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The Central Solenoid (CS), a key component of the ITER Magnet system, using a 45 kA Nb3Sn conductor, includes six identical coils, called modules, to form a solenoid, enclosed inside a structure providing vertical pre-compression and mechanical support. Procurement of the components of the ITER CS is the responsibility of US ITER, the US Domestic Agency (USDA), while the assembly of these components will be carried out by the ITER Organization (IO). Procurement of all the coil modules was awarded in 2011 to General Atomics, while procurement of the structure is split among several manufacturers, using existing equipment, sometimes among the largest ones in the world. Assembly of the ITER CS will require a dedicated area in the ITER Assembly Hall, conventional tooling and special tooling. US ITER is in charge of the procurement of special tooling, while IO is responsible for the procurement of the conventional ones. A detailed assembly procedure is under development at US ITER, in close collaboration with IO and with the support of CEA. Procurement of the special Assembly Tooling is carried out by US ITER and the main part of the first item, the Assembly Platform, was delivered to IO in 2017.

The correction coils (CCs) are a system of 18 superconducting coils composed of top CCs, side CCs, and bottom CCs, and are required to correct asymmetries and reduce magnetic error fields detrimental to the plasma performance of the ITER machine. The CC terminals will be connected with those of the superconducting busbars with twin-box joints. Qualification of the manufacturing procedure of the coil terminals is achieved by performing electrical tests of prototype joints in relevant conditions of current, temperature, and background field (4.5 K, 10 kA, and 2.5 T). In order to control the dc resistance and ac loss in series production, special tooling and manufacturing processes were developed. The main technological steps are conductor dejacketing, nickel removal, and silvering of the cable in the terminal, followed by cable compaction and tungsten inert gas welding of the termination box. To this end, two joint samples were manufactured in 2015 and 2016. The performance of the joint samples was verified before and after 1000 electromagnetic cycles. The qualification tests were carried out in the SULTAN facility and ASIPP facility. The test results show the dc resistance below 5- $\text{n}\Omega $ criterion and ac losses below 7 J/cycle.

The superconducting coils of the ITER magnet system have hundreds of electrical lap joints interconnecting superconducting cables. The joints operate in a magnetic field of up to 4 T, field derivatives of 0.5 T/s, and currents up to 70 kA. The acceptance tests for the toroidal field (TF), poloidal field (PF), and correction coil (CC) coils will be performed at 77 K, before they are assembled in the pit. Hence there will be no possibility to measure the resistance of the joints in the superconducting state before the whole magnet system is enclosed in the Tokamak cryostat. In addition, no reliable nondestructive method has been found to spot the joints with a failure at room temperature. Therefore, the production of the joints relies on the strict adhesion to established robust manufacturing procedures during the qualification phase. As additional quality monitoring, a periodic test of the joint samples manufactured in parallel with a coil fabrication is foreseen to control the reproducibility of the joint electrical performance. In order to qualify the manufacturing procedures, to establish a series production tools and worker teams, a comprehensive qualification program has been set up for manufacturers of the coils in Russia (Poloidal Coil 1, PF1), China (PF6, feeders, CC), Japan (TF), Europe (TF and PF), and USA (Central Solenoid, CS). This program includes a set of mockups manufactured according to the process to be used for the coils and submitted to different tests. They include mechanical testing of materials, electrical tests of full size joint samples, destructive microscopic examination of the joint mockups, and mechanical testing of the full size joint mockups. All tests are carried out in specialized laboratories qualified for this type of work. This paper describes the main items of the qualification program, the tests performed, and the acceptance criteria. The test results are reported and compared to the criteria.

