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.

Tore Supra/WEST is a Tokamak operating at CEA/IRFM since 1988. Its toroidal field system is composed of 18 NbTi toroidal field coils (TFC) cooled by superfluid helium bath at 1.8 K. The system is used to operate at full current during every day of a plasma campaign, and is protected by a quench detection system (QDS) which can trigger a fast safety discharge (FSD). In December 2017, an FSD was triggered. After analysis, we could show that it was caused by the quench of TFC9, and that the quench had been triggered by a high flux of runaway electrons at the end of a plasma shot. The detection and protection systems were effective to protect the TF magnet, which is presently back in operation at nominal current without any fault. This paper presents the analyses conducted to determine the origin of the quench, the details of the QDS operation, and the impact of the quench on our cryomagnetic system.

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.

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.

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.

JT-60 SA 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-60 U experiment will be upgraded to JT-60 SA by using superconducting coils. The 18 Toroidal Field Coils (TFC) of the JT-60 SA device are provided by European industry and tested in a Cold Test Facility (CTF) at CEA Saclay. At the summer 2017, 14 TFCs will have been tested at the nominal current of 25.7 kA and at a temperature between 5 and 7.5 K. The main objective of these tests is to check the TF coils performances and hence mitigate the fabrication risks. These tests allowed checking a certain number of performances of the TFCs: 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 show the testing program progress and give an overview of the main experimental results obtained during these tests. The main performances of each coil will be summarized, analyzed, and discussed in the light of the expected TFCs performances.

In the framework of the EUROfusion DEMO project, design studies for the tokamak magnet systems are conducted across several European institutions. Superconducting magnets are required to generate the high magnetic fields needed for the plasma confinement and control. The reference conductor design is based on Cable-In-Conduit Conductors (CICC) cooled at cryogenic temperatures by forced circulation of supercritical helium. The proposed Toroidal Field (TF) coils Winding Pack (WP) design should satisfy the design/safety criteria during operation (e.g., minimal temperature margin) and in off-normal conditions (e.g., hotspot temperature). Quenches are studied to ensure that the proposed conductor design and associated quench protection system guarantee the integrity of the magnet; such events are a safety and investment protection issue, and as such deserve high levels of scrutiny. Quench propagation in a coil is a 3-dimensional (3D) problem. For this reason, a transient pseudo-3D modeling tool was developed for coupled thermal and thermo-hydraulic calculation in a tokamak superconducting coil. The coupling tool is based on a 1D model of the cable using the THEA code, considering current distribution, helium flow, thermal conduction in the strands and propagation of the quench along the conductor; the 2D transverse thermal diffusion across turns is modeled using the Cast3M code, considering the conductor jacket and insulation, on a selected set of cross-sections along the coil. The aim of the analysis is to assess the quench behavior of the CEA proposal for DEMO TF coil. The hotspot temperatures as well as the propagation of the normal zone along the conductor length are evaluated in a realistic quench scenario, taking into account the impact of transverse heat diffusion.

Coupling losses induced in cable-in-conduit conductors (CICC) when subject to a time-varying magnetic field are a major issue commonly encountered in large fusion tokamaks (e.g., JT-60SA, ITER, DEMO). The knowledge of these losses is crucial to determine the stability of CICC but is yet difficult to achieve analytically (thus in a short computation time) given the specific and complex architecture of these conductors although numerical solutions such as THELMA and JACKPOT already exist. In an attempt to ease the resolution of this problem, we have previously presented a theoretical generic study of a group of elements twisted together (representing a cabling stage of a CICC) and derived the analytical expression of its coupling losses. We have now extended this study to a two cabling stage conductor by establishing an analytical model to calculate its coupling losses as function of its effective features. In a second part, we compare our results to these of THELMA and JACKPOT on geometries representing ITER CS and JT-60SA TF conductors. Finally, we have set up a specific algorithm to reconstruct strand trajectories from X-ray images and have extracted the effective geometrical parameters of a JT-60SA TF conductor. Our next objective is then to extract its effective electrical parameters from interstrand resistivity measurements to be able to compare the coupling losses predicted by our analytical model with those measured within the SULTAN facility.

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).

