Toroidal Field Magnet System Research Articles

Divertor Tokamak Test facility project: status of design and implementation

Abstract An overview is presented of the progress since 2021 in the construction and scientific programme preparation of the Divertor Tokamak Test (DTT) facility. Licensing for building construction has been granted at the end of 2021. Licensing for Cat. A radiologic source has been also granted in 2022. The construction of the toroidal field magnet system is progressing. The prototype of the 170GHz gyrotron has been produced and it is now under test on the FALCON facility. The design of the vacuum vessel, the poloidal field coils and the civil infrastructures has been completed. The shape of the first DTT divertor has been agreed with EUROfusion to test different plasma and exhaust scenarios: single null, double null, XD divertor and negative triangularity plasmas. A detailed research plan is being elaborated with the involvement of the EUROfusion laboratories.

DEMO toroidal field coil fast discharge unit integration studies

The Fast Discharge Units (FDUs) of the Superconducting (SC) Toroidal Field (TF) coils in the European demonstration fusion power plant DEMO warrant the machine integrity over its full lifetime against severe failure events, such as SC coil quenches or any other plant events requiring the safe TF magnet system discharge. A low (75 kA) and a high current (105 kA) configuration are under study for the TF coils for DEMO. The FDUs must be extremely reliable for the purpose of commutating in short time (∼1 s) the currents and to discharge the TF magnets safely into resistors outside of the tokamak building. Malfunctions of the FDUs must be avoided. The FDUs are considered as Safety Important Class (SIC) components that need to discharge high amounts of energy of about 161 GJ (@75 kA) resp. 118 GJ (@105 kA) stored in the DEMO TF coils.The TF FDUs Circuit Breakers (CBs) shall be installed in the lower level of the tokamak to minimize the length of the connecting busbars. The FDUs integration is challenging because of the high neutron and gamma radiation and stray magnetic fields of the tokamak.Since in DEMO the neutron fluence over lifetime is much higher than in ITER, the problems of using FDUs with electronic subsystems was expected to be more severe, so that their integration has been considered from the beginning of the DEMO project. Sufficient shielding or possible re-positioning of the whole FDUs or sensitive FDU components compared to ITER are being investigated, to reduce the neutron fluxes and neutron and gamma ray fluences. Alternative concepts, e.g., fully mechanical CBs are studied in the EUROfusion Work Package Plant Electrical System (WPPES) in parallel.This paper presents the CAD integration work on the DEMO TF FDUs supported by neutronics assessments. It is assumed the same FDU technology as in ITER. The magnet feeder´s integration is commenced at the same time.

Final design of the DTT Toroidal power supply circuit

The Divertor Tokamak Test (DTT) facility is in advanced design and construction phase at the ENEA laboratories in Frascati, Italy, to contribute to the study, the design and assessment of systems for the treatment of heat exhaust. The DTT Toroidal Field (TF) magnet system consists of 18 superconducting coils, fed in series by a single power supply (TFPS), which is a 24-pulses thyristor-based converter with an output DC current up to of 44 kA. In case of a quench in the superconductors, the magnetic energy stored in the coils is extracted by 3 Fast Discharge Units (FDUs) in series with 3 sectors of 6 TF coils. After the respective contracts were awarded, the TFPS and the FDUs are currently in the final design phase which is addressed in this paper. In particular, the analysis is focused on the most innovative aspects of the design, as the use of silicon carbide (SiC) varistors instead of discharge resistors and of a fully electronic and completely redundant high-current switch without any mechanical device.

Optimization of the overall Toroidal Field Coil cryomagnetic system at the pre-conceptual design phase of the European DEMO fusion reactor

Different superconducting conductor design options are considered for the Toroidal Field (TF) Coil magnets of the future European DEMO reactor. Four TF Winding Pack (WP) options have been proposed and consist of two layer-wound concepts and two pancake-wound concepts, each of them having specific conductor design. These TF WPs are cooled by supercritical helium forced flow driven by a cold circulator at about 4 K. At the pre-conceptual design stage, it is particularly relevant to also consider cryogenic system merits while designing the TF magnet system. Indeed, both cost drivers such as the superconductor amount and the refrigeration loads depend on the cooling parameters: temperature, pressure, mass flow/pressure drop across the conductor. Hence, these parameters shall be optimized in order to minimize the total cost for the cryomagnetic system. For the pancake-wound option designed by CEA, a new approach that combines the magnet design calculation and the refrigeration load calculation using CEA tools (an adapted version of MADMACS-TF code and the simulation library Simcryogenics) was implemented. An optimization problem that considers both magnet and cryogenic merits was assessed. This problem consists in optimizing the superconducting amount and the refrigeration loads while ensuring that the temperature margin of the superconductor remains at 1.5 K. This article presents the method for cost minimization and the information workflow between the different codes, as well as preliminary results. This method is relevant to optimize simultaneously the design of the magnets and the cryogenic system and to make comparisons between the different conductor designs during the next phase of EU DEMO, moving forwards to the conceptual design. With such a method, recommendations can be made to ensure that the combined investment and energetic consumption cost is kept as low as possible.

