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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.

Alarge superconducting machine, JT-60SA has been constructed to provide major contributions to the ITER program and DEMO design. For the success of the ITER project and fusion reactor, understanding and development of plasma controllability in ITER and DEMO relevant higher beta regimes are essential. JT-60SA has focused the program on the plasma controllability for scenario development and risk mitigation in ITER as well as on investigating DEMO relevant regimes. This paper summarizes the high research priorities and strategy for the JT-60SA project. Recent works on simulation studies to prepare the plasma physics and control experiments are presented, such as plasma breakdown and equilibrium controls, hybrid and steady-state scenario development, and risk mitigation techniques. Contributions of JT-60SA to ITER and DEMOhave been clarified through those studies. Peer Reviewed Article escrit per 127 autors/autores: M Yoshida, G Giruzzi, N Aiba, J F Artaud, J Ayllon-Guerola, L Balbinot, OBeeke, E Belonohy, P Bettini, W Bin, A Bierwage, T Bolzonella, M Bonotto, CBoulbe, J Buermans, M Chernyshova, S Coda, R Coelho, S Davis, C Day, GDeTommasi, M Dibon, A Ejiri, G Falchetto, A Fassina, B Faugeras, L Figini, M Fukumoto, S Futatani, K Galazka, J Garcia, M Garcia-Muñoz, L Garzotti, L Giacomelli, L Giudicotti, S Hall, N Hayashi, C Hoa, M Honda, K Hoshino, M Iafrati, A Iantchenko, S Ide, S Iio, R Imazawa, S Inoue, A Isayama, E Joffrin, K Kamiya, Y Ko, M Kobayashi, T Kobayashi, G Kocsis, A Kovacsik, T Kurki-Suonio, B Lacroix, P Lang, Ph Lauber, A Louzguiti, E de la Luna, G Marchiori, M Mattei, A Matsuyama, S Mazzi, A Mele, F Michel, Y Miyata, J Morales, P Moreau, A Moro, T Nakano, M Nakata, E Narita, R Neu, S Nicollet, M Nocente, S Nowak, F P Orsitto, V Ostuni, Y Ohtani, N Oyama, R Pasqualotto, B Pégourié, E Perelli, L Pigatto, C Piccinni, A Pironti, P Platania, B Ploeckl, D Ricci, P Roussel, G Rubino, R Sano, K Särkimäki, K Shinohara, S Soare, C Sozzi, S Sumida, T Suzuki, Y Suzuki, T Szabolics, T Szepesi, Y Takase, M Takech, N Tamura, K Tanaka, H Tanaka, M Tardocchi, A Terakado, H Tojo, T Tokuzawa, A Torre, N Tsujii, H Tsutsui, Y Ueda, H Urano, M Valisa, M Vallar, J Vega, F Villone, T Wakatsuki, T Wauters, M Wischmeier, S Yamoto, L Zani

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.

International audience; The JT-60SA cryogenic system was commissioned in 2016 in closed loop, without the cryogenic users (superconducting magnets, current leads, thermal shields). The first plasma operation is expected in 2020. This paper updates the heat load profiles resulting from the cooling loop of TF magnets, and received by the refrigerator and its thermal damper. The heat load profiles are calculated through thermal-hydraulic simulations of the magnets and the associated cryo-distribution, also named as supercritical helium loops. This update was performed by taking into account new data from the TF magnets (measured pressure drops, updated heat loads coming from the plasma), as well as a more accurate thermal model of the TF magnet. Previous simulation had been performed using the Vincenta code in 2010 and were used for the cryogenic system acceptance tests. The new thermal-hydraulic model is performed by using Simcryogenics, the modeling tool dedicated to refrigeration and cryo-distribution developed by CEA (Commissariat à l’énergie atomique et aux énergies alternatives). The differences between the two simulation results are highlighted and analyzed. These simulations provide the pulsed heat load profiles smoothed by the cryo-distribution and deposited into the thermal damper.

Sedlak, K. et al., 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., This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014–2018 and 2019–2020 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission. This work was supported in part by the Swiss National Science Foundation (SNF) under contract number 200021_179134.

The Toroidal Field system of the JT-60SA tokamak is composed of 18 NbTi superconducting coils all individually tested in the Cold Test Facility (CTF, CEA/IRFU Saclay, France). The test program includes for each coil a progressive operating temperature increase at nominal current (25.7 kA) up to quench followed by a safety discharge. To account for the heat coming from the casing to the winding pack (WP), a transient test procedure was settled so as to trigger the quench at peak field in the coil; nevertheless the quench location was identified on a side double-pancake in the first two tested coils. We present the CEA analyses for the quench test of the second coil which showed similar behavior as the first one but received a more extensive instrumentation. The coil performance is analyzed in light of the NbTi strand superconducting properties coming from the strand characterization program and of the full-size conductor tests carried out by CEA in the SULTAN facility. These analyses involve cable thermal and electrical modelling developed at CEA and already used in ITER R&D, design and characterization programs.