The ITER central solenoid (CS) components are currently being manufactured. This Nb3Sn superconducting magnet will provide the magnetic flux swing required to induce up to 15 MA as plasma current. It includes six identical coils, called modules, stacked on top of each other to form a solenoid, enclosed inside a structure split into nine subsets, to provide vertical precompression and mechanical support. High mechanical stresses in materials and high voltages call for the use of structural materials with high strength and toughness and high dielectric strength insulating materials, respectively. The pulsed operation imposes materials with high fatigue strength at cryogenic temperatures. Unlike for the structure, where large existing manufacturing tools were usable, the modules required the construction of a dedicated manufacturing line. A comprehensive qualification programme is performed at the manufacturers before applying procedures for the production of the CS components. The main characteristics of the CS components, their manufacturing routes and the different elements of the qualification programme are described. The overall plan for the manufacture is reported. The status of the first series production components manufactured is presented as well as the planned delivery schedule to the ITER site.

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.

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.

The manufacturing line of the ITER Correction Coils (CC) at ASIPP in Hefei (China) was completed in 2013 and the manufacturing line of the ITER Central Solenoid (CS) modules is under installation at General Atomic premises in Poway (USA). In both cases, before starting production of the first coils, qualification of the manufacturing procedures is achieved by the construction of a set of mock-ups and prototypes to demonstrate that design requirements defined by the ITER Organization are effectively met. For each qualification item, the corresponding mock-ups are presented with the tests to be performed and the related acceptance criteria. The first qualification results are discussed.

The ITER correction coil (CC) case reinforces the winding packs against the electromagnetic loads, minimizes stresses and deformations to the winding pack. The cases are made of high strength and high toughness austenitic stainless steel (316LN) hot rolled heavy plate and have a thickness of 20 mm. Considering the small cross-section and large dimensions of the case, deformation of the case when welding becomes a challenge in the case manufacturing. Therefore, laser welding was developed as the main welding technology for manufacturing. In this paper, multi-pass laser welding technology is used, the laser weldability of a 20 mm thick 316LN austenitic stainless steel plate is studied and the microstructure of the welded joint is analyzed. The welding experiment used an YLS-6000 fiber laser (IPG) and weld filler of 316LMn to match the base metal was used. The result shows that the welded joint has no obvious surface and internal defects based on the optimized welding parameters. The weld joint have a fine austenite microstructure and display columnar dendrites and cellular grains with strong directional characteristics. No apparent heat affected zone is observed and approximately 2 μm an austenite microstructure of the fusion line is clearly presented.

ITER correction coil (CC) cases have characteristics of small cross section, large dimensions, and complex structure. The cases are made of heavy thick (20 mm), high strength and high toughness austenitic stainless steel 316LN. The multi-pass laser welding with hot wire technology is used for the case closure welding, due to its low heat input and deformation. In order to evaluate the reliability of this welding technology, 20 mm welding samples with the same groove structure and welding depth as the cases were welded. High purity argon was used as the shielding gas to prevent oxidation because of the narrowness and depth of the weld. In this paper investigation of, microstructure characteristics and mechanical properties of welded joints using optimized welding parameters are presented. The results show that the base metal, fusion metal, and heat affected zone (HAZ) are all have fully austenitic microstructure, and that the grain size of fusion metal was finer than that of the base metal. The welding resulted in an increase of hardness in the fusion metal and HAZ. It was confirmed that the tensile strength of fusion metal was higher than that of base metal and the impact toughness value is higher than industry standard requirement. Thus, this welding process was determined to be reliable for manufacture of the ITER CC cases manufacture.

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.

The ITER correction coils (CC) include three sets of six coils each, distributed symmetrically around the tokamak and inserted between the toroidal field (TF) and the poloidal field (PF) coils. Each pair of coils located on opposite sides with respect to the plasma is series connected with polarity such to produce asymmetric fields. These superconducting coils use a cable-in-conduit conductor, insulated, wound into multiple pancakes and inserted inside an austenitic stainless steel case. The requirements and the main features of the design are presented and the selected options reviewed in terms of their criticality in achieving the specified tolerances. The requested qualification trials are identified and reports the results obtained so far.

