In the framework of EU-DEMO superconducting magnets cryogenics and thermalhydraulical analyses, the FAOW-STREAM model developed (CEA/IRFM) has been used to predict the behaviour of TFC WP4 coils (ITER like design with radial plates) in the case of a Fast current Safety Discharge (FSD). This FAOW-STREAM model has already been partly validated by WEST TFC quench and JT-60SA TFC FSD analyses. The paper presents the implementation of the model and calculation inputs, the outcome and results, in particular the increase of helium pressure and temperature (in the coil itself and in exhaust relief circuit), and main conclusion showing effects in the operation domain, in particular a potential quench induced (and at which current) and safety considerations.

In the framework of international development on fusion devices, the tokamak JT-60SA was equipped with superconducting magnets within a Europe-Japan collaboration. After a first integrated commissioning (IC) in 2021, another IC was conducted in 2023 reaching magnet energization levels compliant with 1 MA-class plasmas. The Toroidal Field (TF) coil system was loaded at nominal current and the Poloidal Field (PF) at 50% of nominal current. In the 2023 IC various energization pulse patterns were applied on PF coils to gain experience on the magnets control and protection system in operation conditions. Some tailored current pulses were applied to learn about PF coils coupling parameters and protection system, which sensitivity was found to be improved. In this environment, we developed tools and methods to model and predict the magnets behaviour to possibly identify coils limits in operation conditions. First step is with thermo-hydraulic evaluation of PF coil temperature safety margin evaluation along their pulsed current patterns. Second step is a first order analysis of the TF quench detection system sensitivity, found higher than expected to PF transients, and the contribution to possible mitigation measures to avoid spurious false positive detections. Both steps were benchmarked against experimental data collected during 2023 IC to consolidate models prediction capacities. Foreseen applications aim at checking plasma scenarios repetability rate or avoiding pulsed coil current to generate apparent fault signal in absence of real risk (mostly quench). The tools and methods are described and their rationales explained, together with their intrinsic limits and room for improvement.

The DEMO project is a magnetic-confinement fusion reactor conceived as a technology demonstrator for a fusion power plant. In order to search for a cryogenic and magnetic optimal design such a large and complex project requires dynamic mutli-parameters studies. Thus, regarding a magnet design [1], using Simcryogenics, 0D cryogenic components in the forced-flow cooling loop and 1D Cable-In-Conduct-Conductors (CICC) and pipes are modeled. This code is particularly efficient in performing thermal-hydraulic parametric calculations with pulsed heat loads. This paper presents the exploration study of the Central Solenoid (CS), using the CEA 2022 magnet design [1]. Thermal-hydraulic calculations are performed under pulsed magnetic field operation and variable cryogenic operating conditions keeping constant temperature margin. The 1.5 K temperature margin requirement is obtained by adjusting the inlet temperature for a given pressure head at constant inlet pressure. The resulting data are analyzed through the lens of the global cryogenic cost, considering the expected operating cost of exergy at 4.5 K and the expected cryogenic capital cost. It then allows to estimate the optimal region in term of cryogenic operating parameters.

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

The mission of WEST (tungsten-W Environment in Steady-state Tokamak) is to explore long pulse operation in a full tungsten (W) environment for preparing next-step fusion devices (ITER and DEMO) with a focus on testing the ITER actively cooled W divertor in tokamak conditions. Following the successful completion of phase 1 (2016-2021), phase 2 started in December 2022 with the lower divertor made entirely of actively cooled ITER-grade tungsten mono-blocks. A boronization prior the first plasma attempt allowed for a smooth startup with the new divertor. Despite the reduced operating window due to tungsten, rapid progress has been made in long pulse operation, resulting in discharges with a pulse length of 100 s and an injected energy of around 300 MJ per discharge. Plasma startup studies were carried out with equatorial boron nitride limiters to compare them with tungsten limiters, while Ion Cyclotron Resonance Heating assisted startup was attempted. High fluence operation in attached regime, which was the main thrust of the first campaigns, already showed the progressive build up of deposits and appearance of dust, impacting the plasma operation as the plasma fluence increased. In total, the cumulated injected energy during the first campaigns reached 43 GJ and the cumulated plasma time exceeded 5 h. Demonstration of controlled X-Point Radiator regime is also reported, opening a promising route for investigating plasma exhaust and plasma-wall interaction issues in more detached regime. This paper summarises the lessons learned from the manufacturing and the first operation of the ITER-grade divertor, describing the progress achieved in optimising operation in a full W environment with a focus on long pulse operation and plasma wall interaction.

