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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 (Tcs) 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 Tcs 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. [ABSTRACT FROM AUTHOR]

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 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]

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]

Abstract: The use of high temperature superconductor (HTS) materials in future fusion machines could increase the efficiency drastically, but strong boundary conditions exist. To outline the prospects, challenges and problems, first the benefit of using HTS materials is estimated considering the saving in cryogenic power. Next, it is demonstrated that industrial available HTS materials can be used for fusion today. For this purpose, we give a short summary of results that have been obtained from an ITER conform 70kA HTS current lead that was designed, built and tested by the Forschungszentrum Karlsruhe and the CRPP Villigen in the frame of the European Fusion Technology Programme and in cooperation with industry. This current lead consists of an HTS part that covered the temperature range from 4.5 to 70K and a conventional part, making the connection to room temperature. Because the HTS part had no ohmic losses and poor thermal conduction, the refrigerator power necessary for cooling the current lead was reduced drastically. The saving factor could be calculated to be 5.4 at zero current and 3.7 at 68kA. The current lead could even be operated at 80kA and with respect to safety criteria of ITER, a complete loss of He flow was simulated showing that the HTS current lead could hold a current of 68kA for 6min without active cooling. These results demonstrate that today existing HTS materials can be used in ITER for current leads or bus bar systems. For fusion machines beyond ITER, the development of an HTS fusion conductor would be the key to operate the complete magnet system at higher temperatures. The option of developing fusion conductors based on Bi-2223 and YBCO are briefly discussed. For a success of such conductors, the AC loss optimisation is crucial. [Copyright &y& Elsevier]

• The designed gap between IVC bolting system to overcome effects from thermal expansion is reasonable. • Thermal and mechanical cycling brings no obvious effects on the IVC insulation layer. • Long time action of EM load may make the fixed bolts loose. ITER In-vessel coils (IVCs) are the components to be installed in the vacuum vessel for plasma positions compensation controlling. Every coil will be wound continuously and clamped through the welded Inconel 625 brackets. The coils are planned to be bolted to the rails which are welded to the vacuum vessel wall behind the blanket shield modules. The operation condition is severe for the IVC coils, combined with high radiation, high temperature, as well as the high thermal and electromagnetic loads. To make sure the safe operation, tests have been done on the mock-up coil to check its fatigue properties under the combined thermal and mechanical loads before the finalization of the design. The integrity of the mock-up coil has been evaluated after the fatigue cycling tests. In this paper, the details are presented on the test procedures and the measurement results are summarized. [ABSTRACT FROM AUTHOR]

The coils inside the ITER vacuum vessel are part of the overall plasma control system that ensures and maintains stable plasma operations. Two vertical stability coils(VS)will provide fast vertical stabilization of the plasma, while an array of 27 edge localized mode (ELM) coils will create resonant magnetic perturbations to magnetically control the plasma exterior in order to suppress potentially harmful power deposition on plasma-facing components. For magnet coils operating inside of the vacuum vessel, conventional insulation schemes are not an option. After a decade of development and intensive collaboration with industry and research institutes, a copper mineral insulated conductor (MIC) in stainless steel jacket will be used as for its ability to withstand large transient electromagnetic forces, high radiation flux, and high temperature. Complex technologies have been developed by ASIPP for the series of production of IVC conductors, in terms of conductor initial dimension design, MgO block assembly and elaboration of Cu pipe insertion, compaction and winding processes. China has finished all qualification phases, especially for 100-meter unit length conductor manufacturing. The first 100-meter unit length of copper tube is successfully manufactured in China. All conductor samples required for qualification process have successfully passed the electrical tests with good performance. The challenges and status of the production of ITER IVC conductors in China are described in this paper. [ABSTRACT FROM AUTHOR]