The Correction Coils system (CC) within ITER, is intended to reduce the range of magnetic error fields created by assembly or geometrical imperfections of the other coils used to confine, heat, and shape the plasma. The proposed magnet system consists of three sets of 6 coils each, located at the top (TCC), side (SCC) and bottom (BCC) of the Tokamak device and uses a NbTi cable-in-conduit superconducting conductor (CICC) operating at 4.2 K. The ITER Organization (IO) and the Institute of Plasma Physics at the Chinese Academy of Sciences (ASIPP) are jointly preparing the definition of the technical specifications and the upcoming qualification program for the Correction Coils. The proposed design consists of a one in hand conductor winding without internal joint inserted in a structural casing which reacts the electromagnetic loads. The development of major items such as terminal joints, casing manufacture, and vacuum impregnation system, is an essential phase before the series production which will take place at the premises of the supplier. This paper discusses the key technologies on CC coils and future plans for short sample prototypes fabrication.

ITER correction coil (CC) cases have characteristics of small section, large dimensions, and complex structure. The case welding deformation should be controlled within ±2 mm, and the case weld is required to meet the ISO 5817 class B. In order to achieve the requirements of weld, laser welding with hot wire was chosen for the closure welding of the CC case. The biggest advantage of the welding method is the small amount of heat input and small welding deformation. A series of welding experiments and a scaled trial with the same welding joint designed have been carried out and the preliminary welding specifications have been summarized. This paper introduces the prequalification of the case closure welding which describe the details like the welding groove design, the welding specification, and the results of the destructive and nondestructive test. As can be seen from the results, the welding quality and deformation are very close to the design requirements which may be helpful for the real qualification in the future.

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).

The mission of the JT-60SA Tokamak, which will be built in Japan, is to contribute to the early realization of fusion energy in support and supplement of the ITER program. The JT-60SA project is part of the broader approach for fusion energy. In 2008, due to a design change of the TF cross section, and following the redefinition of the global JT-60S A features, mechanical analyses were redone. The first 2D and 3D mechanical analyses were performed on the current TF design, providing useful information on the pros and cons of the new design. For the 2D analysis, the critical area of the inner leg cross section in the location of the equatorial plane was meshed to study the effect of in plane loads. The analysis includes checking the peak field applied to the conductors and the stress on each of the components of the winding pack cross section, in particular in the insulation which is a critical component. The stress in winding pack cross section is due to the accumulation of cool down strain, ?in plane? Lorentz forces and to the quench pressure. Both magnetic and mechanical analyses are performed with the ANSYS V11.0 code. In this paper, we present the main steps of the 2D and 3D FEM calculations which were developed by CEA and used in the following analyses. Associated statements concerning possible TF design optimization with respect to cost, feasibility or risk are also presented.

In the present study a complete neutronic analysis has been performed for the current design of the JT-60SA toroidal field coil (TFC) system. The MCNP5 Monte Carlo code has been used to calculate the nuclear heating, neutron spectra and absorbed dose in the TFC components, assuming a DD neutron emission rate of 1.5 × 1017 n/s (and 1% DT). Nuclear heating of the winding pack is lower than 0.3 mW/cm3 and the maximum nuclear heating of the TFC case is 0.4 mW/cm3. The overall nuclear heating, including the safety margin, is less than 8 kW. Spatial distribution of the nuclear heating has been provided along poloidal, radial and toroidal directions as to be used for thermo-hydraulic analysis and the design of TFC system. The absorbed dose to insulator is as low as to avoid the replacement during the whole life of the machine. Neutron fluxes have been used as input for a preliminary activation analysis performed with FISPACT inventory code. Activity and contact dose rates have been calculated at different cooling times, after 10 years of operations in some representative zone of the winding pack and the case. All the TFC materials can be easily recycled within the first day after shutdown and the hands-on recycling is possible within less than 30 years.

The broader approach agreement between Europe and Japan includes the construction of a fully superconducting tokamak, the JT-60 Super Advanced (JT-60SA), as a satellite experiment to ITER. In particular, the whole Toroidal Field magnet system, described in this paper, will be provided to Japan by the EU. All the TF coil main constituents, i.e. conductor, winding pack, joints, casing, current leads, are here presented and discussed as well as the design criteria adopted to fulfil the machine requirements. The results of the analyses performed by the EU and JA to define and assess the TF magnet system conceptual design are reported and commented. Future work plan is also discussed.

