ITER Magnet System Research Articles

High power high temperature superconductor current leads at KIT

Many large fusion devices presently under construction or in operation who comprise superconducting magnets like EAST, Wendelstein 7-X (W7-X), JT-60SA and ITER use High Temperature Superconductor (HTS) current leads to reduce the cryogenic load and operational cost. The Karlsruhe Institute of Technology (KIT) designed, constructed and successfully tested a 68kA HTS current lead demonstrator for ITER which led to the decision by ITER IO to use HTS current leads for the ITER magnet system. KIT has designed, constructed and tested the HTS current leads for W7-X, which are operational since 2015, and presently manufactures and tests also the HTS current leads for JT-60SA. All these current leads consist of a meander-flow type heat exchanger (HX) which is cooled by 50K helium and an HTS section. In all cases the first generation HTS material BSCCO is used embedded in a low conductivity matrix of AgAu. The technology developed in the past 20 years is now mature and would be able to be used for future fusion reactors like DEMO.In the meantime industry worldwide – with the exception of one company in Japan and another one in China – does not produce BSCCO tapes anymore; the second generation HTS REBCO is now used for application because of the better field performance in particular at high temperature. As the new material can only be produced in a multilayer structure rather than as a multifilamentary tape, the technology developed for BSCCO cannot be directly transferred to REBCO. Therefore several laboratories, in particular in Europe, are presently investigating how an HTS current lead made with REBCO could look like. This development will be the prerequisite for the construction of current leads for future fusion reactors as they have to handle high currents in the range of up to 100kA.The article discusses the status of the current lead development at KIT.

Analysis of ITER Magnet in Safety-Related Fault Condition—Case Study for PF3

The ITER magnet system contains stored energy, 41 GJ in the toroidal field system and up to ~10 GJ in the poloidal field (PF)/central solenoid system during plasma operation. A quench in the ITER magnet is regarded as a normal event, and the ITER magnet system has a redundant quench detection system and a reliable fast discharge system in order to achieve low probability of unmitigated quench and its propagation in the ITER magnet system. An electrical circuit model is developed by using a circuit simulator in order to estimate the arc power during fault conditions. The ANSYS 3-D model of the magnet, including electrical, thermal, and arcing inside the coil, is being constructed to analyze the thermal and electrical behavior of the magnet in an unmitigated quench. In this paper, thermal analyses of the fault condition related to the PF coil and impacts to the vacuum vessel are reported.

Joule Losses in the ITER Cold Structures During Plasma Transients

The ITER magnet system, which is composed of toroidal field (TF) and poloidal field (PF) coils, central solenoid (CS), correction coils, and all their structural supports, is cooled by supercritical helium flow at 4.5 K. Eddy currents are induced in the metallic components during normal operation, due to the variation of PF and CS coils current, and to the plasma ramp-up and ramp-down. In addition, during disruptions, as well as other fast plasma transient events, the eddy currents circulating in these structures reach very high values due to the high induced electric fields. An effective cooling is therefore needed to limit the temperature increase of the magnets and their supports. The Joule energy dissipated in the cold structures of the ITER magnet system has been computed by means of the electromagnetic finite-element code CARIDDI. A model of a 40 degree sector of the ITER magnet has been built in order to represent in detail the connections between the CS and the TF coils, PF and CS supports, and all the intercoil structures and the metallic part of the TF coils (radial plates, case, etc.). Vacuum vessel, thermal shield, and cryostat have also been modeled. The reference 15-MA inductive scenario (with plasma current ramp-up duration of 80 s and ramp-down of 200 s) and a 15-MA fast inductive scenario characterized by the fastest plasma current ramp-up (50 s) and the fastest plasma current ramp-down (65 s) have been analyzed. In addition, several plasma instabilities have been simulated considering the reference electrical connections between the metallic parts.

