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The system of correction coils (CC) is a component of the ITER magnet system, required to correct toroidal asymmetries and reduce error magnetic fields detrimental for physical processes in the plasma. It includes 18 coils, inserted in between toroidal field coils and poloidal field coils and split into 3 sets of 6 coils each: bottom correction coils (BCC), side correction coils (SCC), and top correction coils (TCC). BCC and TCC are planar coils, whereas SCC are wound on a cylinder. All CC coils are wound using a 10 kA NbTi cable-in-conduit conductor and are manufactured by ASIPP laboratory (Institute of Plasma Physics, Chinese Academy of Sciences), under the responsibility of ITER China. A manufacturing line was installed in 2013 at ASIPP in a dedicated workshop for the construction of the CC. In order to qualify the manufacturing procedures, a comprehensive qualification program has been set up. This program includes a set of mock-ups, manufactured according to the process to be used for the coils and submitted to different tests. These qualification items are winding, insulation and vacuum pressure impregnation, helium inlet/outlet, terminal joints, case material, filler material between winding-pack and case, case assembly, and terminal service box. Qualification of conductor winding, He inlet/outlet manufacture, winding-pack turn and ground insulation installation and impregnation, case material, winding-pack-case filler material is achieved. This included mechanical testing of materials at room and cryogenic temperature in specialized testing laboratories and high-voltage tests performed at the CC workshop. Joint qualification, relying on electrical tests of joints in a dedicated test facility, is nearly complete. Remaining qualification items are case assembly, winding-pack insertion into case, and case closure welding. Manufacture of the first coil started in 2015 and its winding-pack is near completion. [ABSTRACT FROM PUBLISHER]

The International Thermonuclear Experimental Reactor (ITER) has 60 high-temperature superconducting (HTS) current leads (CL) which have a nominal current of 68, 55 and 10 kA (depending on the type of magnet they have to supply) and provide a total current of 2.7 MA to the magnet system. The HTS module is a key part of CLs. This paper describes the simulation of the critical current in a multi-layer tape model as well as the special soldering technique developed for the HTS stack. In order to validate this technique, a 10 kA pre-prototype HTS module was made and tested in a setup that reproduces ITER conditions at ASIPP. The paper also briefly describes the overheating time test setup and the results. [ABSTRACT FROM PUBLISHER]

The ITER magnet system consists of 4 sub-systems: 3 sets of 6 Correction Coils (CC), a Central Solenoid (CS), 6 Poloidal Field coils (PF) and 18 Toroidal Field Coils (TFC). The Correction Coil system aims to correct particular asymmetries and reduce error magnetic fields responsible for locked modes which can induce plasma disruptions. During the plasma scenario, thermal shield radiation, eddy current losses in structures, conduction from the TFC casing and AC losses in the NbTi superconductor will decrease the temperature margin of the CC conductor. [ABSTRACT FROM PUBLISHER]

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]

In this paper we discuss the predictive analysis performed in support of the test program of the International Thermonuclear Experimental Reactor (ITER) Poloidal Field Conductor Insert (PFCI). A subset of the test program items was considered, with particular emphasis on DC performance and AC losses. The results and implications of the comparison of selected predictions from different laboratories will be presented. [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]

The ITER CS contains 6 identical but electrically independent modules. ITER plasma shaping requirements yield a wide range of radial (or hoop) loads and a wide range of vertical loads on each CS module. Pre-compression is used to maintain contact between the modules during all stages of plasma operation, to minimize primary tensile stresses in the coil insulation, and to prevent toroidal ratcheting of the modules during repeat operation. The preload requirements for the ITER Central Solenoid ICS) are the subject of this paper. Axial pre-compression of CS stack is carefully analyzed using finite element (FE) models. Conflicting requirements on the pre-compression structure (namely, the generation of large axial pre-compression with minimal use of radial space outside the modules) are balanced to support the desired plasma shapes. [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]

Cable-in-conduit conductors (CICC) consisting of about 1000 Nb3 Sn strands with an outer diameter of about 0.8 mm, have been designed for the TF and CS coils of the ITER. The rated currents of these coils are 68 and 40 kA, respectively. Two joint types (Butt and Lap) were developed during the CS Model Coil project. The performance of these joints was evaluated during the operating tests and satisfactory results were obtained. The joints of the TF coils are located outside of the winding in a region where the magnetic field is about 2.1 T, a very low value as compared to the maximum field of 11.81 at the winding. The CS joints are located at the coil outer diameter and embedded within the winding pack to minimize the space occupied around the coil. The maximum fields at the CS joint and winding are 3.5 and 13 T, respectively. For the TF coils and the CS, the joints are cooled in series with the conductor at the outlet. The maximum allowable temperature increase due to the joule heating in the joints is 0.15 K to reduce the heat load on the refrigerator. The stability of these joints at the ITER operating condition was analyzed by using the experimental data. The analysis shows that both joint types are applicable to the full size ITER coils. [ABSTRACT FROM AUTHOR]