JT-60SA is a fusion experiment which is jointly constructed by Japan and Europe and which shall contribute to the early realization of fusion energy, by providing support to the operation of ITER, and by addressing key physics issues for ITER and DEMO. In order to achieve these goals, the existing JT-60U experiment will be upgraded to JT-60SA by using superconducting coils. The 18 TF coils of the JT-60SA device are provided by European industry and tested in a Cold Test Facility (CTF) at CEA Saclay. The first coils were tested at the nominal current of 25.7 kA and at a temperature between 5 K and 7.5 K. The main objective of these tests is to check the TF coils performances and hence mitigate the fabrication risks. These first tests allowed checking a certain number of performances of the first TF coils: DC/AC insulation, cooling down characterization, RRR of the conductor, pressure drop in the winding pack and temperature margin against a quench. This paper will give an overview of the main experimental results obtained during these tests. These results will be analyzed and discussed in the light of the expected TF coils performances.

Previous thermohydraulic studies performed for the ITER TF, CS and PF magnet systems have brought some important information on the detection and consequences of a quench as a function of the initial conditions (deposited energy, heated length). Even if the temperature margin of the Correction Coils is high, their behavior during a quench should also be studied since a quench is likely to be triggered by potential anomalies in joints, ground fault on the instrumentation wires, etc. A model has been developed with the SuperMagnet Code (Bagnasco et al., 2010) for a Side Correction Coil (SCC2) with four pancakes cooled in parallel, each of them represented by a Thea module (with the proper Cable In Conduit Conductor characteristics). All the other coils of the PF cooling loop are hydraulically connected in parallel (top/bottom correction coils and six Poloidal Field Coils) are modeled by Flower modules with equivalent hydraulics properties. The model and the analysis results are presented for five quench initiation cases with/without fast discharge: two quenches initiated by a heat input to the innermost turn of one pancake (case 1 and case 2) and two other quenches initiated at the innermost turns of four pancakes (case 3 and case 4). In the 5th case, the quench is initiated at the middle turn of one pancake. The impact on the cooling circuit, e.g. the exceedance of the opening pressure of the quench relief valves, is detailed in case of an undetected quench (i.e. no discharge of the magnet). Particular attention is also paid to a possible secondary quench detection system based on measured thermohydraulic signals (pressure, temperature and/or helium mass flow rate). The maximum cable temperature achieved in case of a fast current discharge (primary detection by voltage) is compared to the design hot spot criterion of 150 K, which includes the contribution of helium and jacket.

In the framework the design activities conducted in EU for the future fusion demonstration reactor (DEMO), extensive analyses were led in the EUROfusion context, aiming at ultimately designing the DEMO magnets system. The DEMO tokamak configuration was updated in view of improving the operation merits versus the previous design defined in 2015. In the objective of properly addressing tokamaks general magnetic systems, CEA has adapted macroscopic design tools and associated methods to size magnets for pre-dimensionning or optimization purposes. Once pre-dimensioned using inputs from system-level analyses the magnet concepts undergo an evaluation process by detailed analyses (thermohydraulic and mechanical) highly accurate but strongly time consuming. To avoid repeating this sequence, we aimed in this paper at improving pre-dimensioning process reliability for the final benefit of the whole design workflow. The material exposed in this paper is the path to this improvement achieved through a sery of pre-dimensioning parametric cases applied on the DEMO TF system baseline newly issued in 2018 by EUROfusion. The optimization steps are described, showing which design TF system variables are chosen for optimization and what is their contribution to ultimately rend the design as efficient as possible with respect to given merits. In this aim, former detailed finite elements analyses, carried out on the previous DEMO configuration, are taken as reference to support the quantitative choices of the pre-dimensioning method improvement (criteria or input modifications). Furthermore, TF design is also explored on a new aspect: the possibility of including radial plates in the winding pack.