Progress in Europe of the Procurement of the EU ITER TF Coils

The ITER magnetic system includes 18 toroidal field (TF) coils constructed using a Nb3Sn cable-in-conduit superconductor. Each TF coil comprises a winding pack (WP) composed of seven double pancake (DP) modules stacked together, impregnated, and inserted into a stainless steel coil case. Ten TF coils are being produced in Europe, under the responsibility of Fusion for Energy (F4E) (the European Domestic Agency), whereas the remaining nine TF coils are being produced in Japan. F4E has implemented a procurement strategy aimed to minimize costs and risks by subdividing the procurement into three main packages, each foreseeing first an R&D and a qualification phase. One procurement package is related to the construction of 72 radial plates (RP), another to the fabrication of the ten WPs, and a third to the cold test and coil-case insertion of ten WPs. All industrial contracts have now been signed and are running. The situation as of September 2015 is as follows: 2 RP prototypes and 32 production RPs (enough for four TF coils) have been successfully (enough for four TF coils) produced and delivered to the winding pack supplier. A full-size superconducting DP prototype has been successfully fabricated and subjected to a thermal cycle at 80 K. So far, 33 DPs have been wound, 27 DPs have been heat treated, and 26 DPs have been successfully transferred into the RP grooves. The cover plate welding has been successfully completed on 18 DPs. Regarding the insertion contract, an alternative way to insert the WP inside the coil case has been devised, and the corresponding transfer tooling is being procured. The qualification for the most important manufacturing processes is underway.

Status of the ITER Conductors in China

The ITER magnet system is made up of four main subsystems: the 18 toroidal field (TF) coils, the central solenoid, the six poloidal field (PF) coils, and the correction coils (CCs). The feeder system, with its main busbar (MB) and CC busbar (CB), represents one of the main magnet components as well. All coils and busbars with different dimensions used cable-in-conduit conductors. China has signed three conductor packages, which are the so-called procurement arrangements, between ITER and the Chinese Domestic Agency (CN DA): a TF conductor package, a PF conductor package, and a CC and feeder conductor package. They include 7.5% of the TF conductors (11); all the PF2 (12), PF3 (16), PF4 (16), and PF5 (16) conductors; all the CC (18) conductors; and the MB (3) and CB (2) conductors for the feeders. Complex technologies have been developed by ASIPP for the serial production of all ITER conductors, in terms of cabling parameter design, welding, and elaboration of cable insertion, compaction, and winding processes. China has finished all qualification phases and is well into the main series production. All conductor samples required for quality control have successfully passed the SULTAN tests with good performance. The status of the production of ITER conductors in China is described in this paper.

Review of Scientific Results Obtained During Production of ITER TF and PF Conductors in Russia

The Russian Scientific R&D Cable Institute was obliged to produce and to deliver 20% toroidal field (TF) conductors and 20% poloidal field (PF) cables for the ITER magnet system. The task has been completed in October 2015. This work was a great technological challenge in the development of tricky technologies. Development of manufacturing processes was accompanied by extensive scientific studies directed to an improvement of technologies and a search of causes resulting in change of cables and conductors properties during manufacturing and testing. In this paper, we present the extended review and summary of results obtained for Russian superconductors that permitted improving technologies and finding out the reasons of changes of properties of cables and conductors during manufacturing and testing. The study of rotation and untwisting of TF cables during insertions into jackets allowed developing the model describing the rotation process and suggesting the device which keeps untwisting within demanded levels. Analyses of mandatory SULTAN high-field electromagnetic tests of PF and TF conductors from Russia confirmed good quality and adequacy of manufacturing processes. Micrographic studies of Nb 3Sn strands before and after SULTAN tests permitted qualitatively understanding the reason for the stability of RF TF conductor behavior during multiple electromagnetic and thermal cycling. Statistical studies of residual resistance ratio (RRR) of Nb 3Sn and NbTi strands permitted tracking of its degradation during manufacturing processes and SULTAN tests. The optimal heat treatment process for Nb 3Sn conductors has been confirmed after RRR statistical study. The hydraulic performance has been measured in the short samples of TF conductor. By using the parameters obtained, it is possible to predict hydraulic performance of helium flow in a conductor at any temperature. These results led to deeper understanding and to improving manufacturing processes and quality of TF and PF conductors for the ITER magnet system and could be useful for suppliers of superconducting cables and conductors using ITER-like technologies.