In the framework of design studies for large fusion magnets, the development of the OLYMPE multi-physic platform is ongoing at CEA. This platform gathers different solvers and codes for addressing the interdependent analyses required for tokamak superconducting magnet design. The present article focuses on the coupling between conductor design, thermohydraulics, and electromagnetism. The corresponding OLYMPE module aims at producing factors of merit such as the amount of superconducting material, together with major parameters impacting the cryogenic system (operating conditions, thermal loads on the magnet). The Thermo-Hydraulic Design (THD) loop was first developed for TF magnet, considering the most critical conductor and a single burn cycle. The loop is improved by extending the model to coil scale, thus including the whole winding pack, and by considering four burn-dwell cycles so as to reach a periodic behaviour. The case of an EU-DEMO TF magnet design is applied, and the comparison of results before and after THD loop upgrade confirms the validity of first loop version for exploratory studies. The second part of the study focuses on the development of CS magnet design module. The dimensioning process specific to CS results in significant differences with the TF design loop. In particular, the CS design loop requires updating the magnetic field distribution at each design iteration, while the assessment of AC losses thermal load requires defining a scenario for the current in each CS module. Consequently, the CS design process merges electromagnetic and thermohydraulic loops. The workflow is applied to an EU-DEMO CS magnet design.

The 500 MJ superconducting toroidal field magnet of TORE SUPRA - WEST has been providing a 4 T magnetic field on the tokamak plasma axis since 1988. The magnet is bath-cooled with superfluid helium at 1.8 K, and operates in steady state each experimental day. During those years, the magnet internal structure has not been modified, but several interfaces and auxiliary systems required for its operation have been upgraded. This paper aims at giving an overview of the Tore Supra-WEST toroidal field magnet system. Main issues and upgrades are described while providing the community a valuable long-life operation feedback and return on experience. We will go through the normal operation sequences and the incidental events, trying to point out key components failures and recovery actions undertaken. We will also develop the main subsystems evolutions such as the quench protection system, the data acquisition and the control systems. We will outline the major risks and mitigation schemes followed as well as the strategies adopted to ensure adequate maintenance, to preserve skills and knowledge continuity over such a long and eventful operating time-frame.

JT-60SA Toroidal Field Coils (TFC) comprise each a Winding Pack (WP), with 6 Cable-In-Conduit Conductor (CICC) lengths wound in 6 double-pancakes operating at 25.7 kA nominal current and 4.7 K temperature as well as stainless steel casing cooled by forced flow channels. Second phase of Integrated Commissioning (IC) have been performed during 2023 with current discharge and induced AC losses in the CICC and eddy currents in the structures. A Fast Assessment of Operating Window (FAOW) model was developed for this purpose and permits to assess heat loads inputs, the heat transfer from casing to WP with inertia effect and time delay, and the cooling capacities (through heat exchanger). This FAOW model is coupled to Superconductors Thermohydraulical and Resistive Electrical Analytical Model (STREAM) and permits to evaluate helium pressure and temperature increase in cooling loop. The IC 18 kA current discharge measurements are compared to calculated results thus validating models and analyses. Predictive calculated results for other current discharge values (15, 18, 20, 23.7 and 25.7 kA) are presented and discussed with the maximal helium pressure limits, and safety aspects (induced relief valves opening or not). These analyses of current values operating windows have been confirmed by further plasma operation at 23.7 kA. Even if the limits of the model are discussed, the main interest consists of rapid execution time duration (only few minutes) compared to numerical codes (several hours). This work and fast analysis tool and model (FAOW-STREAM) can be also useful for further performances analyses in view to recommend their use in similar ITER or EU-DEMO magnets, with possibility of real-time feed-forward calculations/control.

A simplified thermal hydraulic tool dedicated to quench propagation in superconducting magnets was developed. It models 1-D flow and heat transfers in cable-in-conduit conductor coupled with a helium hydraulic network. The model relies on assumptions saving computational resources for engineering design purposes and system performances assessment. It can simulate the quench development for both the early stage before primary quench detection as well as for the current discharge stage. The code robustness and accuracy are investigated on quench simulations of JT-60SA toroidal field coil in nominal tokamak conditions. Furthermore, a comparison with the widely used SuperMagnet code depicting consistent results confirms the simplified assumptions.