Design updates and analysis of the Korean fusion demonstration reactor superconducting toroidal field magnet system

Abstract After a conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) was initiated in 2012, the preliminary design of superconducting magnet system which is one of the key components of the K-DEMO has been done in 2015. The superconducting magnet system of K-DEMO has 16 Toroidal Field (TF) coils, 8 Central Solenoids (CS) modules and 12 Poloidal Field (PF) coils which use Cable-In-Conduit Conductors (CICC). For a high 7.4 T magnetic field generation at plasma center, high performance Nb3Sn based superconductor will be developed and used for TF magnet. Starting from the preliminary conceptual design, design updates of K-DEMO TF magnet is being conducted for checking engineering issues. For a validation of the magnet design, thermo-hydraulic stability and mechanical stability are being considered. In this work, the conceptual design studies for updates of K-DEMO TF magnet system is described.

Structural Assessment Procedure of the Toroidal Field Magnet System for the Divertor Tokamak Test

The Divertor Tokamak Test facility (DTT) is an experimental tokamak machine to be built in Frascati, Italy, at the ENEA research center. During its development, the DTT has gone through several design iterations. Developing a rigorous finite-element methodology to evaluate the performance of all its components has thus been a critical part of the verification phase of each new update. This work summarizes the assessment procedure that currently supports the design of the Toroidal Field magnet system, including the superconducting winding pack (Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn), the casing structure and all of the inter-coil structures. Given the high complexity of the 3D structure to reproduce, we implemented some modeling simplifications to solve the problems. Here we describe the finally adopted methodology next to all the motivations we examined to make sure the results were sound and reliable from an engineering point of view.

Qualification of the U.S. Conductors for ITER TF Magnet System

The U.S. Domestic Agency is one of the six suppliers of the toroidal field (TF) conductor for ITER. To qualify conductors per ITER requirements, we prepared or provided to the Swiss Plasma Center (SPC) eight test articles, sixteen conductors total, that were tested in the SULTAN facility at SPC in the ITER relevant conditions. The strands that were used in these SULTAN samples were fully characterized in the laboratories. In this paper, we report both test results and analysis of the conductors’ performance against expected strand performance. The U.S. TF conductors showed a slightly better than average current sharing temperature and a relatively low sensitivity to warm up-cooldown cycles in comparison with other suppliers’ conductors. However, the trend in current sharing temperature versus cycles and warm ups did not saturate, which means a continuing slow degradation of the conductor performance if the number of warm up and cooldown cycles will be significantly higher than expected. The ac losses in the U.S. TF conductors are in line with losses in the other TF conductor suppliers.

Progressing in cable-in-conduit for fusion magnets: from ITER to low cost, high performance DEMO

The performance of ITER toroidal field (TF) conductors still have a significant margin for improvement because the effective strain between −0.62% and −0.95% limits the strands’ critical current between 15% and 45% of the maximum achievable. Prototype Nb3Sn cable-in-conduit conductors have been designed, manufactured and tested in the frame of the EUROfusion DEMO activities. In these conductors the effective strain has shown a clear improvement with respect to the ITER conductors, reaching values between −0.55% and −0.28%, resulting in a strand critical current which is two to three times higher than in ITER conductors. In terms of the amount of Nb3Sn strand required for the construction of the DEMO TF magnet system, such improvement may lead to a reduction of at least a factor of two with respect to a similar magnet built with ITER type conductors; a further saving of Nb3Sn is possible if graded conductors/windings are employed. In the best case the DEMO TF magnet could require fewer Nb3Sn strands than the ITER one, despite the larger size of DEMO. Moreover high performance conductors could be operated at higher fields than ITER TF conductors, enabling the construction of low cost, compact, high field tokamaks.

