operating current of 68 kA and the design peak field of 11.8 T. Several samples, including at least one from each of the six ITER Domestic Agencies participating in TF conductor fabrication (China, European Union, Japan, Russia, South Korea and the United States), have been qualified by the ITER Organization after achieving Tcs values of 6.0-6.9 K, after 700-1000 electromagnetic cycles. These Tcs values exceed the ITER specification and enabled the industrial production of these long-lead items for the ITER tokamak to begin in each Domestic Agency. Some of these samples did not pass the qualification test. In this paper, we summarize the performance of the qualified samples, analyze the effect of strand performance on conductor performance, and discuss the details of the test results. [ABSTRACT FROM AUTHOR]

We analyzed the ITER TFEU5 cable-in-conduit conductor (CICC) after the full SULTAN conductor qualification test in order to explore whether Lorentz force induced strand movement inside the CICC produces any fracture of the brittle Nb3Sn filaments. Metallographic image analysis was used to quantify the change in void fraction of each sub-cable (petal); strands move in the direction of the Lorentz force, increasing the void space on the low force side of the CICC and producing a densification on the high force side. Adjacent strand counting shows that local increases in void space result in lower local strand-strand support. Extensive metallographic sampling unambiguously confirms that Nb3Sn filament fracture occurred in the TFEU5 CICC, but the filament fracture was highly localized to strand sections with high local curvature (likely produced during cabling, where strands are pivoted around each other). More than 95% of the straighter strand sections were free of filament cracks, while less than 60% of the bent strand sections were crack free. The high concentration of filament fractures on the tensile side of the strand-strand pivot points indicates that these pivot points are responsible for the vast majority of filament fracture. Much lower crack densities were observed in CICC sections extracted from a lower, gradient-field region of the SULTAN-tested cable. We conclude that localized filament fracture is induced by high Lorentz forces during SULTAN testing of this prototype toroidal field CICC and that the strand sections with the most damage are located at the petal corners of the high field zone. [ABSTRACT FROM AUTHOR]

The ITER magnet coils are wound from Cable-In-Conduit Conductors (CICC) made up of superconducting and copper strands assembled into a multistage, rope-type cable inserted into a conduit of butt-welded austenitic steel tubes. The conductors for the Toroidal Field (TF) and Central Solenoid (CS) coils require about 500 tons of Nb3Sn strands while the Poloidal Field (PF) and Correction Coil (CC) conductors need around 250 tons of Nb-Ti strands. The required amount of Nb3Sn strands far exceeds pre-existing industrial capacity and calls for a significant worldwide production scale up. [ABSTRACT FROM PUBLISHER]

Superconductors for the ITER Poloidal Field Coils are large cable-in-conduit conductors (CICC) made of NbTi strands encased in a round-in-square stainless steel jacket. Three prototype conductor sections for poloidal field coils PF1/6, PF2/3/4 and PF5 have been fabricated in collaboration of the domestic RF, CN and EU agencies and tested in SULTAN Test Facility at the nominal operating conditions. [ABSTRACT FROM PUBLISHER]

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. [ABSTRACT FROM AUTHOR]

In the scope of the qualification tests of the ITER TF conductors, four samples for the SULTAN facility have been prepared, each consisting of two conductor sections manufactured by the US ITER Domestic Agency. The samples are nicknamed USTF1, USTF2, USTF3 and USTF4. The conductor sections used for USTF1 and USTF3, as well as those for USTF2 and USTF4, are identical. The conductor terminations for the samples USTF1 and USTF4 are prepared by the "solder filling" method, where a copper sleeve is swaged on the dismantled conductor before the heat treatment, and the cable space is filled with a soldering alloy by siphon after the heat treatment. In USTF2 and USTF3, the conductor terminations are dismantled after the heat treatment, and the cable ends are dipped into a bath of molten solder. The main test of the conductor characterization in SULTAN is the current sharing temperature (Tcs) test at 10.78 T background field. The current is raised up to 68 kA current in steps of 10 kA. Then, the coolant temperature is raised in steps of 0.25 K till quench occurs. Early voltage is observed during the current ramp before heating even at relatively low current. Due to the broad voltage-temperature (V-T) transition, the early voltage causes an uncertainty in the assessment of the conductor performance. The comparison of the test result of USTF1 vs. USTF3 and USTF2 vs. USTF4 enables to clarify the impact of the termination preparation procedure on the early voltage. In the case the early voltage is caused by non-uniform current distribution and high inter-strand resistance at the joints, the V-T behavior should be clearly different. In this paper, the joint preparation procedures are described, and the impact of the joint preparation on the conductor performance is discussed. [ABSTRACT FROM AUTHOR]

