Superconducting (SC) tokamak JT-60SA plays an essential role in fusion research and development by supporting and complementing the ITER project, providing directions to the DEMO design activity and fostering next generation scientists and engineers. Since the short circuit incident at the terminal joints of equilibrium field coil #1 during the integrated commissioning (IC) in March 2021, both EU and JA implementing agencies (IAs) have examined how to ensure safe operation of JT-60SA by mitigating the risk of possible discharge occurrence inside the cryostat. Based on the experience of the global Paschen tests, the IAs have established a strategy of risk mitigation measures, which is a combination of (i) reinforcement of insulation, (ii) avoiding unnecessary voltage application to the coil systems and (iii) immediate de-energization of the coils when deteriorated vacuum conditions are detected. Thanks to the considerable efforts of the Integrated Project Team members, the IC restarted in May 2023. After confirmation of the SC state of the coil systems (TF, EF and CS), the coil energization test and the plasma operation phase 1 (OP-1) started. The first plasma was successfully achieved on 23 October 2023 with a limited value of voltage and current applied to the coils. The plasma configuration control was also confirmed with low plasma current and low auxiliary heating power conditions. Based on the IO-F4E-QST collaboration, activities of JT-60SA have been shared with the IO and provided an important lesson for ITER assembly and commissioning, and will provide an outstanding contribution to fusion research at large. After OP-1, maintenance & enhancement phase 1 (M/E-1) starts from January 2024, in which in-vessel components are installed, and heating and diagnostic systems are extensively upgraded to allow a high power heating experiment planned in OP-2. In order to make the best use of JT-60SA, a newly organized JT-60SA experiment team will refine the research plan for the future high heating power operation phase.

This paper describes critical assembly process qualifications, acceptance criteria, and worker training of ITER Central Solenoid Assembly. Coaxial joint assembly was qualified and trained to achieve resistance below 4.1 nOhm. The electrical insulation process was qualified to 30 kV and 15 kV under Paschen conditions. Shear pin drilling and installation within specified tolerance was achieved. Modules were lifted and positioned within 2 mm. The lessons learned, and subsequent work execution onsite within the IO facility assembly hall are also presented.

The Central Solenoid (CS) is a key element of the ITER Magnet system, including six identical coils, called modules, assembled together to form a 4 m outer diameter, 13 m high solenoid. It is a superconducting magnet, with a peak field up to 13 T. The magnet is enclosed inside a steel structure providing vertical pre-compression and mechanical support. The room temperature pre-compression load is necessary to prevent displacement and gap between coils during the exploitation phase and therefore reduce the CS performance and create mechanical stress. The target CS pre-compression is 210 MN realized by the tightening of 45 Multi Jack Tensioners (MJTs) divided in 9 sectors. Procurement of the components and the procedures of the ITER CS is the responsibility of the USA Domestic Agency (USDA), while the ITER Organization (IO) will carry out the assembly of these components. IO signed in 2018 a collaboration agreement with CEA/IRFM to implement the pre-compression procedure proposed by the USDA using a dedicated mock-up with the main objectives: - Simulating the all modules stack rigidity with Spring plates; - Characterize the young's modulus of one CS section module; - Validating the technical procedures proposed - Estimating the time impact to accomplish the CS pre-compression process in accordance to the complete ITER Tokamak assembly. The CS mock-up Pre-compression Verification (CSPV) uses a 1/9 segment of the overall CS with a mix of ITER Tokamak and mock-up components. The target load of the mock-up is 23.3 MN to be in accordance with the overall CS pre-compression load value. This paper presents the CSPV mock-up activities, the challenges encountered and the results gained. As a conclusion, the tightening process is highlighted to estimate its applicability for the ITER CS assembly in term of technical and organisational aspects. The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.

