D. Ciazynski’s research while affiliated with Atomic Energy and Alternative Energies Commission and other places

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 the present work at improving predimensioning process reliability for the final benefit of the whole design workflow. The material exposed in the present paper is the path to this improvement achieved through a sery of predimensioning 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 FEAs, 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 (WP)

In the framework of the EU-JA Broader Approach agreement, the 20 Toroidal Field (TF) coils (included 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 standardised 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 assessed it stands as a 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.

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 design activities conducted in EU for dimensioning the future fusion demonstration reactor (DEMO), extensive analyses were led through in the EUROfusion context, aiming at defining the design of the DEMO magnets system. CEA has developed ad-hoc pre-dimensioning tools and associated methods in order to size the different magnets: Toroidal Field (TF) coils, Central Solenoid (CS) and Poloidal Field (PF) coils. Once pre-dimensioned the magnet concepts undergo an evaluation process with detailed analyses. The tools address various aspects driving the operational limits of the magnets while energized in the tokamak, related to electromagnetic, thermic and mechanic phenomena. Here we present the latest developments on the pre-dimensioning tools and their assessment on reference configurations (ITER). The application of the above described tools on the DEMO configuration will be shown, and their outcome proposals for: - The PF system, composed of 6 coils, design study being focused on one selected PF. - The CS, considering modular geometry and pancake winding. The approach concerns the check of CS resilience in a fast-transient step (breakdown). Discussions on the outcomes of a sensitivity study from parameters / criteria is provided with some tentative recommendations on DEMO PF and CS magnets design.

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.

The Toroidal Field (TF) system of JT-60SA tokamak comprises 18 NbTi superconducting coils. In each TF coil (TFC), 6 Cable-In-Conduit Conductor (CICC) lengths are wound into 6 double-pancakes (DP) 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 (CTF, CEA Saclay), up to quench. A SuperMagnet (CryoSoft) model has been developed, each of the 12 pancakes being modelled with THEA and cryogenic circuit being modelled 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 which 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 6 cable-in-conduit conductor lengths, wound in double-pancakes (DP). 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. Thanks to the accuracy of the voltage measurements it has been possible to follow the 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. This early quench development over one conductor length was also simulated using the THEA code with relevant boundary conditions. Two different quenchs (two different coils) were analyzed: one starting on a central pancake (peak magnetic field), the other 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.