The world-wide procurement of Nb3Sn and NbTi for the ITER superconducting magnet systems will involve eight to ten strand suppliers from six Domestic Agencies (DAs) on three continents. To ensure accurate and consistent measurement of the physical and superconducting properties of the composite strand, a strand test facility benchmarking effort was initiated in August 2008. The objectives of this effort are to assess and improve the superconducting strand test and sample preparation technologies at each DA and supplier, in preparation for the more than ten thousand samples that will be tested during ITER procurement. The present benchmarking includes tests for critical current (Ic), n-index, hysteresis loss (Qhys), residual resistivity ratio (RRR), strand diameter, Cu fraction, twist pitch, twist direction, and metal plating thickness (Cr or Ni). Nineteen participants from six parties (China, EU, Japan, South Korea, Russia, and the United States) have participated in the benchmarking. This round, conducted with a bronze-route Nb3Sn strand, involved samples prepared by a common laboratory (CERN) and sent out to the participants ("IO-prepared samples") and also samples prepared by each individual participant ("self-prepared samples"). Ic samples prepared and measured by CERN were found to have an average Ic of 188.7 A and a standard deviation of 1.8 A (1.0%), while those same set of samples measured by all the participating labs in round-robin fashion were found to have a standard deviation of 2.1 A. Self-prepared samples had an average Ic of 188.1 A, and showed a lab-to-lab standard deviation of 3.5 A. The results demonstrate significant progress in the world-wide capability to accurately and reproducibly measure Nb3Sn critical current over the past decade. Future benchmarking efforts will include an annual cross-check of supplier and DA facilities, and also a round of internal tin Nb3Sn samples to assess each contributor's sample-preparation techniques. A separate round of NbTi benchmarking is also envisioned. [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]

Abstract: Fusion for Energy (F4E) will manage the in-kind contribution from Europe (EU) to ITER of about one quarter of the machine magnet components. This contribution will consist in 10 Toroidal Field (TF) coils, 5 (out of 6) Poloidal Field (PF) coils, 20% of the Nb3Sn conductor used in the TF coils and 15% of the PF coil NbTi conductor. In addition, 12 pre-compression fibreglass rings for the TF system will be procured. This paper presents the complex organization and status of the procurement activities of the F4E magnet components. The technical and managerial challenges, in terms of fabrication, quality assurance and schedule, are also highlighted. [Copyright &y& Elsevier]

The aim of this paper is to present an up to date review of the current status on the superconducting dipole magnet EDIPO, scheduled to be operational at CRPP-PSI (Switzerland) in mid-2010. This saddle-shaped magnet is being built in Europe by a close collaboration among. EURATOM, European Associations (CRPP, ENEA) and European Industries (BNG, LUVATA). Detailed analyses (3D electromagnetic, stress-analysis, thermo-hydraulic, etc.) of the coil confirms the soundness of the design chosen for its projected performances (12.5 T in a bore of 100 mm x 150 mm over a length of ∼1.5 m). The conductors and inter-layer joints have been qualified. The winding line has been set up at BNG and the first dummy double layers are being produced. Impregnation trials and shear tests have been made (and numerically simulated) at room as well as at operating temperature. The installation work at CRPP proceeds with the procurement of the main components (cryostat, transformer, etc.) and the detailed design of the main interface installations (support platform, cryolines etc.). [ABSTRACT FROM AUTHOR]