AC loss, interstrand resistance and mechanical properties of prototype EU DEMO TF conductors up to 30 000 load cycles

Two prototype Nb3Sn cable-in-conduit conductors conductors were designed and manufactured for the toroidal field (TF) magnet system of the envisaged European DEMO fusion reactor. The AC loss, contact resistance and mechanical properties of two sample conductors were tested in the Twente Cryogenic Cable Press under cyclic load up to 30 000 cycles. Though both conductors were designed to operate at 82 kA in a background magnetic field of 13.6 T, they reflect different approaches with respect to the magnet winding pack assembly. The first approach is based on react and wind technology while the second is the more common wind and react technology. Each conductor was tested first for AC loss in virgin condition without handling. The impact of Lorentz load during magnet operation was simulated using the cable press. In the press each conductor specimen was subjected to transverse cyclic load up to 30 000 cycles in liquid helium bath at 4.2 K. Here a summary of results for AC loss, contact resistance, conductor deformation, mechanical heat production and conductor stiffness evolution during cycling of the load is presented. Both conductors showed similar mechanical behaviour but quite different AC loss. In comparison with previously tested ITER TF conductors, both DEMO TF conductors possess very low contact resistance resulting in high coupling loss. At the same time, load cycling has limited impact on properties of DEMO TF conductors in comparison with ITER TF conductors.

High-Performance Superconductors for Fusion Nuclear Science Facility

High-performance superconducting magnets play an important role in the design of the next step large-scale, high-field fusion reactors such as the fusion nuclear science facility (FNSF) and the spherical tokamak (ST) pilot plant beyond ITER. Princeton Plasma Physics Laboratory is currently leading the design studies of the FNSF and the ST pilot plant study. ITER, which is under construction in the south of France, utilizes the state-of-the-art low temperature superconducting magnet technology based on the cable-in-conduit conductor design, where over a thousand multifilament Nb 3 Sn superconducting strands are twisted together to form a high-current-carrying cable inserted into a steel jacket for coil windings. We present design options of the high-performance superconductors in the winding pack for the FNSF toroidal field magnet system based on the toroidal field radial build from the system code. For the low temperature superconductor options, the advanced J c Nb 3 Sn RRP strands (J c > 1000 A/mm 2 at 16 T, 4 K) from Oxford Superconducting Technology are under consideration. For the high-temperature superconductor options, the rectangular-shaped high-current HTS cable made of stacked YBCO tapes will be considered to validate feasibility of TF coil winding pack design for the ST-FNSF magnets.

Russia's Contribution to the ITER TF Magnets

The ITER toroidal field (TF) magnet system consists of 18 D-shaped coils, each of which is formed by five regular double pancakes (rDP) and two side double pancakes (sDP). Manufacturing of conductors for the TF coils, which are Nb 3 Sn-based cable-in-conduit conductors, has been shared among six ITER Domestic Agencies, including the Russian Federation (RF). The RF share includes manufacturing of eighteen 760-m conductors for rDP and eight 415-m conductors for sDP, i.e., almost 4 out of 18 TF coils will be produced from the conductors supplied by RF. In order to fulfil project obligations, facilities for Nb 3 Sn strand, cable, and conductor production had been established, and after successful qualifications, the manufacturing phase began. The manufacturing activities went on for six years, starting with the first strand production in 2009 and finishing with a successful wrap-up of conductor manufacturing and delivery in 2015. In this paper, highlights of the TF conductor production from the superconducting strand manufacturing up to the results of conductor full-size sample tests in the SULTAN facility are presented.

Overview of Progress on the EU DEMO Reactor Magnet System Design

The DEMO reactor is expected to be the first application of fusion for electricity generation in the near future. To this aim, conceptual design activities are progressing in Europe (EU) under the lead of the EUROfusion Consortium in order to drive on the development of the major tokamak systems. In 2014, the activities carried out by the magnet system project team were focused on the toroidal field (TF) magnet system design and demonstrated major achievements in terms of concept proposals and of consolidated evaluations against design criteria. Several magnet system RD and 2) the HTS R&D activities building upon the consolidated knowledge acquired over the past years.

