Search

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 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]

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 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]

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 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 contains 18 Toroidal Field Coils (TFC). These are large D-shaped coils of about 300 t, 17.5-m height and 9-m width. They consist of a Winding Pack (WP) enclosed in a rigid structural steel case, the Toroidal Field Coil Case (TFCC). The WP is a bonded structure of 7 Double Pancakes (DP), each made up of a radial plate (RP) housing the reacted Nb3Sn cable-in-conduit superconductor (CICC), which operate at 4.5 K in supercritical helium. The conductor carries a current of 68 kA in operation to produce a nominal peak field of 11.8 T. The total stored magnetic energy in the 18 TFCs is 41 GJ. While the Japanese and European Domestic Agencies that are in charge of the procurement of the TFCs are progressing with the manufacturing design and the fabrication trials prior to launch the production of the real coils, the ITER Organization (IO) is completing the development and qualification of the most critical items, e.g. cyanate ester and resin blends for the conductor and WP insulation system, the terminal region, the helium inlet, a charged resin system for the filling of the gap between the WP and the TFCC and the general tolerancing especially at the interfaces between the neighboring systems. This paper presents the final design of the TFCs and the results of the developments carried out in the aforementioned areas in the last 2 years. [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]

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]

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 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]