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Now that ITER is entering construction, many of its systems are in the final stages of design and analysis. Among them the 31 feeders, which will be supplied in-kind by the Chinese ITER partner. The feeders supply the electrical power and cryogens through the warm-cold barrier to the ITER superconducting magnet systems. They are complex systems with their independent cryostats and thermal shields, densely packed with many components, such as the current feeds, the cryogenic valves and High-Voltage (HV) instrumentation hardware. Some of the feeder components are particularly critical and have been designed with great care. Among them the High-Temperature Superconductor (HTS) current leads, designed for unprecedented currents, the 30 kV class, Paschen-hard HV insulation and the bus bar support system, designed to react the multi-ton Lorentz-forces from the bus bars at minimal heat load. This paper discusses the design challenges for these (and other) key components of the ITER magnet feeders. [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]

The Correction Coils (CC) of the ITER Tokamak are developed to reduce the range of magnetic error fields created by imperfections in the location and geometry of the other coils used to confine, heat, and shape the plasma. The proposed system consists of three sets of 6 coils each, located at the top (TCC), side (SCC) and bottom (BCC) of the Tokamak device and using a NbTi cable-in-conduit superconducting conductor (CICC). Within each set, the coils are connected in pairs to produce a toroidal field to reduce the most troublesome, lower order, poloidal mode number fields (m = 1,2,3) in order to operate below the locked mode threshold. The conductor is designed to operate up to 6 T. The winding uses pancakes of one-in-hand conductor (quadpancakes for SCC, octopancakes for TCC and BCC), thus avoiding internal joints. The winding-pack is enclosed inside a 20 mm thick stainless steel casing. The coils are supported by rigid connections to the Toroidal Field (TF) coils. The structural design of the CC is mainly driven by the allowable fatigue stress levels in the conductor jacket, in the case material and in the glass-polyimide electrical insulation system. The boundary conditions on the CC are imposed by the TF coils deformation and the electromagnetic interactions with the PF coils system. The thermo-hydraulic and electrical performance of the CICC is also addressed. [ABSTRACT FROM AUTHOR]

The Central Solenoid (CS) of the ITER tokamak has to provide the flux variation needed to induce the plasma current and to shape the field lines in the divertor region. It is designed as a stack of 6 identical coils, independently power supplied. Repulsing forces arising between the coils during a scenario are withstood by a precompression structure installed around the coils. Studies were carried out to simplify the winding manufacture, to optimize the precompression structure and procedure, to optimize the stack assembly of the 6 coils and the assembly of the central solenoid inside the tokamak which allows withdrawal from the machine, while meeting the ITER design criteria and in particular the Magnet Structural Design Criteria (static and fatigue). [ABSTRACT FROM AUTHOR]

High Temperature Superconductor (HTS) current leads provide the transition of the current transfer system from room temperature to the 4.5 K. For the ITER superconducting magnet system, a total of 60 HTS current leads are required, with a total nominal current capacity of 2.64 MA. By the use of HTS leads significant power savings can be achieved during the operation of the ITER magnet system. Based on the functional specification by the ITER Organization (IO), ASIPP manufactured and tested a first pair of trial HTS current leads for the ITER coils. The cryogenic test of these 68 kA and 52 kA trial HTS leads was performed at ASIPP using the cryo-plant and power supplies of the EAST Tokamak. The following paper presents the results of the tests and concludes with a discussion of the design features of the next generation prototype of the ITER HTS leads based on the results of the trial lead tests. [ABSTRACT FROM AUTHOR]

The ITER TF Coils will consist of two main components; the WP which is formed by the assembly of seven double pancakes, superconducting windings inserted into stainless steel radial plates and the stainless steel structural casing which forms the mechanical interface between TF coils and the remainder of the ITER machine. The final step in assembly of the TF Coils will be the insertion of the WP into the structural casing. The insertion procedure must achieve two critical objectives; accurate geometric location, +/-1 mm, of the WP current center line with respect to the casing reference faces and the full mechanical location of the WP within the casing to facilitate the uniform transfer of the electromagnetic forces to the casing and the ITER TF keystone structure. [ABSTRACT FROM AUTHOR]

The magnet system for ITER comprises 18 Toroidal Field (TF) Coils using Nb3Sn cable-in-conduit superconductor, which operate at 4.5 K in supercritical helium. The procurement of the TF Coils and Structures is amongst the first which have been launched following the creation of the ITER Organization (IO). It is organized in 4 phases. A Procurement Design Readiness Review held in April 2008 confirmed the readiness of the design to proceed with Phases I and II. Procurement Arrangements (PA) were signed with the European and Japanese Domestic Agencies (DA) respectively in June and November 2008. After a brief description of the TF Coils and Structures, the paper gives an overview of the PA showing the milestones towards series production. The procurement strategy of both DA involved is described, in particular the first step which covers pre-production activities: qualification of raw materials, manufacturing trials, mock-ups and full-scale prototype radial plates, impregnation tests and, possibly, winding trials. The work carried out by IO is also presented: optimization of the cover plate welding to satisfy the allowable stress criteria while minimizing the associated distortions, qualification of blends of cyanate ester with epoxy resin for the impregnation of the winding packs and design of the coil terminal region including integration of the needed instrumentation. [ABSTRACT FROM AUTHOR]