The Toroidal Field (TF) system of the Tore Supra/WEST tokamak comprises 18 NbTi superconducting coils, cooled by a static superfluid helium bath at 1.8 K and carrying a nominal current of 1255 A. The 19th December 2017, at the end of plasma run #52205, a quench of TFC-09 was detected first on a secondary thermohydraulical signal (helium liquid level) and triggered the current Fast Safety Discharge (FSD). A numerical model of this quench has been developed with SuperMagnet (CryoSoft). The whole TFC-09 circular coil is modelled by THEA as a single large Cable-In-Conduit Conductor (CICC) with 2028 large rectangular strands (monolithic conductors of length equal to coil average perimeter). The external quench helium relief circuit (cold and warm safety valves, rupture disk and magnetic valve with corresponding pressure set) is modelled by FLOWER. The helium pressure in the coil, upstream of the cold safety valve (maximum value of 9 bar), and the expulsed helium temperature have been used as comparison between measurements and calculations which depend on the Minimum Quench Energy (MQE) used in THEA with small initial heat deposition length (few tens of centimeters) at low field region (external leg). This energy, in the order of few kJ, is compared to the real shape and energy of neutron and gamma flux caused by highly energetic runaway electrons colliding the outboard plasma facing components and which induced the quench. The expulsed helium mass flow rate of nearly 6 kg/s during 5 s has also been calculated. This event confirms on one hand the criticality and the possible occurrence of a so called “smooth quench” caused by small initial heat deposition length and at low field region and on the other hand the important interest of secondary quench detection which can be useful for other tokamak magnets safe operation and protection.

In the framework of the European fusion program for energy, EUROfusion funds the studies for the future fusion power demonstrator reactor DEMOnstration fusion reactor (DEMO). Commissariat a l'energie atomique et aux energies alternatives (CEA) is involved in the conceptual design of the superconducting conductors for the toroidal field (TF) and central solenoid (CS) magnets. The CEA design proposal corresponds to wind and react Nb3Sn pancake-wound coils using cable-in-conduit conductors cooled at about 4.5 K by forced flow of supercritical helium. This paper presents the latest thermal-hydraulic analyses performed on both TF and CS conductors. Two TF conductor designs have been analyzed, with nominal currents of 111 kA and 88 kA, respectively. The analyses were performed in normal (burn) and off-normal (quench) conditions. Burn simulations focused on the central and lateral pancakes with heat load corresponding to neutron heating. The central pancake is the most critical one regarding the magnetic field, while the lateral one receives a more important heat load from the casing. The influence of case cooling on temperature margin (ΔTma) was analyzed by means of a dedicated 2D Cast3M model. The ΔTma sensitivity to driving parameters such as conductor friction factor correlations was analyzed. The impact of interturn thermal coupling on temperature margin was also assessed. Quench studies were performed on the TF conductor, considering a quench initiated either on the innermost turn or at the middle of hydraulic length. Regarding the CS conductor, burn scenario was led on the latest design featuring a nominal current of 53.7 kA, focusing on the impact of ac losses and of dwell duration.

Within the framework of the Broader Approach, the French voluntary contributor represented by CEA awarded a contract for the production of 9 toroidal field coils for the JT-60SA project to Alstom, Belfort, France in mid-2011. A first preparatory phase was led to establish the team, produce the manufacture drawings, define the manufacturing process, procure the required tooling and prepare the quality documentation. In parallel, a qualification phase on the critical major processes has proved Alstom's ability to master the processes and reach the requirements. After reviewing of the qualification results and modification of the processes and tooling to overcome the encountered difficulties, a Production Readiness Review has authorized Alstom to start the production winding. A prototype double pancake was wound as the first of series. In addition to complying with the pancake width all around the D shape, the straightness of the centreline in the critical straight leg part was correct. The production of the successive double pancakes to constitute the first winding pack was then completed and the joints and terminals were manufactured. The paper describes the completion of the last qualifications and the status of the winding production.

In case of a magnet quench, the primary quench detection system is classically based on voltage detection. In addition, a secondary quench detection is required and it could rely on signals of thermohydraulic nature. A model based on the coupled GANDALF [1] and FLOWER [2] codes has been developed in previous papers for the Central Solenoid [3] , [4] , and for the Toroidal Field Coils [5] , [6] of ITER. This study focuses on Poloidal Field Coil PF5 as it has the highest magnetic field value along the conductor. Results of some parametric studies are presented and the influence of the initiation parameters of the quench (energy, location and length of heat deposition, time duration for detection). The possibility of a secondary thermohydraulic detection and the main impact on cryogenic system are analyzed and discussed.