Starting Manufacture of the ITER Central Solenoid

The central solenoid (CS) is a key component of the ITER magnet system to provide the magnetic flux swing required to drive induced plasma current up to 15 MA. The manufacture of its different subcomponents has now started, following completion of the design analyses and achievement of the qualification of the manufacturing procedures. A comprehensive set of analyses has been produced to demonstrate that the CS final design meets all requirements. This includes in particular structural analyses carried out with different finite-element models and addressing normal and fault conditions. Following the Final Design Review, held in November 2013, and the subsequent design modifications, the analyses were updated for consistency with the final design details and provide evidence that the Magnet Structural Design Criteria are fully met. Before starting any manufacturing activity of a CS component, a corresponding dedicated qualification program has been carried out. This includes manufacture of mockups using the real manufacturing tools to be tested in relevant conditions. Acceptance criteria have been established for materials and components, winding including joints, cooling inlets and outlets, insulation, precompression, and support structure elements.

Developments of non-destructive test method of jacket section for ITER Poloidal Field (PF) coils

The ITER magnet system is one of the largest and most complex superconducting installations ever undertaken. Therefore, all subcomponents have very demanding requirements, including the conductors which are of cable-in-conduit conductor type. Based on the structural analysis and fatigue life requirements of the coil systems, the maximum acceptable defect sizes inside the jacket material were defined. This paper presents the nondestructive test (NDT) method developed for the poloidal field (PF) coil conductor jackets made of 316L stainless steel. The difficulty in the development of the PF jacket NDT arises from the circular-in-square shape, i.e., extruded bars of a square cross section with an inner circular bore. Furthermore, in addition to volumetric defects, the inspection procedure has to detect planar defects oriented longitudinal (i.e., along the jacket) and transversal on the surfaces and inside the jacket material. After a series of experiments and optimization, the phased array ultrasonic test (PAUT) combined with eddy current testing (ECT) was developed for the PF jacket inspection. The outer surface and volume are inspected by PAUT consisting of 11 separate runs per jacket surface and using four different probes. The inner bore surface up to a depth of 2–3 mm is inspected by multifrequency ECT. The characteristics and limits of all inspections runs are discussed, particularly for the most difficult one with detecting transversal planar defects.

Hybrid integral-differential simulator of EM force interactions/scenario-assessment tool with pre-computed influence matrix in applications to ITER

A necessity to address a wide spectrum of engineering problems in ITER determined the need for efficient tools for modeling of the magnetic environment and force interactions between the main components of the magnet system. The assessment of the operating window for the machine, determined by the electro-magnetic (EM) forces, and the check of feasibility of particular scenarios play an important role for ensuring the safety of exploitation. Such analysis-powered prevention of damages forms an element of the Machine Operations and Investment Protection strategy. The corresponding analysis is a necessary step in preparation of the commissioning, which finalizes the construction phase. It shall be supported by the development of the efficient and robust simulators and multi-physics/multi-system integration of models. The developed numerical model of interactions in the ITER magnetic system, based on the use of pre-computed influence matrices, facilitated immediate and complete assessment and systematic specification of EM loads on magnets in all foreseen operating regimes, their maximum values, envelopes and the most critical scenarios. The common principles of interaction in typical bilateral configurations have been generalized for asymmetry conditions, inspired by the plasma and by the hardware, including asymmetric plasma event and magnetic system fault cases. The specification of loads is supported by the technology of functional approximation of nodal and distributed data by continuous patterns/analytical interpolants. The global model of interactions together with the mesh-independent analytical format of output provides the source of self-consistent and transferable data on the spatial distribution of the system of forces for assessments of structural performance of the components, assemblies and supporting structures. The numerical model used is fully parametrized, which makes it very suitable for multi-variant and sensitivity studies (positioning, off-normal events, asymmetry, etc). The obtained results and matrices form a basis for a relatively simple and robust force processor as a specialized module of a global simulator for diagnostic, operational instrumentation, monitoring and control, as well as a scenario assessment tool. This paper gives an overview of the model, applied technique, assessed problems and obtained qualitative and quantitative results.

