Fusion for Energy is delivering 5 out of 6 ITER superconducting poloidal field coils (PF coils), which are composed of stacks of 6 to 8 double-layered circular coils - double pancakes (DPs). The double pancakes range from 17 m to 24 m in diameter and the wound conductor has a NbTi core inside a stainless steel squared jacked. Due to the size of the PF 2-4 coils, they have been manufactured on the ITER site, very close to the assembly hall. This article highlights the manufacturing processes, the learning curve, the main challenges and learned lessons during the winding activities of the 30 double pancakes that comprise 4 PF coils.

ObjectivesThe helium cooling tube is the core component of the fusion superconducting magnet system. It connects the superconducting magnet cooling channel and the cryogenic system, and is directly related to the operation stability of the superconducting magnet system. Since helium cooling pipes need to be welded on site, the structural design of helium pipes needs to be comprehensively considered from multiple aspects such as functional realization and on-site process operability, and reasonable solutions need to be proposed.MethodsThrough comprehensive analysis of the polar field structure of the fusion reactor, the structural optimization and pressure loss assessment of the cryogenic helium cooling tube were carried out. A comprehensive comparative analysis of the helium hole size and the optimal design of weld chamfer was carried out from the perspective of structural mechanics. The general requirements for the design criteria of the fusion reactor magnet helium cooling tube were proposed. Based on the structural mechanical analysis and pressure loss analysis of helium cooling tubes, and the comprehensive consideration of field operability, a helium cooling tube design scheme was proposed to meet the requirements of helium cooling tubes for fusion reactor magnets.ResultsAlthough short running helium cooling pipes will cause an increase in local pressure loss, which is equivalent to a conductor of 2.7 meters long, the pressure loss caused by them is basically negligible compared with the total length of the coil.ConclusionsThe adoption of the short track helium tube design scheme can not only meet the functional requirements, but also greatly improve the operability of the field process and meet the requirements of helium cooling tubes for superconducting magnets in fusion reactors.

Fusion for Energy (F4E), the European Domestic Agency for the International Thermonuclear Experimental Reactor (ITER) is responsible for the manufacturing, test and delivery of 5 out of 6 Poloidal Field Coils of ITER. F4E has currently delivered 4 coils: PF2, PF4, PF5 and PF6, the last two are installed in the Tokamak pit in temporary positions while PF2 and PF4 are stored awaiting the Tokamak torus completion. At the PF coils manufacturing production site in Cadarache, France, few meters away from the ITER Tokamak assembly building, F4E is currently finalizing the manufacturing and testing of the largest coil: PF3, with 24 m diameter and a weight of 384 t , approaching to the end of an endeavor that started back in 2013. This article describes all electrical tests the PF coils underwent to meet the ITER technical requirements during their lengthy manufacturing process, and which lasted several years. Tests performed at key points of the manufacturing process avoided discovering non-conformities at later stages that could otherwise become critical for the project accomplition. The main electrical tests consisted of high voltage DC/AC, partial discharge tests, and local and global Paschen tests. The experimental setups and procedures are presented and discussed in more detail. We also summarize the acceptance test results for the finalized coils i.e., 80 K forced flow pressure drop testing, leak tightness in vacuum before, during and after the thermal cycle to 80 K .

ITER envisages the use of two heating neutral beam injectors plus an optional one as part of the auxiliary heating and current drive system. The 16.5 MW expected neutral beam power per injector is several notches higher than worldwide existing facilities. A Neutral Beam Test Facility (NBTF) was established at Consorzio RFX, exploiting the synergy of two test beds, SPIDER and MITICA. SPIDER is dedicated to developing and characterizing large efficient negative ion sources at relevant parameters in ITER-like conditions: source and accelerator located in the same vacuum where the beam propagates, immunity to electromagnetic interferences of multiple radio-frequency (RF) antennas, avoidance of RF-induced discharges on the outside of the source. Three years of experiments on SPIDER have addressed to the necessary design modifications to enable full performances. The source is presently under a long shut-down phase to incorporate learnings from the experimental campaign. Parallelly, developments on MITICA, the full-scale prototype of the ITER NBI featuring a 1 MV accelerator and ion neutralization, are underway including manufacturing of in-vessel components, while power supplies and auxiliary plants are already under final testing and commissioning. Integration, commissioning and tests of the 1MV power supplies are essential for this first-of-kind system, unparalleled both in research and industry field. The integrated test to confirm 1MV output by combining invertor systems, DC generators and transmission lines extracted errors/accidents in some components. To realize a concrete system for ITER, solutions for the repair and the improvement of the system were developed. Hence, NBTF is emerging as a necessary facility, due to the large gap with existing injectors, effectively dedicated to identify issues and find solutions to enable successful ITER NBI operations in a time bound fashion.