The DEMO reactor is expected to be the first application of fusion for electricity generation in the near future. In the DEMO power plant the management of magnet system is of central importance as being driver on many crucial aspects such as nominal power plant performance (toroidal field scales fusion power), overall investment budget (about 1/3 of the total construction cost), production efficiency (full power total availability heavily impacted by magnet off-normal events). Therefore a careful approach is requested for this kind of component to ensure a safe design compatible with a power plant production conditions, keeping a control on the factors prone to degrade the economic model (cost, risk). The derivation of those considerations into practical activities results in a constant attempt to lead in parallel extensive design activities and the mastering of upstream knowledge in magnet behavior. In this purpose design activities on DEMO magnet system were continuously conducted in Europe, particularly evolving since 2012 in structured environments, always backed on the association of several laboratories. Since then, the actors underwent preparatory design phase and then the preconceptual design activity (CDA) phase, that led to evolutions of design in many aspects, from design features bottom lien themselves to evolution of associated tools and methods up to global strategic considerations.

In order to ensure the safety of the ITER TF magnets, a primary quench detection system has been foreseen, based on voltage detection. In addition, a secondary quench detection could rely on signals of thermo-hydraulic nature. As a matter of fact, the development of a quench in a CICC leads to significant variations of pressure and mass flow at the quenched pancake extremities. Analyses of the quench development have thus been performed using the coupled GANDALF and FLOWER codes. This tool allows to simulate the thermo-hydraulic behaviour of one CICC with a model of the external cryogenic circuit. The study has focused on the first seconds of the quench development, supposing that the quench has not been detected earlier by the primary detector. It is shown that signals regarding pressure, mass flow and temperature reach significant high values especially in the connecting feeder associated with the helium inlet. More detailed studies will be needed to select a secondary detector in this region.

In the framework of the EU participation to JT-60SA project, a "pseudo-3D" thermo hydraulic tool was developed in order to address reliably the choice between multiple design options on the TF coils design. The tool based on the ANSYS code was configured to calculate the conductor helium temperature variation with time taking into account the transport of helium in the case and the winding pack. The central criterion used is the conductor temperature margin. The principal loads studied are the nuclear heating, and the eddy current losses in the TF case. A 3D model of the machine with the vacuum vessel, the TF case, and the magnets was built to calculate the eddy currents during a disruption, and the results are used as input in the pseudo 3D thermo hydraulic tool. In the framework of cost and feasibility optimization, this study has permitted to comparatively evaluate pros and cons of TF design options concerning geometry, casing cooling channels and winding pack insulation. Moreover a consolidated basis for definitive choices towards an optimized TF design with respect to feasibility, cost and risks has been proposed.

In the framework of the EU participation to JT-60SA project [1] , a dedicated simulation tool named after Thermo-hydraulic EXtended Tool (TEXTO) was developed at CEA between 2006 and 2007 in order to address in a reliable way the calculation of the magnet conductor temperature increase and temperature margins in different operating conditions. The simulation process is based on three different codes, addressing each specific aspects (MCNP for the 3D nuclear heat calculation, TRAPS for magnetic field, ANSYS for 2D transverse thermal contribution of coil casing), which finally stand as input for the well established code GANDALF (with transient helium properties). Both steady-state operating and disruption transient regimes can be studied with this process and a first application is performed on the basis of the design and operating conditions available at this time on JT-60SA TF magnets, i.e. the first version of the different design stages. The complete analysis is shown together with the associated proposals for the TF conductor layout that could be derived from these studies.

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.

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.

The upgrade of JT-60U magnet system to superconducting coils (JT-60SA) has been decided by both parties of Japanese government (JA) and European commission (EU) in the framework of the Broader Approach (BA) agreement. The magnet system for JT-60SA consists of 18 toroidal field (TF) coils, a Central Solenoid (CS) with four modules, seven Equilibrium Field (EF) coils. The TF case encloses the winding pack and is the main structural component of the magnet system. The CS consists of independent winding pack modules, which is hung from the top of the TF coils through its pre-load structure. The seven EF coils are attached to the TF coil cases through supports which include flexible plates allowing radial displacements. The CS modules operate at high field and use Nb3 Sn type superconductor. The TF coils and EF coils use NbTi superconductor. The magnet system has a large heat load from nuclear heating from DD fusion and large AC loss. This paper describes the technical requirements, the operational interface and the outline of conceptual design of the superconducting magnet system for JT-60SA.