Manufacturing and Fatigue Tests of PF1 Coil Helium Inlet

The Poloidal Field (PF1) coil is one of six PF coils of the ITER magnet system. The PF1 is composed as a stack of eight double pancakes (DPs). The DPs are wound from NbTi superconducting cable “cable-in-conduit” type and have cooling by liquid helium. The supplying of liquid helium is carried out through helium inlets, which are one of the critical structural elements of DPs and PF1. During the PF1 coil operating, helium inlets are undergone by cyclic mechanical loads. In this case, it is necessary to check the design of helium inlets by carry out fatigue tests at cryogenic temperature. In 2013, the design of helium inlets was changed. In accordance to these changes, the new manufacturing process of helium inlet was developed. A new helium inlet sample was manufactured. The fatigue tests of this sample were carried out in accordance to ASTM standard E739-91(1998) “Standard Practice for Statistical Analysis of Linear or Linearized Stress-Life (S-N) and Strain-Life (e-N) Fatigue Data.” This paper presents comparative analysis of manufacturing process of previous and new helium inlet designs, details, and results of fatigue tests.

Control Methodology and Test Modes During the Qualification Test of the ITER Cold Circulator

Supercritical helium cold circulators operating at unprecedented high mass flow rates are required to maintain the ITER magnet and cryopump system at operating conditions. The qualification test plan has been developed for the two cold circulators installed in the Test Auxiliary Cold Box in compliance with the Japan Atomic Energy Agency cryogenic test facility consisting of a cryoplant of ∼ 5 kW at 4.5 K. The process modelling for test modes such as cool-down/warm-up, nominal and 10% higher speed operation, including test approach, assumptions made, and expected results using the control strategies to be implemented during the test are described.

Investigation of the Fine Structure Localized Plastic Deformation Zone of Superconducting Cable Components

Using atomic force, optical and electron microscopy methods, the changes in the microstructure and phase composition were investigated for the alloy Nb47%Ti used for the manufacture of superconducting cable employed as current-carrying elements in the ITER magnetic system (abbr. for International Thermonuclear Experimental Reactor). The test samples were prepared from the superconducting wire at an intermediate step of the drawing process for the area reduction Ø1.3→Ø1.2 mm. The effect of cold drawing and intermediate annealing on the properties of as-worked Nb-Ti alloy was assessed. Local strain zones were found to occur in the rupture area. The shape and chemical composition of Nb-Ti wire was examined for both a defect-free area and the rupture area. A Nb diffusion barrier was found to occur in the copper matrix of Nb-Ti wire.

Progress in the F4E Procurement of the EU ITER TF Coils

The ITER magnetic system includes 18 Toroidal Field (TF) Coils constructed using Nb3Sn cable-in-conduit superconductor. Each TF coil comprises a Winding Pack (WP) composed of 7 Double Pancake (DP) modules stacked together, impregnated and inserted into a stainless steel coil case. Fusion for Energy [the European Domestic Agency (DA)] is responsible for the procurement of the ten while the Japanese DA is responsible for remaining nine coils. The conductors are being produced by 6 different DAs, while the coil cases only by the Japanese DA. F4E has implemented a procurement strategy aimed to minimize costs and risks, consisting of subdividing the procurement into three main procurement packages, each foreseeing first an R&D and qualification phase. One procurement package is related to the construction of 72 radial plates (RP), another to the fabrication of the ten WP, and a third to the cold test and coil-case insertion of ten WP. In collaboration with industry, F4E has successfully produced two RP prototypes. Regarding the DP, the construction of the first DP prototype has started. In this paper, we will report on the results achieved so far and the status of each of the procurement packages.