The MITICA (Megavolt ITER Injector and Concept Advancement) experiment, currently in the installation and commissioning phase at the Neutral Beam Test Facility in Padua-Italy, will be equipped with a Beam Source (BS) expected to generate a 40A negative ion beam to the energy of 1MeV (values for D). It is composed of an ion source to generate a H or D plasma and of a multiple-Aperture multiple-grid (MAMuG) Accelerator [1], where ions extracted are accelerated in five consecutive stages of 200 kV each. The engineering and fabrication of the beam source for MITICA BS started in October 2018, when a contract between Fusion for Energy (F4E) and Alsymex (France) was signed. Presently, all the components and sub-assemblies have been manufactured, except for a few non-critical items. Assembly of the accelerator is well advanced and ion source is on-going. A review of the fabrication of the main components and aspects of their assembly is reported in this paper, with a focus on the main achievements.

Fusion for Energy (F4E), the European Domestic Agency for the International Thermonuclear Experimental Reactor (ITER), is responsible for the supply of 5 out of the 6 Poloidal Field (PF) Coils: PF2-PF6. One coil, PF1, is being manufactured by the Russian Federation Domestic Agency (RFDA). While the 10 m diameter PF6 was manufactured by the Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) and tested in the cold test facility at Cadarache under a collaboration agreement with F4E; coils PF2-PF5 are currently being manufactured on site, close to the Tokamak building, their size ranging from 17 to 24 m diameter and weights from 200 to 400 T. This article describes the final acceptance tests performed on the coils PF5 and PF6, the testing setup, paying special attention to the tests performed before, after and during the cool-down at 80 K. The tests cover a wide range of aspects of the operation at cryogenic temperatures: ranging from the high voltage electrical insulation performance during the potential fault conditions during plasma operation, leak tightness under vacuum and pressure drop behavior of its hydraulic system during operation with forced flow helium. In addition, we will briefly introduce the results for the current center line (CCL) calculation obtained for PF5 and PF6.

Three out of six Poloidal Field Coils (PFC) are already delivered to the ITER Organization. The PF Coils are built winding, impregnating, and vertically stacking double pancakes (DPs) of NbTi Cable-in-Conduit conductors into a Winding Pack (WP). Later, the Winding Pack (WP) is impregnated for ground insulation and clamping devices are installed for structural support and interface with the rest of the ITER machine. One of the main parameters characterizing the PFCs is the Current Centre Line (CCL), defined as the barycentre of its WP conductors. Ideally, the CCL would be in the WP's symmetry plane but due to solutions in the construction design and manufacturing deviations, it may vary. Double Pancakes (DPs) may be wound with different dimensions, or a deviation during their stacking would cause a misalignment of all the conductors contained in that ill-positioned DP, affecting the CCL. The manufacturing process starts with the conductor winding and insulation, forming a DP. The DP is then impregnated with resin and a scan of its surface is used to reconstruct the DPs virtually. The DPs are then stacked, forming a WP, and measurements of points on the surface are used to recreate the process virtually. Finally, the WP is insulated and impregnated with resin. At this stage, both surface scans and point measurements are used to align the WP in the coordinate system of the ITER machine. This paper explains the process to calculate the CCL of the three first PF Coils using manufacturing data, defining the uncertainty associated with the calculation and comparing against the target tolerances defined for the proper ITER machine operation.

The helium inlet is one of the most important components of the ITER Poloidal Field (PF) coils. The insulation structure of the helium inlet is critical to provide sufficient electrical and mechanical properties for a practical application. In this paper, an ITER PF6 coil double pancake helium inlet trial mock-up was designed and manufactured by simulating the actual manufacturing process. A thermal cycling test on the sample was carried out from 280 K to77 K, also before and after the thermal cycling, a turn-to-turn DC HV test and Paschen test were performed. The test results satisfied the ITER requirements. It has therefore been verified that the PF6 coil double pancake helium inlet insulation has good electrical properties, and can be applied to the formal production of the PF6 coil double pancakes.