In the frame of a contract with ITER Organization (IO) on magnets assembly support, CEA designed and built a superconducting joint test facility called SELFIE (ITER SELf-FIEld joint test facility). This facility is installed at CEA Cadarache and started to operate in 2022. This project was initiated by IO for quality control of critical assembly activities. Indeed, the magnet superconducting joints assembly is a special process, for which the performance cannot be verified until the full Tokamak is at cryogenic temperature and obviously repair cannot be envisaged once the machine is assembled. Therefore, the quality control of these joints assembly relies on procedures and qualification of the workers in charge of their implementation.As the joints assemblies will span over three years of the ITER construction, the qualified workers will have to assemble periodically some Production Proof Samples (PPS) joints to train and keep their certification valid. The purpose of SELFIE is to test these PPS in a timely manner. The tests scope is the measurement of the PPS resistance (few nOhms). For that purpose, PPS integrated in ITER conductors length (similar to 200 kg weight and 3600 mm length) are tested in a liquid helium bath (4.2 K), at nominal current (up to 70 kA), in self-field.The current is provided by a superconducting transformer integrated in the same cryostat as the sample.CEA finalized the preliminary design in 2019, complying with the requirement to achieve a full test sequence within one week (controlled cool down, test and warm-up), with an optimised operation cost. The detailed design phase was started in April 2020 followed by the manufacturing phase up to mid 2021. SELFIE integration and installation were achieved in December 2021 and the cold commissioning done in January 2022. This paper presents the SELFIE test facility and the first results.

In the framework of EU design activities for dimensioning the future fusion DEMOnstration reactor (DEMO), in-depth analyses were conducted in EUROfusion context, aiming to define the design of the DEMO magnets system. For the last DEMO baseline, CEA has proposed for Toroidal Field (TF) coils a concept with radial plates (named WP#4) which is ITER-like, with a round conductor embedded in steel plates. In order to consolidate this design, CEA conducted fine analyses that assess thermal hydraulic and mechanical aspects to allow ensuring compliance with design criteria. The outcomes of theses analyses were used to improve the TF design in order to avoid relying on conservative approaches at pre-design stage, which often end up in material over dimensioning, that penalize cost and space occupation. In this regard, two improving approaches will be exposed, that deal with mechanical consideration and temperature operation, together with future perspectives. On the other hand, CEA designed and manufactured in collaboration with ASIPP a full-size conductor sample derived from a previous TF CEA concept (square-in-square conductor) which is expected to operate at 88 kA and 12T. The two conductor legs of this sample were designed by CEA and manufactured by ASIPP following an extensive quality assurance (QA) preparation process. The sample is presently in course of assembly in CEA Cadarache and is expected to be tested in SULTAN facility (Villigen, CH) to assess its behaviour in both DC and AC regimes. An overview of the status of activities and future perspectives is given in this paper.

The Central Solenoid (CS) is the core element of the ITER magnet system, contributed as in-kind procurement by the US Domestic Agency. Made up of six modular inter-exchangeable module coils, vertically stacked, forming a 15 m high and 4 m outer diameter solenoid, to be inserted inside the central tokamak core, after assembly of the 18 D-shaped Toroidal Field (TF) coils. Each module uses a Nb 3 Sn conductor internally cooled by circulation of supercritical helium at 4.5 K and supplied, for each module, with a 45 kA current by the way of two vertical leads exiting from the module outer radius. The available space between the CS and TF magnets being very limited, the US DA has developed a compact - so called Coax Joint (CJ) - devoted to connect the respective twelve module leads to the feeders located at top and bottom of the CS via dedicated busbar extensions. The CS coax joint assembly procedure as developed by US DA would make use of CS assembly onsite soldering process, to be executed under supervision of the ITER Organization (IO) in the tokamak assembly hall. In order to mitigate risks related to these soldering activities needed at assembly stage, IO has proposed an alternative assembly process based on indium wires compaction - so called Coax Compacted Joint (CCJ)- and initiated its prototype development and qualification in time. The hereby presented CCJ design solution is based on four copper quadrants, each including an embedded straight Rutherford-type superconductor cable-strip to transport the current. The current is transferred from the lead to the quadrants by the way of compacted indium wires. A steel jacket welded around the joint ensures the mechanical support as well as the needed leak tightness. The paper describes developments made by CEA under IO supervision from the conceptual design to the testing of joints in relevant cryogenic and operative conditions.