Thermal Hydraulic Analysis of the ITER PF and Correction Coils in 15 MA Scenario Operation Using the SuperMagnet Suite of Codes

The poloidal field (PF) coils and correction coils (CCs) constitute one of the four hydraulic loops of the ITER magnet system. They are cooled in parallel by the same heat exchanger. The PF coils consist of 6 different pulsed coils while the CCs include 18 coils divided into 3 different sets: bottom, side and top. The PF coils are made of dual-channel NbTi conductors whereas the CCs consist of single-channel NbTi conductors. The SuperMagnet suite of codes can combine 1-D cooling with supercritical helium flow in the ITER conductors and the 2-D heat diffusion across the coils into a full-scale quasi 3-D model. We analyzed the in-coil NbTi conductor performance and the heat load to the cryogenic plant. We also compared the results from Vincenta and SuperMagnet.

Advanced Examination Techniques Applied to the Assessment of Vacuum Pressure Impregnation (VPI) of ITER Correction Coils

The ITER Magnet System includes a set of 18 superconducting correction coils (CC) which are used to compensate the error field modes arising from geometrical deviations caused by manufacturing and assembly tolerances. The turn and ground insulation are electrically insulated with a multi-layer fiberglass polyimide interleaved composite, impregnated with epoxy resin using vacuum pressure impregnation (VPI). Adequate high voltage insulation (5 kV), mechanical strength and rigidity of the winding pack should be achieved after impregnation and curing of the insulation system. VPI is an effective process to avoid defects such dry spots and incomplete wet out. This insulation technology has also been developed since several years for application to large superconducting coils and more recently to ITER CC. It allows the coils to be impregnated without impacting on their functional characteristics. One of the critical challenges associated with the construction of the CC is the qualification of the VPI insulation. Sections issued from representative VPI test samples with real scale side correction Coil (SCC) cross-section have been delivered and characterized at CERN. High resolution micro-optical inspections have been carried out on large areas through digital microscopy. The aim was to identify lack of impregnation, areas of pure resin and void entrapments. The areas near the filling fibre glass rope received special attention. High precision dimensional and geometrical assessments have been performed with the help of image analysis. Compression and pull-out tests have been also carried out. Finally, high-resolution 3D-computed tomography has been applied for a full volumetric inspection of the sections, enabling the reconstruction in three dimensions of the VPI samples and allowing to fully detect, confirm, and image the volume defects already identified by micro-optical observations.

Moving Toward Manufacture of the ITER Central Solenoid

After several years of design optimization, the Central Solenoid (CS) of the ITER Magnet system is now moving towards manufacture. The design has evolved to take into account on one hand the results of the R&D carried out by the US ITER team in charge of the development of the design and on the other hand the feedback provided by the involvement of industry in preparation of the manufacture. To address specific issues, dedicated mock-ups have been manufactured and tested. Electromagnetic, structural and thermo-hydraulic analyses have been carried out to verify the compliance of the design with the ITER design criteria. A review of the Final Design is planned in 2013, preparing then to move into the manufacturing phase.

Overview of magnetic control in ITER

ITER is targeting Q=10 with 500MW of fusion power. To meet this target, the plasma needs to be controlled and shaped for a period of hundreds of seconds, avoiding contact with internal components, and acting against instabilities that could result in the loss of control of the plasma and in its disruptive termination.Axisymmetric magnetic control is a well-understood area being the basic control for any tokamak device. ITER adds more stringent constraints to the control primarily due to machine protection and engineering limits. The limits on the actuators by means of the maximum current and voltage at the coils and the few hundred ms time response of the vacuum vessel requires optimization of the control strategies and the validation of the capabilities of the machine in controlling the designed scenarios.Scenarios have been optimized with realistic control strategies able to guarantee robust control against plasma behavior and engineering limits due to recent changes in the ITER design. Technological issues such as performance changes associated with the optimization of the final design of the central solenoid, control of fast transitions like H to L mode to avoid plasma-wall contact, and optimization of the plasma ramp-down have been modeled to demonstrate the successful operability of ITER and compatibility with the latest refinements in the magnetic system design.Validation and optimization of the scenarios refining the operational space available for ITER and associated control strategies will be proposed. The present capabilities of magnetic control will be assessed and the remaining critical aspects that still need to be refined will be presented. The paper will also demonstrate the capabilities of the diagnostic system for magnetic control as a basic element for control. In fact, the noisy environment (affecting primarily vertical stability), the non-axisymmetric elements in the machine structure (affecting the accuracy of the identification of the plasma boundary), and the strong component of eddy current at the start-up (resulting in a poor S/N ratio for plasma reconstruction for Ip<2MA requiring a robust plasma control) make the ITER magnetic diagnostic system a demanding part of the magnetic control and investment protection systems. Finally the paper will illustrate the identified roles of magnetic control in the PCS (plasma control system) as formally defined in the recent first step of the design and development of the system.

