The first series components of large D-shaped toroidal field coils (TFC) on the ITER Tokamak project are being fabricated and assembled at European Fusion for Energy (F4E) and Japanese Domestic Agency (JADA) premises since 2013. The TF magnet system consists of 18 individual coils connected in series based on a Nb 3 Sn cable-in-conduit conductors supplied by a 68-kA rated current with an overall 41-GJ stored energy and a peak magnetic field of 11.8 T. One of the key challenges of the construction of the 18 TFCs and their assembly resides in the control of the integration of the large individually manufactured coil components and in the ultimate management of tolerances on the final assembly into the Tokamak pit. This paper presents the integration aspects related to main TFCs subcomponents under fabrication starting from the TF conductor production, the winding of individual double pancakes, and their heat treatment and impregnation. This includes the fabrication of key prototypes for qualification purpose such as helium supply inlets, the electrical joints, and the design of the winding pack insertion into the structural TFC case during the final welding enclosure. Each preassembled 40° sector of a TFCs pair is then integrated into the torus according to tight tolerance requirements to provide both the so-called TF magnetic center line data and to guarantee the final operating wedged design into the inner leg region. The assembly of the coil's terminal is then completed by connecting services through the power feeder busbars, the quench detection high voltage cables and the cryogenics interfaces pipe system.

The ITER organization has performed design and its validation tests on a helium inlet structure for the ITER Toroidal Field (TF) coil under collaboration with CERN, KIT, and CEA-Cadarache. Detailed structural analysis was performed in order to optimize the weld shape. A fatigue resistant design on the fillet weld between the shell covers and the jacket is an important point on the helium inlet structure. A weld filler material was selected based on tensile test at liquid helium temperature after Nb3Sn reaction heat treatment. To validate the design of the weld joint, fatigue tests at 7 K were performed using heat-treated butt weld samples. A pressure drop measurement of a helium inlet mock-up was performed by using nitrogen gas at room temperature in order to confirm uniform flow distribution and pressure drop characteristic. These tests have validated the helium inlet design. Based on the validation, Japanese and European Union domestic agencies, which have responsibilities of the TF coil procurement, are preparing the helium inlet mock-up for a qualification test.

The final design for the inlets of the supercritical He (SHe) coolant of the toroidal field (TF) superconducting magnets of the International Thermonuclear Experimental Reactor (ITER) has to be evaluated on the basis of different aspects, including the interest to minimize the associated localized pressure drop. Based on previous experience made on the analysis of the SHe inlets for the ITER superconducting central solenoid, we have developed and applied a computational fluid dynamics (CFD) model to compute the pressure drop versus mass flow rate characteristic in the TF inlet mock-up which was recently tested at CEA Cadarache, France. The cable model is first calibrated and verified against experimental data from short straight samples. The calibrated model is then applied to the inlet geometry and the results of the calculation are compared with the available measurements, showing very good agreement for sufficiently anisotropic permeability of the cable region. The thus validated model is finally used to investigate the distribution of the coolant flow among the different petals and the central channel, downstream of the inlet, as well as to quantitatively assess the role of the petal wrapping on the localized pressure drop at the inlet.

In the International Thermonuclear Experimental Reactor, the nuclear radiation escaping from the vacuum vessel reaches the superconducting toroidal field (TF) magnets, affecting the temperature margin Delta T-mar, that is, the difference between the current sharing temperature and the operating temperature. The TF magnets are designed to operate at a minimum margin Delta T-mar(min) = 0.7K. Recent design activity on in-vessel components, for example, blanket, in-vessel coils for plasma stability, suggests a potential enhancement of the nuclear heat load, leading to a reduction of Delta T-mar, which is accurately assessed in the paper using the validated 4C code. For the case when the margin goes below the minimum, different possible mitigation strategies are investigated: the first considers the possible reduction of the He bath temperature from the nominal 4.3 K down to 3.8 K, and is proven to be successful. The others consider the possible increase of the dwell time between plasma pulses, and is shown to be inadequate, or the decrease of plasma pulse duration, which turns out to be effective below 300 s.