Latest design of superconducting magnet system in JT-60SA is presented. This magnet system consists of the TF coils and PF coils (CS and EF coils). For this magnet system, stress analyses are systematically carried out to optimize the structural design. In the analysis of TF coil, it is clarified that jacket of conductor has significant strength under the designed electromagnetic (EM) load. In addition, coil case has margin of mechanical strength. For the PF coils, maximum stress appears at jackets of CS and EF conductor are less than 500 MPa, which is within fatigue limit. The latest design of CS support structure is very strong, because the no gap is made between CS modules even though total repulsive force of CS is maximum with no pre-compression. Therefore, optimization or simplification of structure design is possible for all magnet systems from the latest design in the detail design phase.

The upgrade of JT-60U to JT-60 Super Advanced (JT-60SA), a fully superconducting tokamak, will be performed in the framework of the Broader Approach (BA) agreement between Europe (EU) and Japan. In particular, the Toroidal Field (TF) system, which includes 18 coils, is foreseen to be procured by France, Italy and Germany. This work covers activities from design and manufacturing to shipping to Japan. The present paper is mainly devoted to the analyses that lead to the conductor design and to the technical specifications of the joints for the JT-60SA TF coils. The conductor geometry is described, which is derived from Cable-In-Conduit concept and adapted to the actual JT-60SA tokamak operating conditions, principally the ITER-like scenario. The reported simulations and calculations are particularly dealing with the stability analysis and the power deposition during normal and off-normal conditions (AC losses, nuclear heating). The final conductor solution was selected through a trade-off between scientific approach and industrial technical orientation. Besides, the TF system connections layout is shown, derived from the industrially assessed twin-box concept, together with the associated thermo-hydraulic calculations ensuring a proper temperature margin.

In the cable-in-conduit-conductor (CICC) design of the toroidal field system for the international thermonuclear reactor (ITER) Nb3Sn is used as superconductor material. Considering the single strand performance, the crucial characteristic is the strain dependence of the critical current. Within this context, the performance of the CICC under strain is determined by the behaviour of the single strands and additional effects related to the manufacturing process. In the framework of the European fusion technology program a task has been started to investigate single strands as well as sub-size CICC performance using different cable layouts (9, 45 and 180 strands). For this systematic approach, parameters such as the void fraction, the number of pure copper strands, the void fraction or the cabling pattern have been varied. To examine the critical properties in detail, the available test facility, consisting of two experimental setups, is capable to measure the strain dependence in magnetic fields up to 14 T at 4.2 K, by applying an axial load to the samples. Measurements on such sub-size CICC samples are presented and compared to the expected performance.

The Final Design of the Central Solenoid (CS) of the ITER Magnet system is currently being completed by the US ITER Domestic Agency (USDA) and the manufacturing line of the coil under installation at the supplier's premises in the USA. The Central Solenoid includes 6 identical Nb3Sn coil modules independently powered and enclosed inside a precompression structure preventing their separation. The CS structure includes 9 subsets, made of Nitronic 50 high strength austenitic stainless steel, evenly distributed around the stack of the 6 modules.

The ITER Poloidal Field (PF) winding pack consists of a stack of double pancakes made of NbTi CIC conductor. A structural element is required, at each electrical joint, to react the operational hoop load. In the present design, this is provided by a hollow profiled steel part welded to the conductor jacket and bonded to the adjacent turns, named "coil-tail". As part of the present development, prototype coil-tails have been manufactured and welded to the PF conductor jacket. A full-size mechanical mock-up, in the shape of a straight beam has been manufactured. It is made of coil-tails and steel plates simulating the adjacent conductors, which are insulated and impregnated. The mock-up is to be subjected soon to fatigue tests at the ENEA-Brasimone laboratory at LN2 temperature

The ATLAS B0 model coil has been extensively tested, reproducing the operational conditions of the final ATLAS barrel toroid coils (ten Kate, 1999). Two test campaigns have taken place on B0, at the CERN facility where the individual BT coils are about to be tested. The first campaign aimed to test the cool-down, warm-up phases and to commission the coil up to its nominal current of 20.5 kA, reproducing Lorentz forces similar to the ones on the BT coil. The second campaign aimed to evaluate the margins above the nominal conditions. The B0 was tested up to 24 kA and specific tests were performed to assess: the coil temperature margin with respect to the design value, the performance of the double pancake internal joints, static and dynamic heat loads, behavior of the coil under quench conditions. The paper reviews the overall test program with emphasis on second campaign results not covered before.