Seismic analysis of ITER fourth PF (Poloidal Field Coil) feeder

The ITER feeder systems connect the ITER magnet systems located inside the main cryostat to the cryo-plant, power-supply and control system interfaces outside the cryostat. The main purpose of the feeders is to convey the cryogenic supply and electrical power to the coils as well as house the instrumentation wiring. The PF busbar which carries 52kA current will suffer from high Lorentz force due to the background magnetic field inspired by the coils and the self-field between every pair of busbars. Except their mechanical strength and thermal insulation performance must be achieved, the dynamic mechanism on PF structure should be assessed. This paper presents the simulation and seismic analysis on ITER 4th PF feeder including the Coil Terminal Box and S-bend Box (CTB and SBB), the Cryostat Feed-through (CFT), the In-Cryostat-Feeder (ICF), especially for the ground supports and main outer-tube firstly. This analysis aims to study seismic resistance on system design under local seismograms with floor response spectrum, the structural response vibration mode and response duration results of displacement, membrane stress, and bending stress on structure under different directions actuating signals were obtained by using the single-seismic spectrum analysis and Dead Weight analysis respectively. Based on the simulative and analytical results, the system seismic resistance and the integrity of the support structure in the 4th PF feeder have been studied and the detail design confirmed.

Challenges and status of ITER conductor production

Taking the relay of the large Hadron collider (LHC) at CERN, ITER has become the largest project in applied superconductivity. In addition to its technical complexity, ITER is also a management challenge as it relies on an unprecedented collaboration of seven partners, representing more than half of the world population, who provide 90% of the components as in-kind contributions. The ITER magnet system is one of the most sophisticated superconducting magnet systems ever designed, with an enormous stored energy of 51 GJ. It involves six of the ITER partners. The coils are wound from cable-in-conduit conductors (CICCs) made up of superconducting and copper strands assembled into a multistage cable, inserted into a conduit of butt-welded austenitic steel tubes. The conductors for the toroidal field (TF) and central solenoid (CS) coils require about 600 t of Nb3Sn strands while the poloidal field (PF) and correction coil (CC) and busbar conductors need around 275 t of Nb–Ti strands. The required amount of Nb3Sn strands far exceeds pre-existing industrial capacity and has called for a significant worldwide production scale up. The TF conductors are the first ITER components to be mass produced and are more than 50% complete. During its life time, the CS coil will have to sustain several tens of thousands of electromagnetic (EM) cycles to high current and field conditions, way beyond anything a large Nb3Sn coil has ever experienced. Following a comprehensive R&D program, a technical solution has been found for the CS conductor, which ensures stable performance versus EM and thermal cycling. Productions of PF, CC and busbar conductors are also underway. After an introduction to the ITER project and magnet system, we describe the ITER conductor procurements and the quality assurance/quality control programs that have been implemented to ensure production uniformity across numerous suppliers. Then, we provide examples of technical challenges that have been encountered and we present the status of ITER conductor production worldwide.

Aging Tests of the High Current Aluminum–Copper Contact Connections in the ITER DC Busbar System

In ITER, the superconducting coils will be connected with ac/dc converters by water-cooled aluminum busbars, whose segments will be interconnected by copper flexible inserts to compensate for cyclic thermal expansion. This paper presents the experimental data obtained during development and testing of high-current aluminum-copper contact connections with various coatings protecting against corrosion and oxidation including multilayer and composite coatings, coatings with an additional interlayer. With the testing results, the coating method has been chosen for the protection of the contact surfaces of aluminum busbars and flexible copper inserts, which will be used during manufacturing of the busbars for the power supply system of the ITER magnet system coils.