Abstract: The central solenoid (CS) of the ITER tokamak contributes to the inductive flux to drive the plasma, to the shaping of the field lines in the divertor region and to vertical stability control. It is made of 6 independent coils, using a Nb3Sn cable-in-conduit superconducting conductor, held together by a vertical precompression structure. This design enables ITER to access a wide operating window of plasma parameters, up to 17MA and covering inductive and non-inductive operation. Each coil is based on a stack of multiple pancake winding units to minimise joints. A glass–polyimide electrical insulation, impregnated with epoxy resin, is giving a high voltage operating capability, tested up to 29kV. The CS performance is fatigue driven mainly by the stress levels in the conductor jacket and in the precompression structure needed to keep the modules in contact during the repulsive forces which can arise in operation. A rigid connection to the TF coils provided at one end and a centering support at the other end allow to resist net vertical forces as well as unbalanced radial forces while avoiding torsion transmission from the TF Coils to the CS assembly. [Copyright &y& Elsevier]

Abstract: The first 2 years of the ITER IO has seen substantial progress towards the construction of the magnets, in three main areas. Firstly, the design has been developed under the conflicting constraints to minimise construction costs and to maximise plasma physics performance. Building construction momentum while updating the design to take account of new physics assessments of the coil requirements has been challenging. Secondly, with a stabilising design, it has been possible for the Domestic Agencies to launch the first industrial procurement contracts. And thirdly, critical R&D to confirm the performance of the Nb3Sn cable in conduit design is proceeding successfully. The design consolidation has been accompanied by design reviews involving the international community. The reviews conducted by magnet experts have enabled a consensus to be built on choosing between some of the design options in the original ITER basic design in 2001. The major design decisions were to maintain the circular Nb3Sn conductor embedded in radial plates for the toroidal field (TF) coils and to maintain NbTi-based conductors for the PF coils. Cold testing, at low current, is also being introduced for quality control purposes for all coils. [Copyright &y& Elsevier]

An R&D program is being carried out in view of the qualification of the Toroidal Field (TF) coil conductor. It consists of an extensive characterization of several Nb3Sn strand candidates, the cabling of those strands into prototype TF conductors with various combinations of twist pitch and void fraction and the measurement of the performance of conductor short samples in representative condition of transport current, magnetic field and temperature. Industries and laboratories are involved worldwide in this R&D program which is expected to provide the basis to finalize the TF conductor design. The Nb3Sn strands proposed by each of the six Participant Teams involved in the procurement of the TF conductor will then be qualified in the final cable configuration by the successful test of a short sample in the Sultan facility. The status and the first results of the R&D program are presented together with the qualification plan. [ABSTRACT FROM AUTHOR]

Procurement of the ITER magnets is due to start at the end of 2007/early 2008, with the launch of the longest lead time items, the Nb3 Sn conductor and the TF coil windings. The base design for procurement was established in 2001, and the build up of the Cadarache ITER team has been accompanied by a review of the most critical, or controversial, features of the 2001 design. At the same time, an urgent R&D program has been launched to complete the necessary verification of the design solutions that are proposed. In this paper an overview will be presented of the main design features and drivers, and some of the recent issues and R&D results will be summarized. [ABSTRACT FROM AUTHOR]

The toroidal magnetic field of ITER is provided by 18 Nb3 Sn superconducting coils that provide 5.3 Tin the Tokamak magnetic centre thanks to a nominal current of 68 kA. Two major procurement packages will be integrated to accomplish the final assembly of the coils: 1) the winding pack which houses the double pancakes in 7 stacked radial plates (RP) and 2) the TF coil case which houses the winding pack (WP) and is the main structural component of the ITER machine. 19 TF coil cases (18+1 spare) will be entirely procured by Japan, whereas the WP and the TF coil assembly will be shared between Japan (9) and Europe (10). The insertion process of the WP inside the coil case, VPI and final closure welding is presented with an emphasis on the R&D still remaining. An appropriate metrology strategy and correct benchmarking of the WP current center line (CCL) in the coil case external walls is essential to allow the correction of the error fields during operation. The TF coils will be assembled and pre-aligned in pairs together with their corresponding VV sector and craned to the Tokamak hall hanging from 2 independent hooks. The integration of the IF coils in the Tokamak primary datum system will have to bc accomplished within the maximum tolerances defined by the error field analysis. As-built 3D data collected during manufacture will be used to ensure the correct positioning of the CCL against the Tokamak datum system and ease the alignment process. Shimming will be used in the gravity support for vertical alignment and in the upper and lower OIS for toroidal alignment. A subsequent survey of the `as-installed' TF coils will determine the final Tokamak Datum System. [ABSTRACT FROM AUTHOR]