In the ITER superconducting magnets, the CIC conductors cooling is insured by supercritical helium forced flow in the central and annular parallel channels. In the Central Solenoid, with a thick square jacket conductor, the helium inlet is in the highest field region at the inner bore pancake joggle, and it has to support the high hoop force stress level with a very low stress concentration factor. In the TF magnets, a thin jacket circular conductor is wound in double pancakes, inserted into radial plates, stacked and embedded into a steel case. The helium inlets are located at the inner bore, in the limited space between the radial plates and the coil casing. The PF coils helium inlets, not studied here, are similar to the CS ones, but with a lower stress level. A complete qualification work on the CS and TF helium inlets is presented. A design optimization was performed, by FEM analysis, resulting in acceptable stress level for both helium inlets. The welding procedure was qualified and specific fatigue life mock-ups were designed, analyzed and manufactured using representative jacket materials. Fatigue life qualification at 4 K was performed in the FZK test facility applying the relevant loading and number of cycles. Hydraulic mock-ups were manufactured and qualified in the CEA test facility using GN2 at relevant Reynolds number. Pressure drop as well as flow repartition inside the conductors' petals were measured. The mechanical and hydraulic results are presented

The JOSEFA (joint subsize experiment facility) experimental test facility, installed at CEA/Cadarache is devoted to perform tests at cryogenic temperature on subsize superconducting conductor and joint samples under parallel or transverse magnetic field. This facility was built in 1993 to investigate the performances of joints of cable-in-conduit conductors at subsize level and further upgraded in the framework of European tasks. The samples of hairpin type using subsize ITER conductors are cooled by a circulation of supercritical helium in a temperature range from 5 to 15 K and tested at a maximum current up to 10 kA. Two different helium bath cooled magnets allow to apply DC or AC transverse magnetic field up to 3.5 T or longitudinal magnetic field up to 7.5 T. A sliding system with a 240 mm stroke on the sample cryostat allows to test separately in the same sample either the conductor or the joint performances. The paper reports on how, through the conductor and joint development tasks, the facility performances were successfully increased and tested. The ITER TFMC joints using Nb/sub 3/Sn conductors were first developed on this facility. The last developments, performed on the ITER PF NbTi conductors and joints proved this facility to be a versatile and useful tool for superconducting magnet developments and showed the interest of possible upgrading to finalize conductor design.

As the main objective of the toroidal field model coil (TFMC) is to demonstrate the industrial feasibility of the ITER TF coils, the TFMC design includes the particular features of these coils, such as the insertion of the conductors inside stainless steel radial plates (see Fig. 1 ). During ITER transients such as the safety discharge of the magnetic system, eddy currents are induced in the plates, generating a power, which is transferred into the conductor helium channels by a long diffusion process through the conductor steel and insulation, and eventually quenches the system. In this case, the delay introduced by this diffusion is a key parameter in the quench evolution. The model of this diffusion is presented. The diffusion equation is added to the thermohydraulics code gandalf, which enables to take into account the helium circulation along the channel. The ITER TFMC experiment offers an unique possibility to check this model. Specific safety discharges were intentionally initiated from 25 kA to study this phenomenon. The results of these experiments are compared with the predictions. The model proves to be quite effective and can be used for ITER. The heat conductivity and the specific heat used in the model are discussed.

During a safety discharge of toroidal field type magnets, eddy currents and associated heat generation are induced in the plates. A model has been developed from the thermohydraulic code Gandalf with introduction of the equations of the heat diffusion from plates to conductors through the steel and insulation. The comparison of calculation and experimental results for the ITER toroidal field model coil is presented. Preliminary analysis for the ITER toroidal field coils is also presented, taking into account the conductor parameters, the magnetic field and the external hydraulic circuit. The possible quench of the magnetic system is discussed.

During the last phase of the EDA (Engineering Design Activities) of the ITER (International Thermonuclear Experimental Reactor) project, a fully superconducting PF (Poloidal Field) system has been designed to cope with the requirements of the FEAT (Fusion Energy Advanced Tokamak) version of ITER: providing the magnetic field required to develop, shape and control the 15 MA plasma during the 900 s of a typical plasma scenario. The CS (Central Solenoid), divided into 6 Nb/sub 3/Sn subcoils, and the six outer NbTi PF coils will experience severe heat loads especially during the 400 s of the plasma burn: nuclear heating due to the 400 MW of fusion power, thermal radiation, and ac losses (30 to 300 kJ in the PF coils). The ac losses along the PF coil pancakes are deduced from accurate magnetic field computations performed by the TRAPS code (analytical integration of the Biot-Savart law over the cross section of 3-D current elements). Using as input these heat loads, including thermal radiation and nuclear heating, a thermal-hydraulic analysis of the PF coil cable-in-conduit conductor is performed with a finite element code, GANDALF: the temperature increases (0.1 to 0.3 K) are computed, the temperature margins of the conductor are thus evaluated.