In the International Thermonuclear Experimental Reactor (ITER), the nuclear radiation escaping from the vacuum vessel reaches the superconducting toroidal field (TF) magnets, contributing to the reduction of the temperature margin Delta T-mar, i.e., of the difference between the current sharing temperature and the operating temperature. The TF magnets are designed to operate at a minimum Delta T-mar of 0.7 K. However, recent design activity on in-vessel components, e.g., the blanket and the in-vessel coils for plasma stability, suggests a potential enhancement of the nuclear heat load on the TF coils. A detailed reassessment of Delta T-mar in the TF magnets during plasma operation is thus required, together with the study of possible strategies to mitigate the corresponding margin reduction. The extensively validated Cryogenic Circuit Conductor and Coil (4C) code, developed for the thermal-hydraulic analysis of transients in superconducting magnets, including their cooling circuit, is used here for the simulation of the standard ITER operating scenario. Among the different possible mitigation strategies, we consider the variation of the cooling He mass flow rate, both in the winding and, separately, in the casing, and its effectiveness in terms of Delta T-mar is assessed.

All 21 Procurement Agreements (PAs) related to the ITER superconducting magnet system were signed in March 2011. Six Domestic Agencies (DAs) of the seven, namely China, EU, Japan, Russia, the United States and Korea have to deliver components of the magnet system like TF coils and Feeders. In order to organize the complicated international supply-chain, the ITER Manufacturing Database (MD) has been developed as a sub-project of the ITER Engineering Database (EDB) which aims to become the official Product Lifecycle Management (PLM) system of the ITER project. The first version of the PLM was released in August 2011. One of the advantages of the MD is to provide clear definitions of the work conditions and resultant deliverables to all stakeholders for a common understanding and better mutual communication. The work conditions (e.g. applicable code/standard) are basically agreed by all stakeholders after the preparation phase. They are then prepared in the MD as four customized User I/F (interface) screens: the interactive Parts List, the Manufacturing Inspection Plan (MIP), the Welding Quality Inspection and Production Plan (WQIPP) and the Tooling List. Users only upload the required reports (e.g. the mill certificate for each part, test reports) into the corresponding User I/F. Consequently, the filled up User I/F becomes a deliverable, the Manufacturing Dossier.

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 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.

The Toroidal Field Coils (TFC) for the ITER magnet system are large `D' shaped coils consisting of a Winding Pack (WP) enclosed in a stainless steel (316LN) casing. The WP is a bonded structure of 7 Double Pancakes (DP), each made up of a stainless steel radial plate (RP) housing the reacted Nb3Sn circular cable-in-conduit superconductor (CICC), which operate at 4.5 K. The cooling of the WP is assured by helium feed in the CICC through one inlet for each DP at an average mass flow of 7.9 g/s in the conductor. The helium inlet is critical from a structural point of view because it has to withstand both static and cyclic strains in the order of (10.2±2.3)×10-4 coming from energization of the TFC and plasma operation conditions. The assembly of the helium inlet around the conductor includes a fillet weld that makes it even more critical. This paper describes the analyses results performed for the (static and fatigue) structural assessment of the helium inlet. It describes the statistical approach followed to determine the number of cycles to be used in the validation tests.

It has been observed in the superconducting helical coils of the Large Helical Device (LHD) that the balance voltage signals measured between the corresponding pairs of the coil blocks contain a number of spike signals during ramp-up and ramp-down processes of excitation. The spike signals might be generated by rapid changes of the self-inductances of the coil windings due to mechanical disturbances caused by large electromagnetic forces. Pulse height analysis (PHA) has been applied to analyze these signals in order to investigate the changes of mechanical properties of the coil windings as the excitation and cooling cycles proceed. In addition, acoustic emission (AE) sensors attached to the helical coil cans are also used to detect mechanical disturbances.

The ITER Superconducting Magnet Manufacturing Database, MMD, is not only a data archive, but also a common communication platform, on which contributors to magnet manufacturing collaborate and take coordinated actions. In-kind procurement is a feature of the ITER construction. The magnet system construction involves six Domestic Agencies (DAs) plus contractors in these DAs. The six DAs are EU, Japan, Russia, the USA, Korea and China. The magnet system consists of many components like TF coils and current feeders. The ITER Organization (IO) monitors and controls quality throughout the manufacturing process. This is fundamental because the IO takes responsibility for the ITER machine even though not all large components can be tested under nominal conditions before their acceptance. For many contributors though (who have different cultures, quality assurance systems and languages), quality monitoring and control (QA/QC) represent big challenges. MMD is the web-based application that gives all contributors access to the IO server computer according to deflned privileges. Users can upload and consult updated manufacturing processes, associated drawings, procedures, inspection reports, etc.; more over, they can communicate internationally and visualize identical, updated and systematically stored datasets. The data stored in the database will be available in future site assembling and operation phases. Technically, the database is an application of ICP (ITER Collaborative Platform). ICP is a software framework for implementing database-driven applications for the ITER project. It provides a standardized web browser interface and data storage. This paper presents the design and functionality of MMD.