EU-DEMO - the European DEMOnstration Fusion Power Plant, is being designed as an intermediate stage between the ITER experimental reactor and the future fusion power plant. The fully superconducting magnet system of the DEMO tokamak includes six Poloidal Field (PF) coils. Two different concepts of the DEMO PF winding packs are being developed by the EPFL-SPC (Switzerland) and CEA IRFM (France) teams. The current cycle of the PF coils consists of the following phases: Premagnetization (10 s), Plasma Current Ramp-Up (PCRU, 80 s), plasma burn (7200 s) and dwell (600 s). The PCRU phase starts with the fast breakdown (0.8 s), during which the operating current and magnetic field change very rapidly causing significant heat generation due to AC losses. Thus, the breakdown is potentially the most critical phase of normal operation of PF conductors regarding the temperature margin. This study was focused on the thermal-hydraulic analysis of the PF coils designed by CEA during breakdown. According to the CEA design, based on the 2018 DEMO reference, the PF coils are double-pancake wound with square NbTi Cable-in-Conduit Conductors with a central cooling channel. Operation of conductors designed for each PF coil was simulated with the THEA code by Cryosoft. Since the detailed reference current breakdown scenario for the PF coils is not available yet, the analysis was performed parametrically. We took into account the external or internal magnetic field profile along each PF conductor as well as heat generation due to the AC hysteresis losses and coupling losses (for the trial values of the parameter nτ = 100, 200 and 300 ms). The study was aimed at the estimation of the minimum temperature margin in each PF conductor during the breakdown and verification if it fulfills the performance criterion min(ΔTmarg) ≥ 1.5 K. The results of analysis should help in further improvements and optimization of the PF winding packs design.

The present study aims at providing quantitative insight, extracted from experimental data, about the impact of Lorentz forces on the distribution of static heat load within JT-60SA Toroidal Field (TF) coils. More accurately, we assess the distribution of these loads over the winding packs (WP) and casings from Helium enthalpy balances when the TF current is either zero or nominal. This study is motivated by the impact of such loads and associated distribution on the quench margin in operation and on the cryogenics system. It is divided in two parts: the first one focuses on the distribution results obtained for the Cold Test Facility (CTF) tests (i.e., in self-field configuration) while the second one focuses on those obtained during JT-60SA integrated commissioning (i.e., in tokamak configuration). We then draw the main outcomes of these studies and discuss the consistency of the results obtained for the two configurations.

Cable-in-conduit conductors (CICC) compose superconducting magnets of tokamaks and are cooled with 4 K supercritical helium forced flow, to provide efficient heat removal and thermal stability of the superconducting state. Nowadays, thermal couplings between fluid and solids are ill-characterized for such cables. We consequently measured global heat transfer interphase coefficients at macro scale. An experimental set-up associated with an inverse method and a direct three temperatures thermal-hydraulic model have been developed to obtain representative data. Transient experiments of forced flow with heat pulses imposing thermal disequilibrium between fluid and solids have been performed. Results are correlated to actual geometric properties obtained from 3D tomography. 1D three temperatures model of the two phases (bundle/nitrogen) along with the jacket, thermally coupled by convection is solved using finite volume method with an implicit and upstream scheme. Results are compared to available literature data.

In the framework of JT-60SA Tokamak commissioning (Japan, 2021), all coils have to be cooled with supercritical helium forced-flow at the temperature of 4.5 K in order to reach superconducting state before energization. An important issue is to predict the Joule energy dissipated in Cable-In-Conduit Conductors (CICC) and the maximal temperature reached in case of an incidental quench occurrence. Therefore, quench simulations were performed on the Toroidal Field Coil (TFC) with the STREAM (Superconductor Thermal hydraulic and Resistive Electrical Analytical Model) code and with the SuperMagnet code (CryoSoft), coupling THEA (thermal hydraulic and electrical 1-D CICC model) and Flower (thermal hydraulic network model). The quench event was simulated for one TFC at nominal conditions of tokamak operation. External energy deposition over each pancake's first turn of the coil at the peak magnetic field location was applied for initiating the resistive transition. This way, the computed Joule energy dissipated by one TFC is conservatively evaluated to be 5.65 MJ. The Joule energy dissipation depends strongly on quench initiation conditions and on the number of fully and rapidly quenched pancakes (maximal quench propagation velocity of 19.5 m/s). Some further analyses were performed on the acceptance quench test realized at the Cold Test Facility (CEA Saclay, 2018) on TFC02. The different calculation results (helium temperature, mass flow rate in upstream and downstream manifolds, normal length propagation) are presented in the following study and are consistent with the measurements. This analysis brings information on the limits and the ability of STREAM code to model quench behaviour in CICC coils cooled by forced flow of supercritical helium. This tool could be useful for tokamak magnets protection during quench event and safe operation, notably thanks to its fast computing time.

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.