The 9-m long superconducting model coil B0 was built to verify design parameters and exercise the construction of the Barrel Toroid magnet of ATLAS Detector. The model coil has been successfully tested at CERN. An intensive test program to study quench propagation through the coil windings as well as the temperature distribution has been carried out. The coil is well equipped with pickup coils, voltage taps, superconducting quench detectors and temperature sensors. The current is applied up to 24 kA and about forty quenches have been induced by firing internal heaters. Characteristic numbers at full current of 24 kA are a normal zone propagation of 15 m/s in the conductor leading to a turn-to-turn propagation of 0.1 m/s, the entire coil in normal state within 5.5 s and a safe peak temperature in the windings of 85 K. The paper summarizes the quench performance of the B0 coil. Based on this experience the full-size coils are now under construction and first test results are awaited by early 2004. 7 Refs.

ATLAS is one of the two major experiments of the LHC project at CERN using cryogenics. The superconducting magnet system of ATLAS is composed of the barrel toroid (BT), two end caps toroids and the Central Solenoid. The BT is formed of 8 race-track superconducting dipoles, each one 25 m long and 5 m wide. A reduced scale prototype (named B0) of one of the 8 dipoles, about one third of the length, has been constructed and tested in a dedicated cryogenic facility at CERN. To simulate the final thermal and hydraulic operating conditions, the B0 was cooled by a forced flow of 4.5 K saturated liquid helium provided by a centrifugal pump of 80 g/s nominal capacity. Both static and dynamic heat loads, generated by the induced currents on the B0 casing during a slow dump or a ramp up, have been measured to verify the expected thermal budget of the entire BT. The instrument used for the heat load measurements was a void fraction meter (VFM) installed on the magnet return line. The instrument constructed at CERN was calibrated in order to provide direct readings of heat loads. An example of application of the VFM measuring method to a large-scale apparatus cooled at liquid helium temperature.

The International Thermonuclear Experimental Reactor (ITER) toroidal field model coil (TFMC) was designed to allow an overall mechanical test representative of the ITER toroidal field (TF) coils design principles and features. The coil, manufactured by the European industry, was tested in two phases at the TOSKA facility of the Forschungszentrum Karlsruhe up to its maximum current of 80 kA. In the single coil tests performed in 2001 the coil was submitted to in-plane loading only, whereas in the two coil tests performed in 2002 the coil experienced also out-of-plane loading, representative of the coil load conditions in the ITER tokamak TF magnet. The paper details the mechanical tests performed, compares them to model predictions and discusses the experience gained in the mechanical behavior of the ITER TF coils.

In 1997, the JET device was operated for an extensive campaign with deuterium–tritium (D–T) plasmas (the DTE1 campaign). A comprehensive network of machine protection systems was necessary so that this experimental campaign could be executed safely without damage to the machine or release of activated material. This network had been developed over many years of JET deuterium plasma operation and therefore the modifications for D–T operation was not a significant problem. The DTE1 campaign was executed successfully and safely and the machine protection systems proved reliable and robust and, in the limited cases where they were required to act, functioned correctly. The machine protection systems at JET are described and their categorisation and development over time are summarised. The management, commissioning and operational experience during DTE1 are discussed and some examples of fault scenarios are described. The experience with protection systems at JET highlights the importance of correct design and philosophy decisions being taken at an early stage. It is shown that this experience will be invaluable data input to the safe operation of future large fusion machines.