In this paper we report the main test results obtained on the Poloidal Field Conductor Insert coil (PFI) for the International Thermonuclear Experimental Reactor (ITER), built jointly by the EU and RF ITER parties, recently installed and tested in the CS Model Coil facility, at JAEA-Naka. During the test we (a) verified the DC and AC operating margin of the NbTi Cable-in-Conduit Conductor in conditions representative of the operation of the ITER PF coils, (b) measured the intermediate conductor joint resistance, margin and loss, and (c) measured the AC loss of the conductor and its changes once subjected to a significant number of Lorentz force cycles. We compare the results obtained to expectations from strand and cable characterization, which were studied extensively earlier. We finally discuss the implications for the ITER PF system. [ABSTRACT FROM AUTHOR]

The performance degradation under electro-magnetic, transverse load has grown to a key issue for the design of Nb3 Sn cable-in-conduit conductors (dcc). Beside the tolerance to bending strain of the basic Nb3 Sn strand and the void fraction of the CICC, a relevant parameter is thought to be the cable pattern. A sequence of "long" twist pitches in the early stages of a multi-stage cable is credited to mitigate the performance degradation compared to "short" 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 Nb3 Sn 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. [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]

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 x 150 mm rectangular bore over a length of about 1.5 m in order to test full size conductor samples that ,hall be produced during the ITER magnets procurement. Th 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 th& key results of the on-going supporting R&D. [ABSTRACT FROM AUTHOR]

ISL 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-hydraulic 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. [ABSTRACT FROM AUTHOR]

Abstract-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 Nb3 Sn superconductor. We have investigated a PF6 coil design based on Nb3 Sn 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. [ABSTRACT FROM AUTHOR]

A new generation of ITER TF conductor samples has been assembled and tested in SULTAN in 2007 following a common procedure agreed among the ITER parties. The test results of six SULTAN samples, made of twelve conductor sections manufactured in Europe, Japan, Korea and Russia, are reported here. The conductor layout reflects the ITER TF conductor design, with minor differences for the Nb3 Sn strand characteristics, void fraction and twist pitch. The object of the test is a straight comparison with the ITER requirement of 5.7 K current sharing temperature at 68 kA current and 11.3 T field. A broad range of behavior is observed. [ABSTRACT FROM AUTHOR]

Abstract: In the framework of the European Fusion Programme, the Forschungszentrum Karlsruhe and the Centre de Recherches en Physique des Plasmas (CRPP) have successfully developed and tested a 70kA high-temperature superconductor (HTS) current lead demonstrator. In the present work, the optimisation of the cooling conditions with respect to the cooling power consumption and the required amount of HTS material is described for a binary current lead consisting of a conduction-cooled HTS part and a gas-cooled heat exchanger. The refrigerator input power, required to cool a binary HTS current lead, is determined by the heat load at the 4.5K level, the helium mass flow rate necessary to cool the heat exchanger part and the helium inlet temperature (T He,in). For a range of copper (T Cu) and helium inlet (T He,in) temperatures at the cold end of the heat exchanger, the optimum values of the length of the heat exchanger and the required helium mass flow rate have been calculated. Furthermore, the main results of the pulsed operation of the 70kA HTS current lead demonstrator are presented. The possibility of pulsed operation of HTS current leads for the CS and PF coils is discussed. [Copyright &y& Elsevier]

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]

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. [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]

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]

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. [ABSTRACT FROM AUTHOR]

For the superconducting magnet system of the International Thermonuclear Experimental Reactor, ITER, 60 current leads for a total current of more than 2500 kA are needed. To reduce the resultant large refrigerator load at 4.5 K, High Temperature Superconductor current leads (HTS-CL) could be used. Therefore, EFDA CSU Garching had launched a development program for a 70 kA HTS-CL demonstrator. The Forschungszentrum Karlsruhe and CRPP developed and built this CL optimized for 50 K Helium operation. In 2004, the CL was successfully tested in the TOSKA facility at the Forschungszentrum Karlsruhe. The very encouraging results lead to testing this CL with 80 K Helium because ITER provides a large 80 K Helium cooling capacity for the thermal shields. At the end of last year, the test could be successfully performed demonstrating that high current capacity current leads can be stably operated at about 80–85 K. Recently, the CL was retested using liquid nitrogen which would be an interesting alternative option. In this paper, the test results for the 80 K He cooling operation as well as for the operation with LN2 are presented and compared to the results obtained before for nominal conditions. [ABSTRACT FROM AUTHOR]