The toroidal field system of the JT-60SA tokamak is composed of 18 NbTi superconducting coils. Each TF coil is composed of six cable-in-conduit conductor lengths, wound in double-pancakes, and carrying a nominal current of 25.7 kA. These coils are being tested in the single-coil configuration at the so-called cold test facility (CEA/IRFU, Saclay, France). The test program includes for all coils a quench test at nominal current obtained by a progressive increase of the operating temperature at helium inlet up to quench (around 7.5 K). Thanks to the accuracy of the voltage measurements, it has been possible to follow the very early development of the quench at the scale of a few millimeters normal length. The paper presents the evolution of the resistive voltages measured on eight coils before the safety discharge is triggered. In addition, this early quench development over one conductor length was also simulated using the THEA code with relevant boundary conditions. Two different quenches, which occurred on two different coils, were analyzed, one starting on a central pancake (peak magnetic field) and the other starting on a side pancake (warmer conductor due to heat transfer from the casing). The simulations show good agreement with experiments, stressing particularly three phases in the quench development clearly identified in the measured resistive voltages of all coils.

In the framework of the DEMOnstration fusion power plant (DEMO) design coordinated by the EUROfusion consortium, a pre-conceptual design of the superconducting magnet system has been developed. For the toroidal field coils (TFCs), three winding pack (WP) options have been proposed; exploring different winding approaches (pancakes vs. layers), and manufacturing techniques (react & wind vs. wind & react Nb3Sn). Thermal-hydraulic and mechanical analyses on the three WPs have produced encouraging results, with some critical issues to be solved in future studies and optimizations. The experimental tests on TF prototype short sample conductors have demonstrated a limited performance degradation with electro-magnetic cycles and significantly lower effective strains than most of the large-size Nb3Sn conductors reported in literature. The toroidal field quench protection circuit has been studied, starting from different topologies and focusing on the most promising one. Two designs are also presented for the central solenoid magnet, with preliminary evaluations on the AC losses during the plasma breakdown. Finally, the design of a TF winding pack based on HTS conductors and the experimental tests on "fusion-relevant" HTS cables are illustrated.

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 à 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.

In the framework of the European-Japanese project JT-60SA, the quench tests are performed for each of the 18 NbTi superconducting tokamak toroidal field coils (TFC) in the Cold Test Facility at the CEA Saclay, Gif-sur-Yvette, France. While launching these experimental quench tests, the conductor current sharing temperature (T-cs) is reached by progressively increasing the inlet helium temperature. The quite complex quench dynamics are then observed due to several coupled physical phenomena influencing the quench propagation. In order to better understand the experimental analyses on coils quench behavior, a numerical model has been used, combining the computation code thermal hydraulic and electric analysis of superconducting cables for the longitudinal quench transient modeling along the cable-in-conduit conductor, and an additional interturn thermal coupling model for transverse heat flux modeling. In this paper, several parametric studies will be done, thanks to simplified numerical simulations in order to identify the predominant physical phenomena driving the JT-60SA TFC quench propagation. The analysis focuses on the linear power impact on the quench initiation, the quench propagation dynamics, and the reverse flow effect.

In the framework of the design activities conducted in the EU for dimensioning the future fusion demonstration reactor (DEMO), extensive analyses were carried out within the framework of the EUROfusion consortium, with the objective of defining the design of the DEMO magnets system. To this aim, CEA has developed ad-hoc predimensioning tools and associated methods in order to size the different magnets: toroidal field coils, central solenoid (CS) and poloidal field (PF) coils. Once predimensioned, the magnet concepts undergo a detailed evaluation procedure with more complex tools and methods. As these detailed analyses are time consuming, the whole design process benefits from an accurate and robust predimensioning process. The predimensioning tools described herein address various aspects driving the operational limits of the magnets while energized in the tokamak, related to electromagnetic, thermal, and mechanic phenomena. In this paper, we expose the latest developments in the predimensioning tools and the methods employed for obtaining a rapid and reliable macroscopic semi-analytical representation of the magnets. The assessment of the predimensioning tools on reference configurations (e.g., ITER) is also described. The application of these design tools on the DEMO configuration issued by EUROfusion is presented, and the resulting magnet design proposals are summarized: First, the PF system, composed of six coils, with one PF selected for study here. Second, the CS, with a modular geometry (five modules) and pancake winding. The approach here assesses the resilience of the CS to a fast-transient (breakdown). The outcomes of a sensitivity study on parameters/criteria are discussed, and extended to some tentative recommendations on the design approach for the DEMO PF and CS magnets. For the PF, a design is obtained for the PF5 coil following a dimensioning methodology consistent with the ITER PF system design with the same level of definition of the plasma operating scenario. As for the CS, it is stressed that consideration of the internal field (from shielding) substantially affects temperature margin results.