All major JET systems have been fully commissioned for D-T and the DTE1 series of experiments has started with the D-T fuel mixture and operating conditions foreseen for ITER. In the area of ITER physics, significant results have been produced in both D-D and D-T. In D-D, the LH threshold power database has been extended, the bounds on edge-electron temperature and density in ELMy H-modes have been defined and the advantages of Types I and III ELMy discharges have been compared. In D-T plasmas, the isotope effect on H-mode threshold power and transport has been determined so that a more accurate assessment can be made of the ignition margin and heating requirements for ITER. Trace tritium experiments have provided first particle transport measurements and an assessment of the ITER reference ion-cyclotron resonance-frequency heating scenarios has been started, In the area of fusion performance, record D-D neutron yields have been obtained by controlling the plasma and current profiles in hot ion ELM-free H-modes and optimized shear modes. In D-T, internal transport barriers have been readily established in optimized shear discharges and Alfven eigenmodes have been observed.

The Central Solenoid (CS) is a critical component in the ITER tokamak providing plasma current drive and shaping. The CS final design is being completed at the US ITER Project Office (USIPO) in Oak Ridge, TN under a Procurement Arrangement with the ITER Organization (IO). Key design decisions have been made and CAD models and drawings developed. Interfaces have been established. An extensive R&D program has been completed. Analyses have been conducted to verify the design meets requirements. Design documentation is being completed in anticipation of a Final Design Review in the fall of 2013. The paper describes the key features of the CS final design.

The JET neutral beam injection system has proved to be both effective and reliable as a plasma heating device. The ion heating and plasma fuelling characteristics of the system are ideally suited to the production of high fusion performance plasmas while the flexibility in the choice of beam species (H, D, T, 3 He or 4 He) and the ability to inject into almost any JET plasma configuration allows a wide variety of related physics experiments to be carried out. The capability to inject (for the first time) tritium beams was essential to the successful execution of the first tritium experiments in which 1.7 MW of power from DT fusion reactions was generated.

Three of the toroidal field (TF) coils of the JET tokamak have developed interturn faults. These faults have not been catastrophic and it has been possible to continue operation and replace the faulty coils with spares at the next pre-planned shutdown. The faults were found to be due to water leaks. The coil coolant was changed from water to an insulating fluid. All known faulty coils have been changed and latest measurements do not detect any faults on the installed coils. >

An account is given of the high performance plasmas established by development of the H-mode regime in JET in the experimental campaigns up to 1992. High performance in this case is measured in terms of the confinement enhancement achieved over the L-mode scaling as measured using the plasma diamagnetism. Three JET H-mode regimes have achieved enhancement factors (H G DIA ) over Goldston L-mode scaling of 2.5 < H G DIA < 4.0. These are the Pellet Enhanced Performance (PEP) H-MODE, the high bootstrap fraction (high β POL ) H-mode and the Hot Ion (HI) H-mode. The phenomenology of these three regimes is reviewed and contrasts and common threads are elucidated

The main results of the Tore Supra experimental programme in the years 2007–2008 are reported. They document significant progress achieved in the domain of steady-state tokamak research, as well as in more general issues relevant for ITER and for fusion physics research. Three areas are covered: ITER relevant technology developments and tests in a real machine environment, tokamak operational issues for high power and long pulses, and fusion plasma physics. Results presented in this paper include test and validation of a new, load-resilient concept of ion cycotron resonance heating antenna and of an inspection robot operated under ultra-high vacuum and high temperature conditions; an extensive experimental campaign (5 h of plasma) aiming at deuterium inventory and carbon migration studies; real-time control of sawteeth by electron cyclotron current drive in the presence of fast ion tails; ECRH-assisted plasma start-up studies; dimensionless scalings of transport and turbulence; transport experiments using active perturbation methods; resistive and fast-particle driven MHD studies. The potential role of Tore Supra in the worldwide fusion programme before the start of ITER operation is also discussed.