The plasma confinement of the International Tokamak Experimental Rector (ITER) is provided by the magnetic field generated by 18 toroidal field (TF) coils while 6 poloidal field (PF) and 6 central solenoid coils have the function to drive, shape and pre-heat the plasma. Fusion for Energy (F4E), the European Domestic Agency for ITER, is responsible for the supply of 10 TF coils and 5 PF coils to the ITER project. The ITER Organization (IO) team is instead responsible for the design of such coils as well for the coordination of the activities of the different Domestic Agencies (DAs) producing the different components, and their assembly into the Tokamak. The PF coils utilize NbTi Cable-in-Conduit-Conductor and have different diameters between 8 and 24 meters and weights of up to 400 tons. Regarding the PF coils produced by F4E, so far one has been completed by the Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) under a collaboration agreement with F4E. The other 4 PF coils are being produced at the ITER site in Saint Paul lez Durance, France, under F4E supervision. The first of these (PF5) will be completed by July 2020 while the last coil (PF3) will be ready be the end of 2023. The TF coils utilize Nb3Sn conductor and are manufactured with the “Wind, React & Transfer” method. The first TF coil is close to completion and will be delivered to the ITER site in early 2020. Subsequent TF coils will follow at a rate of about one every 3-4 months. In this article we will report on the production status of both PF and TF coils and, in particular, the different manufacturing strategies employed. The main challenges faced so far and the results obtained are also described.

The International Thermonuclear Experimental Reactor (ITER) is a full superconducting coil tokamak. The tail is a critical element of the bottom or top double pancakes of the Poloidal Field (PF) coil. The main functions of the tail are to withstand the electromagnetic force of the conductor. The electrical isolation is provided by the strap. The connections of the tail to the last but one turn and conductor terminal are both carried out through the welds. Moreover, this welding is further complicated as it is performed in overhead position due to the narrow operational space, one welding procedure should be developed and qualified to satisfy the ITER requirements. This paper focuses on the assessment and qualification of the weld for PF6 coil tail. One new special J-style groove with the thickness of 10 mm was designed and two full-size samples were prepared by simulating the actual working condition. Meanwhile, one welding procedure was developed and applied to control the weld deformation and welding temperature, etc. Finally, the sample 1 was destructively inspected in two sections after a five cycle thermal test, and no obvious defects appeared. In addition, a fatigue test of sample 2 was carried out with a load range of (60-600) kN at 4 Hz and 77 K. Finally, sample 2 successfully passed 30 000 cycles, validating that the fatigue properties of the PF6 coil tail weld satisfy the ITER requirement for the PF6 coil production.

The Poloidal Field (PF) coils are one of the main sub-systems of the ITER magnets. The PF6 coil is being manufactured by the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) as per the Poloidal Field coils cooperation agreement signed between ASIPP and Fusion for Energy (F4E). ITER PF6 winding pack (WP) is composed by 9 double pancakes (DPs). After the vacuum pressure impregnation (VPI) of each DP, the 9 DPs were stacked together and inter-pancake joints were manufactured to form the WP. This paper focus on the main process of ITER PF6 DPs stacking. During stacking, the DP being stacked was firstly been placed in position by means of vacuum pads lifting system, laser tracker and stacking guiding tooling. To eliminate the thermal stress, the DP being stacked was then lifted up to 80-90% of its weight to allow free rotating and moving while inter-pancake joints copper sole soldering. At the end, the DP being stacked was released, followed by welding of lateral plates and wedges. The DPs stacking has been finished with an overall concentricity of less than 1 mm and toroidal positioning of less than 5 mm for each DP, which well met the technical requirements.

The Poloidal Field (PF) coils are one of the main sub-systems of the ITER magnets. The PF6 coil is being manufactured by the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) as per the Poloidal Field coils cooperation agreement signed between ASIPP and Fusion for Energy (F4E). As one of the critical components, helium inlet locates on the innermost turn and supplies the coil with supercritical helium. During Tokamak operation, the helium inlet will undergo huge cyclic electromagnetic loads and thermal cycling during cooling down and warming up. This paper focus on the main steps of ITER PF6 helium inlet qualification. Helium inlet hole drilling and stainless steel wrapping removal were firstly been carried out. Helium inlet welding with full penetration by automatic welding machine was then performed, followed by leak proof test, non-destructive test, fatigue test and tomography test. At the end, the samples were sectioned for micro and macro inspection. The results show ITER PF6 helium inlet qualification has successfully met the requirements of PF procurement agreement(PA) and was approved by ITER IO.