In the framework of the ITER (International Thermonuclear Experimental Reactor) project, the TFMC (Toroidal Field Model Coil) has been manufactured by European industry, using about 800 m of a 80 kA Nb/sub 3/Sn cable-in-conduit conductor to wind the 98 turns of the coil. These turns are wound in 10 pancakes which are inserted inside machined stainless steel radial plates, and the whole winding pack is enclosed in a thick stainless steel casing. These passive conductive structures are expected to experience severe induced eddy currents and associated Joule heating during the last testing phase of the TFMC, when fast discharges are triggered to study the electromagnetic and thermal-hydraulic transient behavior of the coil. Diffusing through insulation, the Joule heating induced in the radial plates will add to the AC losses induced in the superconducting cable. The AC losses along the TFMC winding pack are deduced from accurate magnetic field computations performed by the 3-D magneto-static code TRAPS, the eddy currents are calculated with an analytical model using circuit equations, and the Joule heating diffusion is studied with a 1-D thermal analysis model. Using these heat loads as input data, the helium temperature and pressure increases in the conductor are computed with the finite element code GANDALF. The possibility of a quench of the whole magnet is discussed.

The preparative experimental and analysis work as well as the code development for testing the ITER (International Thermonuclear Experimental Reactor) TFMC, one of the large projects of the ITER research and development programme, is described. The expected critical currents of the Nb3Sn superconducting cable are derived from the numerous short sample strand and subcable measurements. A data set of the thermo-hydraulic properties of the bundle conductor with a central channel was elaborated in experiments as an input for the thermo-hydraulic calculation codes GANDALF and MITHRANDIR combined with the hydraulic network code FLOWER. On this basis, the quench behaviour and the injection of gaseous heated helium slugs for determining the operation limits are treated. The mechanical behaviour of the test configuration was investigated by a finite element calculation. The winding type in which the conductor is embedded in radial plates has to be considered for transient voltage peaks as an electrical network. Arising over-voltages by resonance frequencies in the winding pack have to be considered for the integration of the electrical insulation system of the TFMC. Numerous sensors are installed to compare the predicted values and behaviour with the measurements. (C) 2001 Elsevier Science B.V. All rights reserved.

In preparation of the ITER FEAT project, a superconducting toroidal field model coil (TFMC) including 10 parallel pancakes of between 72.2 and 82.0 m long cable-in-conduit conductor (CICC) has been built and has to be tested at FZK. A series of preliminary tests were performed at room temperature and in nitrogen at both CEA-Cadarache and Ansaldo. The tests aimed to assess the thermohydraulics of this coil and specifically to address the pressure drop and the friction factor of the dual channel CICC in a range of Reynolds numbers relevant to the TFMC. These measurements are compared with the predictions which include a hydraulic description of the joint region whose contribution to the total pressure drop is non-negligible. The mass flow distribution between the bundle region and the central channel will be calculated for the cable as well as for the joint region. Measurements on the central solenoid model coil (first layer of inner module and insert coil) in normal operating conditions with supercritical helium are presented. They are in good agreement with the calculations performed from the previously validated correlations. A description of the mass flow imbalance between the TFMC pancakes at low temperatures in helium is done. The major influence of the central spiral is confirmed, since two types of spirals have been included in the TFMC pancakes. This distribution varies as a function of the total mass flow of the analysed hydraulic circuits (TF and CS) of ITER–FEAT.

International audience; Tore Supra routinely addresses the physics and technology of very long-duration plasma discharges, thus bringing precious information on critical issues of long pulse operation of ITER. A new ITER relevant lower hybrid current drive (LHCD) launcher has allowed coupling to the plasma a power level of 2.7 MW for 78 s, corresponding to a power density close to the design value foreseen for an ITER LHCD system. In accordance with the expectations, long distance (10 cm) power coupling has been obtained. Successive stationary states of the plasma current pro le have been controlled in real-time featuring (i) control of sawteeth with varying plasma parameters, (ii) obtaining and sustaining a `hot core' plasma regime, (iii) recovery from a voluntarily triggered deleterious magnetohydrodynamic regime. The scrape-off layer (SOL) parameters and power deposition have been documented during L-mode ramp-up phase, a crucial point for ITER before the X-point formation. Disruption mitigation studies have been conducted with massive gas injection, evidencing the difference between He and Ar and the possible role of the q = 2 surface in limiting the gas penetration. ICRF assisted wall conditioning in the presence of magnetic eld has been investigated, culminating in the demonstration that this conditioning scheme allows one to recover normal operation after disruptions. The effect of the magnetic eld ripple on the intrinsic plasma rotation has been studied, showing the competition between turbulent transport processes and ripple toroidal friction. During dedicated dimensionless experiments, the effect of varying the collisionality on turbulence wavenumber spectra has been documented, giving new insight into the turbulence mechanism. Turbulence measurements have also allowed quantitatively comparing experimental results with predictions by 5D gyrokinetic codes: numerical results simultaneously match the magnitude of effective heat diffusivity, rms values of density uctuations and wavenumber spectra. A clear correlation between electron temperature gradient and impurity transport in the very core of the plasma has been observed, strongly suggesting the existence of a threshold above which transport is dominated by turbulent electron modes. Dynamics of edge turbulent uctuations has been studied by correlating data from fast imaging cameras and Langmuir probes, yielding a coherent picture of transport processes involved in the SOL.