We present the pre-concept design of the European DEMO Magnet System, which has successfully passed the DEMO plant-level gate review in 2020. The main design input parameters originate from the so-called DEMO 2018 baseline, which was produced using the PROCESS systems code. It defines a major and minor radius of 9.1 m and 2.9 m, respectively, an on-axis magnetic field of 5.3 T resulting in a peak field on the toroidal field (TF) conductor of 12.0 T. Four variants, all based on low-temperature superconductors (LTS), have been designed for the 16 TF coils. Two of these concepts were selected to be further pursued during the Concept Design Phase (CDP): the first having many similarities to the ITER TF coil concept and the second being the most innovative one, based on react-and-wind (RW) Nb3Sn technology and winding the coils in layers. Two variants for the five Central Solenoid (CS) modules have been investigated: an LTS-only concept resembling to the ITER CS and a hybrid configuration, in which the innermost layers are made of high-temperature superconductors (HTS), which allows either to increase the magnetic flux or to reduce the outer radius of the CS coil. Issues related to fatigue lifetime which emerged in mechanical analyses will be addressed further in the CDP. Both variants proposed for the six poloidal field coils present a lower level of risk for future development. All magnet and conductor design studies included thermal-hydraulic and mechanical analyses, and were accompanied by experimental tests on both LTS and HTS prototype samples (i.e. DC and AC measurements, stability tests, quench evolution etc.). In addition, magnet structures and auxiliary systems, e.g. cryogenics and feeders, were designed at pre-concept level. Important lessons learnt during this first phase of the project were fed into the planning of the CDP. Key aspects to be addressed concern the demonstration and validation of critical technologies (e.g. industrial manufacturing of RW Nb3Sn and HTS long conductors, insulation of penetrations and joints), as well as the detailed design of the overall Magnet System and mechanical structures.

During the integrated commissioning of JT-60SA tokamak, the Toroidal Field (TF) magnet has experienced fast current discharges which have induced AC losses in its winding packs (WP) and in its casings. Using the thermo-hydraulic sensors network installed at the inlet and outlet of the TF coils, we have performed enthalpy balances to estimate the transient heat loads generated by AC losses. In parallel, we have modeled the hysteresis and coupling losses produced in the WP and the eddy currents losses produced in the casings by such events. We then assess and discuss the consistency between these experimental and first theoretical analyses from the energy balance point of view.

In the context of the different fusion machines using superconducting magnets that are currently in development or exploitation phase (DEMO, ITER, JT-60SA…), reliable numerical models are required to perform analyses of the magnet design performances. In the design phase, after a conductor and associated magnet structures are proposed, extensive analyses are carried out to verify the compliance of the design with imposed design criteria (i.e. temperature margin, hotspot temperature). These calculations are also performed during commissioning and exploitation of the machine to ensure that a given scenario will not put the magnet at risk. To meet the need for accurate, fast, and easily usable numerical models for the analysis of magnet systems, the TACTICS code was developed. It consists of a code coupling methodology to perform a pseudo-3D transient co-simulation of coupled physics by linking different codes: THEA for the modeling of the conductors, Cast3M for the magnet structures and Simcryogenics for the cryodistribution simulation. In this paper, the structure of TACTICS is described, detailing the different links and coupling between the codes. The latest developments will be presented, mainly regarding the numerical stability when dealing with fast transients (optimization of the coupling scheme, of the meshing of the structures…). Details will be given on the generation of inputs for the different codes (electromagnetic, thermic, thermohydraulics, cryodistribution) and the post-processing of the results. The paper also presents the application of the tool on the latest CEA design proposal for EU-DEMO TF magnets. The design considered is WP#4 for the DEMO 2018 baseline. It is an “ITER-like” design, based on the double-pancake concept, with the use of radial plates. In burn studies, the minimal temperature margin of the design will be calculated using a TACTICS model accounting for the conductors and the magnet structures. A fast safety discharge simulation of the magnet will be performed, and it will be checked whether or not a quench is initiated on such scenario. Then quench simulations will be performed in order to verify if the hotspot temperature criterion is met.

WEST is an MA class superconducting, actively cooled, full tungsten (W) tokamak, designed to operate in long pulses up to 1000 s. In support of ITER operation and DEMO conceptual activities, key missions of WEST are: (i) qualification of high heat flux plasma-facing components in integrating both technological and physics aspects in relevant heat and particle exhaust conditions, particularly for the tungsten monoblocks foreseen in ITER divertor; (ii) integrated steady-state operation at high confinement, with a focus on power exhaust issues. During the phase 1 of operation (2017-2020), a set of actively cooled ITER-grade plasma facing unit prototypes was integrated into the inertially cooled W coated startup lower divertor. Up to 8.8 MW of RF power has been coupled to the plasma and divertor heat flux of up to 6 MW m(-2) were reached. Long pulse operation was started, using the upper actively cooled divertor, with a discharge of about 1 min achieved. This paper gives an overview of the results achieved in phase 1. Perspectives for phase 2, operating with the full capability of the device with the complete ITER-grade actively cooled lower divertor, are also described.

A large 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 DEMO have been clarified through those studies.