JT-60SA is a fully superconducting coil tokamak to be built under the framework of the EU-JA Broader Approach Agreement, and it aims to contribute to the complement of the ITER experiments and to the DEMO reactor design by the study of steady-state high-beta plasma experiments. The conceptual design of the JT-60SA tokamak and the peripheral systems has been carried out in close collaboration of EU and Japan from middle of 2006, and the JT-60SA Conceptual Design Report which involves an analysis of plasma physics and engineering design is summarized in May 2007. This paper intends to present an overall view of the JT-60SA project, and the latest design status of the key engineering issues such as superconducting magnets, divertor target, plasma heating devices, and magnet power supplies.

The ATLAS B0 model coil has been tested at CERN to verify the design parameters of the Barrel Toroid coils (BT). The mechanical behavior of the B0 superconducting coil and its support structure is reported and compared with coil design calculations. The mechanical stresses and structural force levels during cooling down and excitation phases were monitored using strain gauges, position sensors and capacitive force transducers instrumentation. In the ATLAS magnet test facility, a magnetic mirror is used to reproduce the electromagnetic forces present in the BT coils, once these are assembled in toroid in the underground cavern in 2004. (8 refs).

The JET tokamak magnetic system has been enhanced by adding four resistive magnets inside the vacuum vessel to produce divertor plasmas. In factory the coil parts were manufactured and process techniques qualified. Assembly took place at JET, inside the vacuum vessel, including welding into Inconel casings and impregnation with epoxy resin.

The ATLAS Barrel Toroid (BT) consists of 8 superconducting coils of 25 m length and 5 m width, it provides the magnetic field for the muon spectrometer of the ATLAS detector, one of the experiments of the Large Hadron Collider, under construction at CERN. The cable used for the coils is an Aluminum stabilized NbTi rutherford cable and the coils are built with the double pancake technique, so a resistive joint between the cables of the two single pancake is present. The transverse magnetic field induces eddy currents, during the ramping of the main current, whose intensity can sometimes exceed the transition limit. B0 is a one third length, full large scale model of one of the eight magnet composing the ATLAS barrel toroid. The B0 construction was decided to test the technical construction solutions and reproduce the behaviour of the final coils. In this paper the behaviour of the interpancake joint is analyzed and compared with theoretical models describing the current distribution at the joint. Measurements of the joint resistance are reported.

The authors report on the preparation and technical performance of experiments on the JET (Joint European Torus) tokamak aimed at producing one cycle of alternating plasma current. After analyzing the revised scenario of the poloidal field circuit in relation to the different phases of the plasma current discharge they report in detail on the modification of the JET poloidal field system. Preliminary experiments successfully demonstrated, at large plasma current, one full cycle of an AC tokamak operation. The two half cycles are characterized by the same Z/sub eff/ and the same radiated power at the same density, and the same reaction rate for the same ICRH (ion cyclotron resonance heating) power. A short dwell time, primarily determined by the vertical stray field and the prefill pressure, was demonstrated. The experiment indicates that the duration of the plasma current reversal phase in this mode of operation is dominated by the plasma current rampdown and the plasma current rise. >

The ATLAS superconducting magnet system consists of a Barrel Toroid, two End-Cap Toroids and a Central Solenoid. The Barrel Toroid, with overall dimensions of 20-m diameter by 26-m length, is made of eight individual coils symmetrically assembled around the central axis with a warm structure. The system is presently under construction in industry. In order to verify the construction concepts a model coil B0, a 9-m short version of a single Barrel Toroid coil, was built. Since April 2001, an extensive test program is underway at CERN to characterize the mechanical, thermal, electrical and magnetic properties of the coil. The magnet successfully achieved the 20-kA nominal operating current in July 2001. The test program and the main results are reported. (9 refs).

Publisher Summary The aim of coil protection system is to detect electrical faults and to protect the coils against mechanical or thermal over-stressing. The protection implemented are over-voltage and over-current for all the circuits, circuit equation simulation and comparison with the measured currents, ampere turn protection, and tensile, shear, and thermal stresses of poloidal and divertor coils. The protective actions include immediate removal of the voltage from the coils and circuit breaker trip. The tensile and shear stress of the poloidal coils and divertor coils is computed as a linear combination of the vertical force, radial force, and energy dissipated. The radial and vertical force of each coil is computed with flux loops and ampere turn measurements. During pulses, the TF coils expand in the radial direction and in the vertical directions. This motion can be approximated, on a slow time scale, by a linear combination of the in-plane magnetic force — due to the interaction of the current with the toroidal field — and the dissipated energy.