In the framework of the JT‐60SA project, TF coils design activities have led to the development of TEXTO (Thermo‐hydraulic EXtended TOol), a dedicated simulation tool centred on the GANDALF code for calculating the temperature margin (ΔTma) central criterion. From a first version providing conservative results, this tool has been upgraded to a pseudo 3D model by integrating a 1D thermo‐hydraulic approach in ANSYS, leading to an additional and independent module named TACOS (Thermo‐hydraulic Ansys COmputation Semi 3D).By providing He temperature in conductors, TACOS is well adapted for evaluating the impact of TF coils design choices in the framework of cost and feasibility optimization. TACOS also provides more accurate values of transverse heat flux from case to conductors, which can then be injected into TEXTO and allow the calculation of the temperature margin, thanks to the functionalities of GANDALF.Several validation calculations have been performed for the reference operation scenario, by comparison with GANDALF for validating the simplified thermo‐hydraulic method of TACOS and by comparison with VINCENTA for an overall validation of the TACOS/TEXTO tool. Results have shown a good consistency between the different models, with a conductor temperature dispersion around 0.05 K in the critical zone.

During the plasma reference scenario of ITER, the Central Solenoid (CS) coils are submitted to pulsed heat loads due to pulse operating currents. All Cable‐In‐Conduit‐Conductors (CICC) of these magnets are cooled with a forced flow of supercritical helium. Thermohydraulic calculations aim to estimate the heat loads (static and pulsed) transferred to the thermal bath connected with the cryogenic refrigerator.These calculations are performed on the ITER CS system as its design is practically frozen. Comparison of the models is made through cross checking of two numerical codes: 1‐ Vincenta (Efremov Institute, Saint‐Petersburg) a quasi‐3D code which can model all the CS coil CICCs in parallel and the first cryogenic loop. 2‐ Gandalf (CryoSoft) which models one CICC (1‐D) and an associated cryogenic circuit (Flower) which has to be adapted in order to be relevant.The Gandalf model deals with the pancake 101 experiencing the most severe conditions. The results of Vincenta code are procured by ITER Organization (IO). These results are compared with Gandalf‐Flower model (presented herein) in terms of helium pressure, temperature and mass flow rate at inlet and outlet of the studied conductor. This normal operation should be compared with a quench detection, in a further work.During the plasma reference scenario of ITER, the Central Solenoid (CS) coils are submitted to pulsed heat loads due to pulse operating currents. All Cable‐In‐Conduit‐Conductors (CICC) of these magnets are cooled with a forced flow of supercritical helium. Thermohydraulic calculations aim to estimate the heat loads (static and pulsed) transferred to the thermal bath connected with the cryogenic refrigerator.These calculations are performed on the ITER CS system as its design is practically frozen. Comparison of the models is made through cross checking of two numerical codes: 1‐ Vincenta (Efremov Institute, Saint‐Petersburg) a quasi‐3D code which can model all the CS coil CICCs in parallel and the first cryogenic loop. 2‐ Gandalf (CryoSoft) which models one CICC (1‐D) and an associated cryogenic circuit (Flower) which has to be adapted in order to be relevant.The Gandalf model deals with the pancake 101 experiencing the most severe conditions. The results of Vincenta code are procured by ITER Organization...

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.

New tests and measurements performed (Othello Facility, EFDA Task) on TF mock up cooling inlet and different central spirals (characteristics: hydraulic outer diameter and perforation ratio) are presented, as well as the new model of singular and linear friction factor. The ITER Coils CICC hydraulic length pressure drop is determined in operating conditions (m=8 g/s, P=0.6 MPa and T=5 K): the important result is an increase in linear pressure drop for the TF (290 Pa/m) and CS (430 Pa/m), in comparison with prototype model coils TFMC (100 Pa/m) and CSMC (180 Pa/m). The main reason is the reduction of the central spiral diameter and associated increase of friction factor and bundle to total mass flow ratio α (from 1/3 up to 2/3 typically). The ratio of singular cooling inlet to CICC linear pressure drop is estimated: TF mock up ratio (3 m) is lower than previous CS mock up tested (12 m), due to design changes. The cryogenic power necessary to compensate the CICC pressure drop is calculated for the 4 primary loop circuits: typically 2.3 kW at 5 K for TF winding system represents 40% of the whole average TF winding magnet heat loads during operation.