The measurement of current sharing temperature of Cable in Conduit Conductors is the main test for the acceptance of superconducting cables for the ITER Project. This temperature is defined as the one at which the average electric field along the cable reaches a critical value E0. Due to the complexity of the measurement, the evaluation of Tcs from the experimental results is not straightforward. Different techniques of elaboration of the raw experimental data can result in rather different estimations of Tcs. The aim of this work is to elaborate a standardized procedure of data treatment to be applied to all Tcs measurements. To this purpose, the assessment of the impact of the different measurement methodologies and data treatment techniques applied in the procedure is very important. A sensitivity analysis is therefore presented, to evaluate the influence on the final evaluation of Tcs of the main parameters and methodologies involved in the proposed procedure.

The procurement of the conductor for the ITER Toroidal Field (TF), Central Solenoid (CS), and Poloidal Field (PF) magnet systems will generate a significant amount of raw material and product test data that needs to be managed and analysed in a systematic fashion by the six responsible Domestic Agencies (DAs) and the ITER Organization (IO). To do this, ITER has developed and commissioned a conductor database, modeled after the successful CERN LHC conductor database, that can store, track, and analyse the voluminous data being generated during the production process. The database focuses on the business flow associated with superconducting strand, cabling, jacket sections, jacket assembly, jacketing, and full conductor unit length production. Within the database, individual suppliers can enter and track the performance of both raw materials and finished products, and also monitor the acceptance status of all products sent to their DA. Each DA is given a single portal through which they can manage the progress of each domestic supplier, track product performance results, enter verification test results, approve strand for delivery, and notify IO of production progress. The database can be used for long-term statistical process control of both raw material parameters and final product parameters (like Ic and Ihys), and also utilizes a sophisticated security scheme to protect the confidentiality of suppliers' proprietary data. Additionally, the database can automatically highlight non-conforming products, and can manage the organization of the documentation for deviation requests and non-conformance reports. Technically, the database is an application of ICP (ITER Collaborative Platform). ICP is a software framework for implementing database-driven applications for the ITER project, and it provides a common web browser interface and data storage architecture developed in ASP.NET and MSSQL.

The world-wide procurement of ${\rm Nb}_{3}{\rm Sn}$ and NbTi for the ITER superconducting magnet systems will involve eight to ten strand suppliers from six Domestic Agencies (DAs) on three continents. To ensure accurate and consistent measurement of the physical and superconducting properties of the composite strand, a strand test facility benchmarking effort was initiated in August 2008. The objectives of this effort are to assess and improve the superconducting strand test and sample preparation technologies at each DA and supplier, in preparation for the more than ten thousand samples that will be tested during ITER procurement. The present benchmarking includes tests for critical current $(I_{\rm c})$, $n$-index, hysteresis loss $(Q_{\rm hys})$, residual resistivity ratio $(RRR)$, strand diameter, Cu fraction, twist pitch, twist direction, and metal plating thickness (Cr or Ni).

It was reported that Lorentz force caused degradation of critical current in the ITER-TFMC conductor. We have used our novel experimental setup, which utilizes the closed electric circuit concept for critical current and stability measurements of multi-stand superconducting cables. The feature of this setup is mechanical loading applied to the multi-strand cable in the transverse direction. Significant degradation in the critical current of the cable was observed when the average compressive stress was about 20 MPa. This degradation was found irreversible after unloading. We tested the cable with epoxy or ice molds as well. No degradation was observed in the molded cables. We also tested the cable with smaller void fraction. In this case, significant degradation in critical current was observed.

We have developed a novel critical current and stability measurement experimental setup, which utilizes a closed electric circuit with a multi-strand superconducting cable. The feature of this setup is mechanical loading applied to the multi-strand cable in the transverse direction. It was reported that Lorentz forces caused degradation in the critical current of the ITER-TFMC conductor. Furthermore, these phenomena were mainly observed in the ITER full-size conductors with large Lorentz forces under high magnetic fields. The advantage of our setup is critical current measurement under mechanical stresses comparable to those in the full-size conductor under high magnetic fields. By employing an inductive critical current measurement technique, we conducted an experiment with a transport current of about 10 kA without any power supply or current leads. In our experiments, we observed significant degradation in critical currents due to a compressive stress of about 30 MPa. We applied an innovative technique to mitigate the critical current degradation in mechanically loaded Nb3Sn superconducting multi-strand cables. We molded one such cable with ice and tested it. No degradation occurred in the icemolded cable. In addition, stability was also ensured due to the large thermal conductivity of ice. Thus, we have successfully mitigated the degradation in the critical current of the Nb3Sn conductor by ice molding.