The ITER Central Solenoid (CS) is 12 m high and 4 m in diameter. The CS consists of a stack of 6 electrically independent modules to allow control of plasma shape. The modules are compressed vertically by a p re-compression structure to maintain contact between modules. The CS conductor is CIC conductor with Nb3 S ii strands and a steel conduit. The CS model coil and insert coil test results have shown that the conductor design must be modified to achieve an operation margin. This required either to increase the cable diameter or to use strand with a higher current capability. A bronze-process (NbTi)3Sn strand is proposed to achieve a higher critical magnetic field. A square conduit with a high Mn stainless steel is proposed as it can satisfy fatigue requirements. The inlets are in the high stress region and any stress intensification there must be minimized. The preload structure is composed of a set of tie-plates, flanges, buffer plates, and wedges. These design proposals satisfy all ITER operational requirements. [ABSTRACT FROM AUTHOR]

Testing of the five International Thermonuclear Experimental Reactor (ITER) model coils is nearly complete and it is possible to produce an overall summary of the conductor performance assessment. The interpretation is supported by a large database of tests on components (especially strand) and by analysis of the two main phenomena that influence the strand performance in the cable (current non-uniformity and mechanical loads). Individually, interpretation of individual coils is difficult but taken together the results enable conclusions to be drawn about the expected behaviour of conductors in the ITER coils. The conductor design has successfully resolved many of the problems of stability and pulsed current behaviour associated with large multi-strand cables, and the performance is better than expected in these areas. The current sharing behaviour is below expectations, probably due to an extra unexpected strain effect in the Nb3Sn strands arising from the local magnetic loads, and the paper concentrates on this issue. Design modifications are required to recover the performance and updated design criteria to implement these are proposed. [Copyright &y& Elsevier]

Abstract: Error fields in Tokamaks are small departures of the exact axisymmetry of the ideal magnetic field configuration. Their reduction beyond a threshold value by the error field correction coils is essential since sufficiently large static error fields lead to discharge disruption. The error fields are originated not only by coils fabrication and installation alignment tolerances, joints and busbars but also due to the presence of ferromagnetic elements. The start of plasma current flattop with relatively low plasma density is considered as a critical state of the 15MA scenario for the onset of locked modes causing disruptions. A figure of merit, the ‘3-mode’ criterion, based on the lowest error field harmonics [(1,1); (2,1); (3,1)] was chosen to assess the error fields expected in ITER. Analysis performed last years in an independent way by CREATE (EU) and Efremov Inst. (RF) groups has allowed a deep understanding of the error fields induced by all the possible sources. Both groups were successfully benchmarked with the estimation of the error fields induced by a given deformed shape of a TF coil obtaining a perfect match of the results up to 2 orders of magnitude smaller than the defined threshold of B 3-mode/B to <5×10−5. Three different sets of independent variables based on a 3D rigid body movement of the coils have been provided for the 6 CS modules, 18 TF coils and 6 PF coils tolerances, which have allowed a clear understanding of the weight of each of the variables in the induced error fields. The results obtained in 2008 with a realistic set of magnets tolerances concluded that the system of correction coils provides effective suppression of the error fields with margin to correct possible impact of other sources. In particular, it has been shown that superconducting joints, feeders and busbars play a secondary effect; however the radial position of the TF coils and the tilt and radial shift of the CS stack would have a relevant influence. The ensuing recent sets of variables studied aimed at deepening the understanding of the operational limits. The present paper addresses the impact of the fabrication tolerances and installation misalignments of the magnet system; the impact of ferromagnetic inserts and busbars as well as the design of the correction coils will be thoroughly covered elsewhere (Amoskov et al., in press ). [Copyright &y& Elsevier]

Abstract: The ITER superconducting magnet system generates an average heat load of 23kW at 4K to the cryoplant, from nuclear and thermal radiation, conduction and electromagnetic heating, and requires current supplies 10–68kA to 48 individual coils. The helium flow to remove this heat, consisting of supercritical helium at pressures up to 1.0MPa and temperature between 4.3 and 4.7K, is distributed to the coils and structures through 30 separate feeder lines. The feeders also contain the electrical supplies to the coil, helium supply pipes and the instrumentation lines, and are integrated with the current lead transitions to room temperature. The components consist of the in-cryostat feeders, the cryostat feedthroughs and the coil terminal boxes (CTBs). This paper discusses the functional requirements on the feeder system and presents the latest design concept and parameters of the feeder components. [Copyright &y& Elsevier]

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]