Publisher Summary This chapter presents the testing possibilities and a preliminary design of the ITER superconducting magnet system cold test facility, including the process engineering of a cooling system. The ITER superconducting magnet system consists, besides six poloidal field coils and the central solenoid, of 18 toroidal field coils with the size of 16.6 m x 9 m and a weight of 400 t each. For each of these TF coils an intensive cold test is recommended prior to the installation into the torus to minimize the risk of malfunction which would cause tremendous costs and massive time delay. The cold test of a complete ITER TF coil at 4.5 K at a reduced coil current of about 30 kA is the most favorite option. This test scenario allows to detect high voltage, leak, joint, sensor and cooling failure before installation into the ITER machine, and therefore, to avoid enormous disassembly and reinstallation work and as consequence large additional costs and delay. In addition the proposed cold test facility allows to prove the cooling circuit planned for ITER and will allow to train the ITER staff when the cold test facility is built on ITER site.

The ITER Toroidal Field (TF) magnet is made of 18 coils. The 7 winding‐pack modules constituting their structure, contain each one double pancake Cable‐In‐Conduit‐Conductor (CICC) inserted in a stainless steel radial plate. The winding pack is enclosed inside a thick stainless steel case. During a safety discharge of the TF system, eddy currents and associated heat generation are induced in the plates and in the case. Such fast discharges have been performed during Phase I and Phase II experiment of the Toroidal Field Model Coil (TFMC) tests to investigate heat generation and transfer. Simple models have been developed to predict the behaviour of the coil during these transients. These models have been validated by calorimetric measurements. These models can also be used to estimate the temperature increase of the TF Conductor during a plasma disruption and to investigate whether a quench can happen.

The ITER Toroidal Field Model Coil (TFMC) was designed and manufactured by the European Home Team in collaboration with European industry. The test in the TOSKA facility of the Forschungszentrum Karlsruhe was successfully performed in 2001 and 2002 and has confirmed that the used design and construction principles are applicable for the ITER TF coils.The TFMC was tested up to the rated current of 80 kA as a single coil and in the background field of the EURATOM LCT coil in order to achieve ITER TF coil relevant stress levels. For the operation of the TFMC and LCT coils, special developed forced‐flow‐cooled current leads were used. Both coils with a total weight of 108 t were forced‐flow‐cooled with supercritical He at 4.5 K in a secondary cooling loop connected to the 2 kW refrigerator. However, for currents above 11.4 kA in the LCT coil, its winding had to be cooled at 3.0 K with a separate refrigerator and cooling system.Details of the process engineering of both cooling systems will be described. The o...

In the framework of the ITER (International Thermonuclear Experimental Reactor) FEAT (Fusion Energy Advanced Tokamak) project, a fully superconducting PF (Poloidal Field) system has been designed in detail. The Central Solenoid and the 6 equilibrium coils constituting the PF system provide the magnetic fields which develop, shape and control the 15 MA plasma during the 1800 s of a typical plasma scenario. The 6 PF coils will be wound two-in-hand from a 45 kA niobium-titanium CICC (Cable-In-Conduit-Conductor). These coils will experience severe heat loads specially during the 400 s of the plasma burn: nuclear heating due to the 400 MW of fusion power, thermal radiation and AC losses (30 to 300 kJ). The AC losses along the PF coil pancakes are deduced from accurate magnetic field computations performed with a 3D magnetostatic code, TRAPS. The nuclear heating and the thermal radiation are assumed to be uniform over a given face of the PF coils. These heat loads are used as input to perform the thermal and hydraulic analysis with a finite element code, GANDALF. The temperature increases (0.1 to 0.4 K) are computed, the margins and performances of the conductor are evaluated.

In the frame of the ITER program, a Toroidal Field Model Coil is being built in Europe. Calculations about the thermohydraulic behaviour of this coil are presented in detail using the codes SARUMAN and GANDALF. The Helium temperature increase along the cooling circuits, the pressure drop and the mass flow partition inside the cable during normal operation will be discussed. One of the most crucial points is the behaviour of the coil during quench with a safety discharge. The general behaviour of the coil will be studied. Particular emphasis will be put on the pressure increase in case of a quench which in fact is strongly dependent on the dimensions of the joints and the external cryogenic circuit. The calculations and the resulting design choices for these parts are presented and discussed. An experimental program will be carried out to assess the validity of these choices as a necessary contribution to the ITER design.