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. (C) 2017 Published by Elsevier B.V.

In the framework of the EUROfusion DEMO project, studies are conducted in several European institutions for designing the tokamak magnet systems. In order to generate the high magnetic fields required for the plasma confinement and control, the reactor should be equipped with superconducting magnets, the reference design being based on Cable-In-Conduit Conductors (CICC) cooled at cryogenic temperatures by forced circulation of supercritical helium. In order to propose a toroidal field (TF) winding pack (WP) design compatible with DEMO requirements, CEA has developed several tools addressing the different areas related to magnet dimensioning. An accurate calculation of magnetic field along the conductors is provided by the TRAPS code, and conductor design is performed by using an integrated macroscopic home design code based on simplified models accounting for superconducting properties, mechanics and thermal. This multi-physic tool gives a realistic but not assessed design. Indeed it is based on an assumed operating temperature that must be validated with an elaborate code, since it is linked with temperature margin design criterion (1.5 K). A dedicated modelling tool was developed by coupling the THEA code for 1D thermo-hydraulics in cables and the Cast3M code for 2D transverse thermal diffusion in a limited number of coil cross-sections, enhancing the accuracy-of the outputs as being a quasi-3D approach. This tool allows a better assessment of the flux exchange between WP and casing, and the modelling of inter-turn and inter-pancake thermal coupling. The coupling methodology is described, as well as its validation on the simulation of a heat exchanger. A calculation was performed on the CEA proposal for DEMO TF coil in a burn (steady state) scenario, and finally providing a realistic assessment of the temperature margin. (C) 2017 Elsevier B.V. All rights reserved.

In order to check the performance of the JT-60SA Toroidal Field (TF) coils and hence mitigate their possible fabrication risks, a series of tests have been carried out in the Cold Test Facility (CTF) at CEA Saclay in nominal conditions at 5 K and 25.7 kA. One major test performed is the so called "temperature margin test" during which the inlet helium temperature of the Winding Pack (WP) is controlled to increase progressively to the calculated current sharing temperature of 7.3 K in the high field region to launch the quench. The measurements of voltage, pressure and temperature signals allow us to study the thermohydraulic behavior of the JT-60SA coils during a quench. This paper will present the 6 quench tests on 5 TF coils accomplished between February and September 2016, of which the first manufactured coil TFC10 has carried out two quench tests and two pure Fast Discharges without any quenches. It will analyze in particular the temperature and pressure rise in helium, the corresponding physical phenomena after each voltage variation, the four quench dynamic phases, the quench propagation velocity and the dissipated power in the coils. (C) 2017 Elsevier B.V. All rights reserved.

While the ITER Tokamak is proceeding with major construction and assembly phases, design and R&D activities regarding the next step experimental reactor, the so-called DEMO machine, are underway in Europe. In particular, the magnet systems enter a refined design phase, under the general guidelines of Eurofusion, where winding packs (WP) are defined, conductor designs are proposed, and subscale analyses are performed in order to crosscheck the systems code (PROCESS) predimensioning. This paper summarizes the design iterations led at CEA on the toroidal field (TF) and central solenoid (CS) magnet systems. It presents, in particular, the optimization of the WP architecture with regard to the filling of the available space (TF) and flux maximization (CS). The paper gives the relevant electrical, thermal, and mechanical parameters for each system design, and tries to give insight on the tradeoffs and possible changes that could be implemented, given some margin on fixed systems code inputs. Finally, it also includes the refined magnetic field maps calculations performed to confirm the design operating field and explore other scenario time points that include poloidal field and plasma contributions.

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. (C) 2017 Published by Elsevier B.V.

Cable-in-conduit conductors (CICCs) are composed of a large number of strands (superconducting composites and copper strands) twisted together in several stages with different twist pitches. They are widely used in large fusion tokamaks such as JT-60SA or ITER. However, because of their complex transposed geometry at a strand scale, the knowledge of ac coupling losses in these conductors is limited and still has some improvement margins to capture its complexity while the prediction of their behavior under transient regimes (e.g., central solenoid) is of first importance to assess a safe operation in tokamaks. Consequently, we have carried out an in-depth theoretical generic study of a single stage of a CICC and analytically derived the expression of coupling losses using physical parameters (time constant and partial shielding coefficient) determined from electromagnetic and geometrical properties. Our approach has been inspired by the MPAS model (extensively used on the experimental ITER database) but starts from the analytical description of a single stage and aims at reaching the CICC scale in an iterative way.