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The ITER correction coils (CCs) are designed to be arranged toroidally around the machine inside the poloidal field coils and mounted on the toroidal field coils in order to compensate field errors arising from misalignment of the coils and winding deviations from the nominal shape as a result of fabrication tolerances, joints, leads, busbars CCs, and assembly tolerances. China is responsible for manufacturing all the CCs and their conductor. The paper presents the qualification progress and results of ITER CCs and CC conductor, including the test results of CC SULTAN samples, double pancakes winding experiments, CC case welding techniques, and manufacturing experiment of CC helium inlet/outlets. [ABSTRACT FROM PUBLISHER]

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]

The ITER magnet procurement is now well underway. The magnet systems consist of 4 superconducting coil sets (toroidal field (TF), poloidal field (PF), central solenoid (CS) and correction coils (CC)) which use both NbTi and Nb3Sn-based conductors. The magnets sit at the core of the ITER machine and are tightly integrated with each other and the main vacuum vessel. The total weight of the system is about 10000 t, of which about 500 t are Nb3Sn strands and 250 t, NbTi. The reaction of the magnetic forces is a delicate balance that requires tight control of tolerances and the use of high-strength, fatigue-resistance steel forgings. Integration and support of the coils and their supplies, while maintaining the necessary tolerances and clearance gaps, have been completed in steps, the last being the inclusion of the feeder systems. Twenty-one procurement agreements have now been signed with 6 of the ITER Domestic Agencies for all of the magnets together with the supporting feeder subsystems. All of them except one (for the CS coils) are so-called Build to Print agreements where the IO provides the detailed design including full three-dimensional CAD models. The production of the first components is underway (about 175 t of Nb3Sn strand was finished by July 2011) and manufacturing prototypes of TF coil components are being completed. The paper will present a design overview and the manufacturing status. [ABSTRACT FROM PUBLISHER]

ITER magnets are all designed using superconducting conductors with a high current carrying capability. The Poloidal Field (PF) coils are operated in pulsed mode with maximum operating currents of 48 kA for PF 1 and 6, 55 kA for PF 2,3, and 4 and 52 kA for PF 5. The PF system consists of 6 ring coils, PF1 through PF6, that serve to stabilize the position and control the shape of the plasma in the tokamak. The six PF coils are attached to the outside of the Toroidal Field (TF) coil cases through flexible plates or sliding supports allowing radial displacements. The PF coil positions and sizes have been optimized for the plasma requirements consistent with the constraints imposed by allowing access to the vacuum vessel and by considering pumping ducts to the in-vessel components. [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 DC performance of the so-called Good Joint Nb3Sn Short Sample conductor, measured in the SULTAN facility in 1999 with a progressively reduced (chopped) joint length, became worse as the joint became shorter. In this paper we present the analysis of those tests using the THELMA code. The results confirm quantitatively that the loss of performance was due to increasingly unbalanced current distribution. Once the unbalance due to the joint is properly modeled, the performance of the conductor can be fitted using the same extra longitudinal strain in the critical current scaling, proportional to the electromagnetic load but independent of the joint length. [ABSTRACT FROM AUTHOR]

The effect of applied and thermal strain on the superconducting performance of Nb3Sn strands is well known. Prediction of the strand performance depends on an accurate knowledge of the strand strain state. As well as operating strains due to mechanical loads on the conductor, the strand has a complex system of internal strains due to the fabrication process and partial annealing at the reaction heat treatment temperature (about 600 C). Nb3Sn strands are now available with a range of different matrices around the superconducting filaments, varying both in mechanical properties and geometry. Starting from a database of mechanical properties, this paper will use finite element analysis to model the elasto-plastic behavior of the matrix, predicting the filament multi-axial strain state. [ABSTRACT FROM AUTHOR]

The toroidal Field Conductor Insert (TFCI) coil was tested at 46 kA in the 13 T background magnetic field generated by the Central Solenoid Model Coil. The single layer TFCI coil uses a Nb3Sn cable-in-conduit conductor in a Titanium jacket. The object of this work is to verify the data self-consistency, and to apply a thermal, hydraulic and electrical simulation to derive the strand performance in the cable. We present the results of the analysis in support of the interpretation of a subset of current sharing temperature measurements. After estimation of the n-value in the voltage-current equation, we perform the simulations using the total longitudinal strain as the only fitting parameter; the resulting value is much larger than the value expected from simple thermal contraction considerations. [ABSTRACT FROM AUTHOR]

The tests of the Toroidal Field Model Coil (TFMC) were completed in 2002 in the TOSKA facility of Forschungszentram 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 (Tc) of the TFMC conductor, possibly the highlight of the whole test campaign. These tests were performed by increasing in steps the helium inlet temperature in double pancake DPI, 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 -Tin characteristic by means of the M&M code allows to self-consistently deriving an estimate of Tc 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. [ABSTRACT FROM AUTHOR]

The full set of TCS measurements performed during 2000–2002 on conductor 1A of the ITER Central Solenoid Model Coil (CSMC) of the International Thermonuclear Experimental Reactor (ITER) is analysed with the extensively validated M&M code. Under the assumptions of uniform strand properties and uniform current distribution among strands, the performance of this “average” strand in the cable is deduced from the best fit of the measured voltage–inlet temperature characteristics. Two fitting parameters are chosen: an ad-hoc additional contribution ‘ϵextra’ to the longitudinal strain of the average strand in the cable, and the average-strand (or cable “effective”) index ‘n’ of the electric field–current density characteristic. It is shown that the average strand in the CSMC performed less well than expected from the strand database––an increasingly negative ϵextra being needed to reproduce the coil behaviour at increasing transport currents, and that the cable effective n is clearly below the measured n of the strand. It is argued that this average-strand performance reduction in the CSMC is most likely related to mechanical load effects. [Copyright &y& Elsevier]

Discusses electric conductor design for international tokamak engineering reactor (ITER). Operating conditions; Design criteria and operating limits; Design optimisation; Conclusions.

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]

Abstract: The ITER correction coils (CC) include three sets of six coils each, distributed symmetrically around the tokamak and inserted between the toroidal field (TF) and the poloidal field (PF) coils. Each pair of coils located on opposite sides with respect to the plasma is series connected with polarity such to produce asymmetric fields. These superconducting coils use a cable-in-conduit conductor, insulated, wound into multiple pancakes and inserted inside an austenitic stainless steel case. The requirements and the main features of the design are presented and the selected options reviewed in terms of their criticality in achieving the specified tolerances. The requested qualification trials are identified and reports the results obtained so far. [Copyright &y& Elsevier]

Abstract: When energized, the ITER central solenoid (CS) coils experience large pulsed electromagnetic forces translating into a combination of tensile hoop load in the toroidal direction and transverse vertical compression to be resisted by the conductor jacket itself. A key parameter for the coil design is fatigue behaviour (60000 cycles) of the CS conductor jacket. The mechanical requirements on the CS conductor jacket are derived from the mechanical analyses performed on the CS stack assembly. A prototype production of the two considered candidate materials, 316LN and JK2LB, a specially developed high Mn-bearing austenitic stainless steel, has been achieved in industry. Manufacturing specifications and first results of production controls are presented. [Copyright &y& Elsevier]