The ITER PF6 coil is manufactured by winding, insulating the superconducting conductors, and further connecting them electrically and hydraulically. To qualify the insulation design and manufacturing process to be applied to the ITER PF6 coil, two beam-shaped mock-ups taking the form of a 3 x 3 array of conductors combined with the same insulation as in the production coil have been tested in simulated Tokamak operating conditions. All the results of the electrical and mechanical tests, including DC and AC high-voltage withstanding at both room and cryogenic temperature, AC partial discharge, Paschen tests, cryogenic compressive fatigue, push-out, and thermal cycling, met the ITER requirements. Performance of the mock-ups confirmed that the insulation design and manufacturing process are suitable for the ITER PF6 coil, based on that the qualification of the insulation manufacturing process for the double pancake can be started.

International Fusion Energy Organization (ITER) is one of the most ambitious energy projects in the world today. The poloidal field (PF) coil is cooled down to the required operating temperature (4.2 K) by a forced flow of supercritical helium, which is introduced through the helium inlet. The mechanical reliability requirement on the irregular structure of helium inlet due to the introduction of orifice make the helium inlet one of the critical components of the PF coil cooling system. This paper focuses on the fatigue tests on ITER PF6 coil full-size helium inlet samples to validate whether the fatigue performance of the conduit with the welded inlet can withstand the lifetime of PF6 coil close to the working condition. According to the ITER requirement, the conductor jacket shall withstand a cyclic longitudinal strain (0.95 to 9.5) × 10−4 during 600 000 cycles at 77K or (1.9–19) × 10−4 during 30 000 cycles at 77 K, and the equivalent nominal tensile load range we calculated is about (33.5–335) kN. Two samples were designed and prepared to assess the structural design and weld quality. The fatigue test of the sample 1 was carried out with a load range of (33.5–335) kN at 4 Hz and 77 K, and it failed after 593 000 cycles. The scanning electron microscope images and finite element analysis showed that crack initiation area of the fractured sample 1 was at the bottom of the hole. The main reasons were that there were some stress concentrations inside the hole when the tensile load was provided, and the residual burrs assisted the crack initiation. Finally, the modified sample 2 with the completely removed burrs succeed after 600 000 cycles with a nominal load range of (33.5–335) kN. Therefore, it is concluded that the present optimized ITER PF6 helium inlet sample has good fatigue performance and meets the requirements of ITER, which can be utilized for practical applications.

The poloidal field (PF) coils are one main subsystem of the ITER magnets. The PF6 coil is being manufactured by the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) as per the PF coils cooperation agreement between ASIPP and Fusion for Energy. The ITER PF6 winding pack is composed by stacking nine double pancakes. Dummy double pancake fabrication aims to fully qualify the critical fabrication processes, before series production. This paper describes ITER PF6 dummy double pancake winding process and results, which was performed with a “two-in hand” configuration. Conductors were wound sequentially in the following steps: despooling, straightening, ultrasonic cleaning, sandblasting, bending, and turn insulation wrapping until deposition on the winding table. Joggle bending, helium inlet manufacturing, and tail assembly were also carried out during winding. The dummy double pancake winding was finally completed with the qualification requirements well met, which demonstrated good synchronization of the control system, fine conductor positioning accuracy, tight radial buildup, and so on.

International Thermonuclear Experimental Reactor (ITER) is a full superconducting coil tokamak. The strap composed of fiber glass composite materials is an important component of Poloidal Field (PF) coil tail. The main function of the strap is to maintain the electrical isolation and mechanical connection between the last turn and the last but one turn. The thermal contraction of the strap shall be measured to assess the mismatch between the strap and adjacent conductor while the coil is cooled down, and further, in order to balance the thermal stresses, pretension on the strap is needed while the elastic modulus of the full-size strap shall also be measured. In addition, the fatigue performance of the strap shall be examined in consideration of the cyclic electromagnetic load on the tail in the tokamak operation condition. In this paper, the thermal contraction test, the elastic modulus test and the fatigue test on the full-size strap for ITER PF coils were carried out. The test results show that the thermal contraction from 300 K to 77 K is 0.19% and the elastic modules is about 60 GPa at 77 K. The fatigue results indicate that the strap can meet the requirementof ITER and can be utilized to the practical applications. (C) 2017 Elsevier B.V. All rights reserved.

