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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 Procurement of Feeder system in Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) for The International Thermonuclear Experimental Reactor (ITER) Tokamak device has been carrying out after the signature of Procurement Arrangement in October 2011. The relevant main tasks in 2012 and 2013 are the qualification of some key subcomponents, including high temperature superconducting current leads (HTS-CL), superconducting joints, 80-K thermal shield (TS), vacuum barrier, in-cryostat feeder, cold mass support, etc. The production of manufacturing drawings, preparation of manufacturing plan, manufacture and qualification of tooling, and qualification of subcomponents are all involved in this phase. This paper describes the progress made by ASIPP and some sub-contractors for the feeder's qualification work. [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]

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]

The superconducting magnet system of ITER has \sim5,000 tons of cold mass as steel structures to support the gravity and electromagnetic forces. These steel structures need to be maintained at \sim4.4 K during magnet operations. The magnet system has three STR (structure) feeders for distribution, monitoring and control of helium to remove heat loads from these large steel structures during cool down and normal operation. One SHe (supercritical helium) circulation pump for flow rate of \sim2.7 kg/s, in a closed loop is used only for cooling of structures, and these 3 STR feeders are part of this loop. Each STR feeder has 3 main components: SCVB (structure cooling valve box), CFT (cryostat feed through) and ICF (in-cryostat feeder). Unlike the coil feeders, STR feeders do not have bus bars, current leads & SBB (“S”-bend box), and present some difference in the design approach. The SCVB is placed outside the enclosure of the nuclear radiation shield and hence contains elements needing maintenance like valves and instruments required for diagnosis and operation of the cooling paths. Each SCVB has to accommodate about 20 valves within a limited available space. Some SHe lines will have running length of about 50 m from SCVB to coil structures, but the layout is such that the use of expansion joints can be avoided. The details of the design approach and analysis results will be presented in this paper. [ABSTRACT FROM PUBLISHER]

The ITER magnet system is to be operated at about 4.5 K. It has 31 feeders for high-current power supply, helium distribution and instrumentation cable routing. The feeders (except the smaller ones which are for instrumentation only) are 25 to 30 m long for about 40 tons. They have 3 main components: coil terminal box-cum-S-bend box (CTB-SBB), Cryostat feedthrough (CFT) and in-cryostat feeder (ICF). The CTB is the feeder component placed farthest from the coil (about 25 m away from the magnet center) and outside the bio-shield (the nuclear radiation restriction wall) so that it can be easily accessible in case the critical elements placed within it need maintenance: high-temperature superconducting current leads (HTSCL), large number of cryogenic control valves and pressure-release valves for quench protection. This paper explains the cryogenic design that considers these factors and main components. [ABSTRACT FROM PUBLISHER]

ITER has a large superconducting magnet system with about 10,000 tons of cold mass to be maintained at about 4.5 K during magnet operation. This cold mass has to be cooled down from \sim300 to 4.5 K simultaneously and uniformly to avoid high thermal stress. More than 5,000 tons of cold mass corresponds to steel structures to support the gravity and electromagnetic forces. A controlled cool down is necessary to avoid high thermal stress in these materials. The variations of hydraulic impedances and cold masses of different magnet components play a major role in making the cool down non-uniform if appropriate control strategies are absent. A detailed analysis has been done to quantify the factors preventing a uniform cool down. The slowest cooling component will dictate the cool down rate of the whole magnet system. The analysis finds the maximum cool down rates possible for different components and controlled flow rate required for different components to have uniform cool down of all components. Based on this, the cool down rates and duration have been found. This paper will discuss details of these analysis 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 ITER Poloidal Field (PF) magnet system consists of six coils. Niobium-Titanium (NbTi) is used as superconducting material and cable-in-conduit conductor(CICC) type are used as a conductor. All coils are fabricated by stacking 6 to 9 double-pancakes wound by two-in-hand winding scheme. The six PF coils (PF1 to PF6) are attached to the Toroidal Field (TF) coil cases through flexible plates or sliding supports to allow small radial and vertical displacements. The outer diameters of the coils vary between 8 m and 24 m. Since the PF coil system provides magnetic field for plasma shaping and position control together with the Central Solenoid (CS) coil, it needs to operate in fast pulse mode, leading to induced voltages of up to 14 kV on the coil terminals during operation. [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 International Thermonuclear Experimental Reactor (ITER) superconducting magnet system requires current supplies from 10 kA to 68 kA. The components relating the electrical power and cryogens through the warm-cold barrier to the ITER magnets are the so-called feeders. In total there are 31 such feeders for the ITER magnet system, 9 of them necessary for the toroidal field magnet system are called TF feeders. Each feeder consists of three main units: the In-Cryostat Feeder (ICF), the Cryostat Feed-Through (CFT) and the S-Bend Box (SBB)/Coil Terminal Box (CTB) assembly. This paper presents the latest design of the TF feeder components. [ABSTRACT FROM AUTHOR]

The nuclear radiation levels, magnetic fields and general access restrictions in the ITER Tokamak are such that it is undesirable to place remote control helium supply valves close to the coils. The Coil Terminal Boxes (CTB), which are further away from the coils, are therefore designed for easy access and contain as much as possible the components needing maintenance, such as the cryo-valves. The CTBs also house the electrical cold-warm transitions with the HTS current leads and the NbTi feeder busbars. Also most of the feeder instrumentation is located in the CTB. Based on the functional specification by ITER, ASIPP has manufactured and tested a prototype CTB for the ITER magnet system. This test allowed for the verification of the mechanical design of the cryostat and the thermal shield design. Following a discussion of the ITER CTB design, with particular emphasis on the concepts of flow distribution for the large ITER magnet system, considering safety and maintenance, this paper will also present the results of the prototype CTB development. [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]

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]

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]

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]

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]

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]