The Central Solenoid (CS) is a key element of the ITER Magnet system, including six identical coils, called modules, assembled together to form a 4 m outer diameter, 13 m high solenoid. It is a superconducting magnet, using a 45 kA Nb3Sn conductor internally cooled by circulation of supercritical helium at 4.5 K with a peak field up to 13 T. It is enclosed inside a structure providing vertical pre-compression and mechanical support. Procurement of the components and the special assembly tooling of the ITER CS is the responsibility of US ITER, the ITER Domestic Agency of the USA, while the ITER Organization (IO) will carry out the assembly of these components. US ITER has awarded several contracts since 2011 to supply seven modules, including a spare, structure components, and the special tooling required for the CS pre-assembly. All deliveries are organized with the objective to start the CS assembly at IO site by the end 2021. IO has organized special process training and is now starting first phases of the assembly mostly focused on on-site assembly contractor. The paper describes the CS module manufacturing and delivery status, reports the special process final development and training, and assembly status. In particular, the outcomes from first modules factory acceptance tests results will be reported as well as the development and training for special activities to start the first module stacking. The bus bar joint and pre-compression processes with their related tooling qualifications will be detailed.

The European DEMO, i.e. the demonstration fusion power plant designed in the framework of the Roadmap to Fusion Electricity by the EUROfusion Consortium, is approaching the end of the pre-conceptual design phase, to be accomplished with a Gate Review in 2020, in which all DEMO subsystems will be reviewed by panels of independent experts. The latest 2018 DEMO baseline has major and minor radius of 9.1 m and 2.9 m, plasma current 17.9 MA, toroidal field on the plasma axis 5.2 T, and the peak field in the toroidal-field (TF) conductor 12.0 T. The 900 ton heavy TF coil is prepared in four low-temperature-superconductor (LTS) variants, some of them differing slightly, other significantly, from the ITER TF coil design. Two variants of the CS coils are investigated—a purely LTS one resembling the ITER CS, and a hybrid coil, in which the innermost layers made of HTS allow the designers either to increase the magnetic flux, and thus the duration of the fusion pulse, or to reduce the outer radius of the CS coil. An issue presently investigated by mechanical analyzes is the fatigue load. Two variants of the poloidal field coils are being investigated. The magnet and conductor design studies are accompanied by the experimental tests on both LTS and HTS prototype samples, covering a broad range of DC and AC tests. Testing of quench behavior of the 15 kA HTS cables, with size and layout relevant for the fusion magnets and cooled by forced flow helium, is in preparation.

The ITER Central Solenoid (CS) will be one of the world's largest and most powerful pulsed superconducting electromagnet ever built; at an approximate weight of 1300 tons and a total height of 18 m consisting of a stack of six electrically independent 4.1 m diameter modules. In order to electrically connect the CS with the feeder busbars, 12 twin box joints are used to assure an efficient high current transfer while avoiding excessive AC losses. The fabrication of the box entails a succession of steps: explosion bonding of the stainless steel and the copper, precision machining of the internal part of the box and the cover, introduction of the conductor bundle followed by a controlled compaction to achieve the required void fraction, closure welding the cover onto the box, and subsequent reaction heat treatment (HT) for the formation of the Nb3Sn superconductor. The combined effect of all these fabrication processes, if not optimized, can lead to significant residual stresses and large localized plastic deformation acting during HT, which have empirically shown to result into microstructural heterogeneities and in the worst cases, cracking, and thereby component disqualification for use into a nuclear environment. The paper summarizes design optimizations investigated through mock-ups and could be implemented to remedy the present manufacturing fabrication technology process qualification failure(s). Various solutions have been realized by changing different design parameters whose effect on the response to the HT is studied. Dimensional metrology and residual stress measurements via hole - drilling method complemented with metallographic investigations were performed to assess the suitability of each of the solutions. Additionally, an innovative test bench is described, that was implemented for in - situ monitoring of a twin box mock ups during HT.

The ITER Central Solenoid (CS) has terminal butt-type joints called Coaxial joints. It was decided to study a design of this joint with rutherford shunts, and to build models for its resistive and inductive behaviors. In particular, the behavior of the joint under magnetic field transients is investigated with various analytical models that are compared with a FEM model. The key point of the study was to verify that the induced currents were reasonable and would not induce flux jumps in the rutherfords. To validate the model, a prototype with simplified geometry was tested in the CEA Josefa facility under various field ramps. The comparison between model and experimental results are presented and discussed.