International Thermonuclear Experimental Reactor (ITER) is a full superconducting coil tokamak. The tail is an important component of Poloidal Field (PF) coil, of which the main functions are to provide the electrical isolation and transfer the longitudinal load from the last turn to the last-but-one turn. The paper focuses on an optimized mechanical structure of PF6 coil tail, which is made up of two main parts. One was welded to the last turn and the other was welded to the last-but-one turn. Both of them were connected by the mechanical coupling. The electrical isolation between the two parts was maintained by a strap made of insulating composite. In addition, as the PF6 coil is operated under the cyclic electromagnetic load during the tokamak operation, the fatigue property of the tail should be assessed and qualified at low temperature. Moreover, taking into consideration the complexity of the insulation winding process which is performed in a confined space, the wrapping process of the insulation needs to be established. Meanwhile, the high voltage (HV) tests of the tail insulation, including the direct current (DC) and alternating current (AC) tests, need to be assessed before and after the fatigue test. In this paper, a fully bonded PF6 coil tail partial mock-up (not including the weld of the tail to the last conductor turn) was designed and manufactured by simulating the actual manufacturing processes. In addition, the fatigue tests on the sample were carried out at 77 K, and the results showed the sample had good and stable fatigue properties at cryogenic temperature. The HV tests before and after the fatigue test, also including the final 30 kV breakdown DC test after the fatigue test, were carried out. The test results satisfied the requirements of ITER and were discussed in depth. Finally, the sample was destructively inspected to validate the integrity of the insulation by mechanical cross sectioning, and no voids and cracks were observed. Therefore it can be verified from the test results that the designed PF6 coil tail has good comprehensive properties, which can be applied to the formal production of the PF6 coil.

The European Union contributes around 20% of the cable-in-conduit conductor lengths needed for the ITER toroidal field (TF) magnet coils. For that purpose, 97 tons of Nb 3 Sn superconducting strand have been fabricated over five years, the production being completed in 2014. This superconducting strand has been manufactured by two companies, namely, Bruker EAS (Germany) and OST (USA), through the bronze route and the internal tin diffusion, respectively. This paper reports the outcomes of this strand mass production and of the strand characterization as performed by the suppliers and cross-checked on a regular basis by Durham University.

The European share of the ITER magnet superconductors includes the supply of around 20 km of Toroidal Field and 7 km of Poloidal Field conductors. This represents 20% and 11% of the total conductor amounts respectively needed for all ITER Toroidal Field (TF) and Poloidal Field (PF) coils. For TF conductor, around 97 tons of superconducting niobium tin strand and 60 tons of copper wire are needed to be purchased, cabled and inserted within a stainless steel jacket tube to form the Cable-In-Conduit-Conductor (CICC). For PF conductor, about 45 tons of superconducting multifilamentary niobium titanium wire need to be cabled and jacketed. Altogether, Fusion for Energy (F4E) placed 4 supply contracts for the European TF and PF coil conductors (one for copper strand, two for niobium tin strand and a single contract for cabling and jacketing). In addition, the PF conductor procurement is based on a bi-lateral agreement with the Russian Federation Domestic Agency (RF DA) where NbTi cables are supplied by RF DA and their jacketing is provided by F4E. In this article, the procurement strategy and the current procurement status are reported for the European contribution of TF and PF conductors.

This paper presents the results of the latest European Conductor Performance Qualification Sample for the ITER toroidal field coils. The standard qualification test was performed in the SULTAN facility at CRPP. The sample consisted of two 3.5-m lengths of cable in conduit conductor (CICC), which utilized bronze route Nb 3 Sn strands from Bruker-EAS. The sample was prepared at CRPP with the conductor terminations and the bottom joint fabricated by the solder filling method. In order to prevent slippage between the jacket and cable, SULTAN samples systematically have crimping rings added to the extremity of each leg. In this case, one leg also had additional crimping rings on both sides of the high field region, which experiences a greater electromagnetic load during the test. As per a standard SULTAN sample, temperature sensors and voltage taps were installed along each conductor. In order to monitor the strain effects of the electromagnetic load cycles on the sample, strain gauges were bonded to each leg both in the high and low field regions. Pickup coils were wound on the high field region of each leg for critical temperature measurements via the magnetization method. The results of the SULTAN test are described in this paper and the effect of the additional crimping rings in the high field region discussed. Additional investigative work into residual jacket strain of CICCs is also presented.