The current design of the ITER Toroidal Field coils and structures, the main critical design and manufacturing issues, and the status of the procurement arrangements for these components, which will be released to the ITER parties in early 2008 to start the manufacturing contracts, are described. Some qualification and R&D work still required in preparation for the manufacture are also mentioned. [ABSTRACT FROM AUTHOR]

The problem of how representative a short Nb3Sn conductor sample is for simulating conductor-in-coil conditions in the ITER magnets is addressed here by considering the particular case of the Central Solenoid Model Coil (CSMC), tested in 2001-2003 at JAERI Naka, Japan, and its associated Good Joint IGJ) sample (tested in 1999 in the SULTAN facility). While the basic thermal-hydraulic conditions are interpreted from experimental data by the M&M code, the sensitivity of the voltage-temperature characteristics to mechanical effects (due to the strain sensitivity of the Nb3Sn) is investigated by analysis. The results are used to assess the errors to be expected in predicting coil performance from measurements made on conductor samples. [ABSTRACT FROM AUTHOR]

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

In the design of CIC conductors using NbTi or Nb3Sn strands cooled with supercritical He, the choice of stability criteria for near steady state operation presents a problem. The disturbances are small, under 10 mJ/cm³ of cable material, too small for design significance. The main parameters affected by the choice of stability criteria are temperature margin and Cu:nonCu ratio, both of which strongly affect cable cost and overall current density. Methods quantifying the copper required for stability have been elementary, mostly using cryogenic stability (the well-cooled condition). Recent experimental and analytical work on the electric field achieved before quench Eq suggest a more exact method. Cable quench is modeled by a simple stability simulation, including copper conduction and the superconducting transition. A heat transfer coefficient is derived from ITER model coil data. As an example it is applied to conductor design (Fig. 1) for the ITER fusion reactor magnets. [ABSTRACT FROM AUTHOR]

The Central Solenoid Model Coil (CSMC) was designed and built by ITER collaboration between the European Union, Japan, Russian Federation and the United States in 1993-2001. Three heavily instrumented insert coils have been also built for testing in the background field of the CSMC to cover a wide operational space. The TF Insert was designed and built by the Russian Federation to simulate the conductor performance under the ITER TF coil conditions. The TF Insert Coil was tested in the CSMC Test Facility at the Japan Atomic Energy Research Institute, Naka, Japan in September-October 2001. Some measurements were performed also on the CSMC to study effects of electromagnetic and cooldown cycles. The TF Insert coil was charged successfully, without training, in the background field of the CSMC to the design current of 46 kA at 13 T peak field. The TF Insert met or exceeded all design objectives, however some interesting results require thorough analyses. This paper presents the overview of main results of the testing—magnet critical parameters, joint performance, effect of cycles on performance, quench and some results of the post-test analysis. [ABSTRACT FROM AUTHOR]

The Central Solenoid Model Coil (CSMC) was designed and built by an ITER collaboration in 1993-2001. Three heavily instrumented Inserts have been also built for testing in the background field of the CSMC. The Nb[sub 3]Al Insert was designed and built by Japan to explore the feasibility of an alternative to Nb[sub 3]Sn superconductor for fusion magnets. The Nb[sub 3]Al Insert coil was tested in the CSMC Test Facility at the Japan Atomic Energy Research Institute, Naka, Japan in March-May 2002. It was the third Insert tested in this facility under this program. The Nb[sub 3]Al Insert coil was charged successfully without training in the background field of the CSMC to the design current of 46 kA at 13 T peak field and later was successfully charged up to 60 kA in 12.5 T field. This paper presents the test results overview. [ABSTRACT FROM AUTHOR]

Presents an analysis of the impact of strand joints and copper location on the current distribution and stability in cable-in-conduit conductors. Use of a time dependent model of current diffusion and heat transfer; Contribution to cable stability by separate copper strands; Current ramping with joined strands; Effects of separate copper strands.

Describes the design of the Nb[sub 3]Sn and NbTi cable-in-conduit conductors of the International Thermonuclear Experimental Reactor magnets. Optimization of the amount of superconducting strand; Conductor thermohydraulic analysis; Consideration of non-uniform current distribution among superconducting strands; Conductor design rules; Heat deposition; Conductor parameters.

The Poloidal Field (PF) coils of ITER (International Thermonuclear Experimental Reactor) will supply the necessary magnetic field to initiate, shape, control and shutdown burning plasmas. The PF coils use NbTi cable-in-conduit superconductors, which operate at maximum currents of the order of 45–60 kA and experience large variations of current and magnetic fields. In order to test full-scale NbTi superconductors at operational conditions similar to ITER, the European Team has been asked to design and manufacture a PF conductor insert (PFCI). The cable has been provided by the Russian Federation. The Insert will be tested in 2004 in the Central Solenoid Model Coil (CSMC) facility at JAERI Naka, Japan. [Copyright &y& Elsevier]