In the framework of the EU-JA Broader Approach agreement, the 20 Toroidal Field (TF) coils (18 coils for the tokamak + 2 spares) produced for the JT-60SA tokamak were qualified in cold conditions and at full current (25.7 kA) in self-field conditions and heated up until quench occurs, along standardized testing procedures, at the Cold Test Facility (CTF) installed at CEA-IRFU (Saclay, France). The analysis of the tests results will be presented using simple dynamic models for the coolant exchange with cable in the steady-state regime. In addition to this approach, the coils operating limits in CTF conditions will be explored, integrating the global statistics of their strands' critical performances. The latter are built by processing TF strand performances data obtained during strand production phase. Those statistical data will be used to calculate the TF coils performances in the CTF configuration. Results will be discussed and compared to the TF coils experimental statistics, followed by a tentative quantified interpretation: The results will be transposed into variation of effective macroscopic parameters such as local heat load on pancakes or temperature margin potential increase. Once this macroscopic study is assessed, it stands as the first step toward more refined future analyses. Finally, using the above method, a first predictive application will be conducted on the JT-60SA tokamak operation configuration.

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 precompression 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 EU DEMO reactor is expected to be among the first applications of fusion for electricity generation in the near future and the design of its magnet system is of central importance as driving power plant performance, budget and production efficiency. In this purpose activities were led by CEA in the framework of EUROfusion Consortium to contribute to the EU DEMO magnet system design. It encompassed design activities (dimensioning and development of associated modelling tools) with R&D (design and tests of mock-ups). The CEA design activity was mainly oriented towards Toroidal Field (TF) coils system to propose a conceptual option (pancake-wound, no radial plates) established with a semi-analytical CEA tool that considers the inter-dependent electromagnetic and mechanical behaviors. Then the proposed design is consolidated by detailed analyses: Thermo-hydraulics evaluation by coupling THEA, TRAPS and CAST3M softwares respectively for thermo-hydraulics, electromagnetic and thermal items. The outcomes obtained in normal and off-normal regimes are exposed and discussed in the paper; Mechanics evaluation with the most stressed zones identified and their criticity evaluated, in particular in the insulation zones. Design optimization analyses were conducted on jacket shape, together with investigations on the thermo-mechanic hotspot criterion. Further to the TF system, the central solenoid design was addressed and an optimization analysis will be presented and discussed. On another side, CEA also conducted R&D activities, mostly regarding the TF system with hydraulic tests at variable void fraction to explore its impact on helium friction and a full-scale TF conductor sample design and manufacture.

The ITER magnet system will be the largest superconducting magnet system ever built. The system, all inside a cryostat, is mainly composed of a Central Solenoid (CS) split in 6 modules, a set of 18 Toroidal Field (TF) D-shaped coils and 6 Poloidal Field (PF) coils. Each of these coils uses variable type of cable-in-conduit-conductors (CICC) actively cooled by a forced supercritical helium flow. Their electrical supply from the current feed-through of the cryostat is done with Main Busbars (MB) using similar CICC. The electrical MB to coils as well as internal PF and TF coils connections rely on the twin box concept developed by CEA in the early R&D phase. After electrical validation of joint prototypes for the PF and the MB conductors on full size samples, specific hydraulic characterization tasks were done through the Magnet Infrastructure Facilities for ITER (MIFI) contract between ITER Organization (IO) and CEA devoted to develop, improve and qualify manufactured components and assembly processes. These tasks were done on the samples using the CEA OTHELLO dedicated facility able to operate with gaseous N-2 in a large Reynolds range at room temperature. The paper explains the way followed to get a full hydraulic characterization of the MB and PF5 joints for the two flow directions. The study of the flow distribution between parallel cooling channels inside the PF5 joint revealed a bypass of the active joint region. The paper reports on this hydraulic behaviour in the relevant magnets operating conditions and outlines the design changes in the joints provoked by the results of this study.

JT-60SA is an advanced superconducting fusion Tokamak jointly constructed by Japan and Europe. It takes the main missions of addressing key physics issues and providing direct operation experiences for ITER and DEMO reactors. In the framework of the JT-60SA project, the 18 NbTi superconducting Toroidal Field (TF) coils have been tested in quench conditions at CEA Paris-Saclay. In our previous study of these coils tests, a typical quench behavior has been identified and analyzed. This behavior shows for the majority of the coils (13 tests over 19), four dynamic phases including a quench initiation phase with a velocity around 3 m/s which is rapidly (several hundreds of milliseconds) followed by a quench acceleration one at around 30 m/s near the start of the current discharge. This is called "early" quench acceleration. Nevertheless, a few number of coils showed a quench acceleration with a certain delay of about 0:5 s to 2 s after the current discharge. This paper will propose a study of this "delayed" quench acceleration phenomenon in two steps: Firstly, a physical analysis of the experimental data of these few tests; Secondly, a numerical study with the code THEA analyzing the testing conditions impact on the beginning delay of the quench acceleration phase.