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.

The ITER Magnet system [1] consists of four main coils sub-systems: 18 Toroidal Field Coils (TF-coil), a Central Solenoid (CS), 6 Poloidal Field Coils (PF-coil) and 18 Correction Coils (EFCC). The main contract of the EFCCs supply is awarded to the Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) by the Chinese Domestic Agency (CNDA). According to the pre-qualification program, ASIPP is implementing the procurement phase to qualify and validate key technologies and manufacturing methods. The Correction coils qualification activities are conducted within the framework of the procurement arrangement set up between the ITER Organization and CNDA. The paper describes the CC development including first results of the coils winding qualification trials and, qualification of a S-Glass fiber-polyimide based insulation system The CC casing assembly process and the first results of the welding trials are reported. The weld qualification, according to ASTM for 316LN austenitic steel is reported in terms of fracture toughness, fatigue crack growth, and tensile property at 4 K. The Vacuum Pressure Impregnation of CC short mock-up, with low viscosity bisphenol-F (DGEBF) epoxy resin, aims to optimization of the curing and insulation mechanical properties.

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.

The superconducting magnet system of ITER has ~5,000 tons of cold mass as steel structures to support the gravity and electromagnetic forces. These steel structures need to be maintained at ~4.4 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 ~2.7 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.

JET was extensively modified in the 1992/93 shutdown. The new pumped divertor and many new systems were brought into operation early in 1994. Operations have progressed to 4MA plasma current and, with substantial additional heating, H-mode confinement results confirm the expected scaling. The high power handling capability of the pumped divertor with sweeping is estimated at 20MW for 20s. H-mode plasmas have large Type I ELMs. With lower hybrid heating alone, 2MA full current drive has been achieved with good efficiency, with ICRF power, effective heating and direct electron heating have been demonstrated.

The ATLAS B0 model coil has been developed and constructed to verify the design parameters and the manufacture techniques of the Barrel Toroid coils (BT) that are under construction for the ATLAS Detector. Essential for successful operation is the mechanical behavior of the superconducting coil and its support structure. In the ATLAS magnet test facility, a magnetic mirror is used to reproduce in the model coil the electromagnetic forces of the BT coils when assembled in the final Barrel Toroid magnet system. The model coil is extensively equipped with mechanical instrumentation to monitor stresses and force levels as well as contraction during a cooling down and excitation up to nominal current. The installed set up of strain gauges, position sensors and capacitive force transducers is presented. Moreover the first mechanical results in terms of expected main stress, strain and deformation values are presented based on detailed mechanical analysis of the design. (7 refs).

The ITER toroidal field model coil (TFMC) was designed, constructed and tested by the European Home Team in the framework of the ITER research and development program of the Engineering Design Activities (EDA). The project was performed under the leadership of European Fusion Development Activity/Close Support Unit (EFDA/CSU), Ciarching. in collaboration with the European superconductor laboratories and the European industry. The TFMC wits developed and constructed in collaboration with the European industry consortium (AGAN) and Europa Metalli LMI supplied the conductor, The TFMC was tested in the test phase I as single coil and in phase 11 in the background field of the EURATOM LCT coil in the TOSKA facility of the Forschungszentrum Karlsruhe. In phase 1, the TFMC achieved an ITER TF coil relevant current of about 80kA and further representative test results before the end of the EDA. In the more complex test phase [I. the coil was exposed to ITER TF coil relevant mechanical stresses in the winding pack and case. The tests confirmed that engineering design principles and manufacturing procedures are sound and suitable for the ITER TF full size coils. The electromagnetic. thermo hydraulic and mechanical operation parameters agree well with predictions. The achieved Lorentz force on the conductor was about 800 kN/m. That has been equivalent to the Lorentz forces in ITER TF coils. (c) 2005 Elsevier B.V. All fights reserved.