The toroidal field (TF) system of JT-60SA tokamak comprises 18 NbTi superconducting coils. In each TF coil, 6 cable-in-conduit conductor (CICC) lengths are wound into 6 double-pancakes and carry a nominal current of 25.7 kA at a temperature of 4.7 K. Each coil is tested in the cold test facility (CEA Saclay), up to quench. A SuperMagnet (CryoSoft) model has been developed, each of the 12 pancakes being modeled with THEA and cryogenic circuit being modeled with FLOWER. The experiments showed that helium inlet temperature increases until quench triggering at about 7.5 K on C11 and C13, with quench starting on a lateral and on a central pancake, respectively. Each test has been simulated, applying (or not) a realistic heat load from casing to winding pack that was estimated from experimental measurements. A parametric analysis has been performed, considering realistic or null heat flux deposition, variation of friction factor (in fabrication quality range), and CICC critical current density (in strand Jc performance range). This last parameter was found to have the largest impact on the localization of the first quenched pancake (central or lateral).

ITER magnets are in the final phases of production and are preparing for the upcoming assembly challenges. The ITER magnet team has expressed the need for close-range support labs in order to perform qualification tests, procedure tuning, mockup testing, site acceptance tests, as well as training of assembly staff and logistics management. In the framework of its general support to the ITER project, CEA's Institute for Magnetic Fusion Research proposed a support structure in line with this need on CEA Cadarache premises. In July 2014, the MIFI agreement (Magnet Infrastructure Facilities for ITER) was signed between ITER organization and CEA. It led to the creation of four laboratories and a storage area dedicated to magnet instrumentation components. This paper gives a general description of the work organization in MIFI, and describes the main ongoing activities with specific focuses on critical qualification activities like high-voltage testing and diagnostic of Glass-Kapton-Glass resin impregnated insulation and the scale 1 mock-up and test of the intermediate outer intercoil structure assembly procedure.

The general and detailed design of the JT-60SA toroidal field (TF) coil system was done by all the Voluntary Contributors in the project, CEA, ENEA, and F4E. The French part including the supply of 9 + 1 spare of the 18 TF winding packs and their integration in casings was entrusted mid-2011 to Alstom (Belfort, France), now General Electric. Manufacturing flow definition, drawings, and quality assurance documentation were achieved prior qualification of the critical processes. The procurement and commissioning of the needed tooling was led in parallel and the production was started in January 2014. The first winding pack was ready in the end of 2014. The integration inside the casing started in March 2015 for coil completion end 2015. Then, the coil was delivered to the coil test facility for testing at nominal operation condition (T = 4.5 K, I = 25, 7 kA) and for qualifying the coil resilience to quench events. In parallel, the successive coils were engaged in the manufacture process. The paper reports the status of the manufacture and gives a feedback on the main issues and on the technical solutions implemented. A focus is made on the mastering of the manufacturing processes, which is highlighted by the learning curve of the TF coils manufacturing.

This paper describes the result of the ITER feeder main busbar joint sample qualification test as confirmation of the requirement of busbar joint resistance, 2 nΩ at 70 kA at zero background field, as well as that of joint performance in various magnetic fields to investigate stability and current distribution characteristics in feeder-type joint box. The results support the quality of the joint manufacturing process for ITER main busbar joint. The qualification sample design was prepared to be tested in SULTAN facility. The SULTAN joint sample consists of joints to be qualified at the level of the peak field and upper terminations. In bottom joints, twin-box feeder-type praying hands configuration is applied. In upper terminations, one of them is made with solder-filled cable for optimum current distribution. The other takes the same length of the copper sole and contact with the busbar cable as those positioned in bottom of the sample. The sample undergoes a test program, which includes joint resistance measurement, ac losses, and stability margin test. The outcomes of those test programs are reported.