ITER TF Magnet system analysis in faulted conditions

The ITER toroidal field (TF) magnet system must be designed, built, and operated, so that failures, which could occur under off-normal conditions, cannot cause damage to the confinement barriers. In fact, even if the magnets are not part of the barrier, the coils store a relevant amount of energy, which, in case of failure, may lead to large displacement of the coils, with the risk of a crash against confinement structures, which cannot be damaged. In this paper, the outcome of a structural analysis of the ITER TF coil magnet system, simulating possible faulted condition scenarios, is described. Two possible electrical failures of the TF coil system are considered: a short, resulting in a relevant current peak in one coil, and a quench without the discharge of the other ones. Since in both scenarios the assumption of cyclic symmetry condition is not valid anymore, a complete 18-coil finite-element (FE) model of the whole ITER TF magnet system has been developed for this purpose. By means of FE models, both electromagnetic and mechanical behaviors have been simulated, determining the magnetic field distribution; the forces acting on the coils and, consequently, stress; the displacement; and the forces on the interfaces. In particular, the mechanical model features a detailed description of all the subsystems which play a structural role in the TF magnet system, such as the intercoil structures, the poloidal shear keys, the TF wedged vault, the TF gravity support, the precompression ring system, and the poloidal field support structures. The model also includes all the bolted connection and all the nonlinearities (pretension, contact, etc.) characterizing the mechanical system. The results show that the deformations of the structure and the loads on the subsystems are within the allowable limits at off-normal conditions, thus demonstrating the soundness of the design, even in the failure scenarios.

EU ITER TF coil: Dimensional metrology, a key player in the Double Pancake integration

The ITER Toroidal Field (TF) magnet system consists of 18 “D” shaped coils. Fusion for Energy (F4E), the European Domestic Agency for ITER, is responsible for the supply of 10 out the 19 TF coils (18 installed plus one spare coil). Each TF coil, about 300 t in weight, is made of a stainless steel case containing a Winding Pack (WP).The European manufacturing of the Radial Plates (RPs) and WPs has been awarded to two different industrial partners, whose activities are strongly linked with each other. In order to manufacture a Double Pancake (DP), first, the conductor has to be bent onto a D-shaped double spiral trajectory, then heat treated and inserted in the grooves of the RP. This represents the most challenging manufacturing step: in order to fit inside the groove, the double spiral trajectory of the conductor must match almost perfectly the trajectory of the groove, over a length above 700m. In order to achieve this, the conductor trajectory length must be controlled with an accuracy of 1mm over a length of 350m while the radial plate groove has to be machined with tolerances of ±0.2mm over dimensions of more than 10m. In order to succeed, it has been essential to develop a metrology process capable to control with high accuracy both the DP conductor and the RP groove trajectories.This paper reports on the work carried out on the development and qualification of the dimensional metrology to monitor the manufacturing of the conductor. Reference is made to the final dimensional check of the RP focusing on the groove centreline length. In addition the results obtained on the one to one scaled prototype DP are described. Finally, the strategy and foreseen improvements for the production of DPs are discussed.

Optimal design of a toroidal field magnet system and cost of electricity implications for a tokamak using high temperature superconductors

The potential for reducing the Cost of Electricity (CoE) by using High Temperature Superconductors (HTS) in the Toroidal Field (TF) coils of a fusion tokamak power plant has been investigated using a new HTS module in the PROCESS systems code. We report the CoE and the design of HTS tokamaks that have been optimised by minimising the major radius of the plasma. Potential future improvements in both the superconducting properties and the structural materials for TF coils operating at 4.8K and 30K are considered. Increasing the critical current density by a factor of 10 (with a commensurate reduction in costskA−1m−1) results in a CoE 4.4% less than equivalent tokamaks using current low temperature superconductors (LTS). If the yield strength of the TF casing material is increased by 40% to 1400MPa, the CoE is further reduced by 3.4%. Implementing both improvements and operating the TF coils at 4.8K leads to CoE of 19.1 (10.1) €centkW−1h−1 for a 500MW (1.5GW) HTS reactor compared to 20.7 (11.1) €centkW−1h−1 for an LTS reactor (2013 costs). Operating the HTS TF coils at 30K with both improvements, gives a similar CoE for HTS and LTS tokamaks.

Extended Quality Control Process Applied on TF Strand and TF Conductor Production for JT-60SA Project

In the framework of the JT-60 Super Advanced (JT-60SA) project, as part of its contribution to the Broader Approach Agreement, Europe provides the full toroidal field (TF) magnet system. For this purpose, Fusion for Energy (F4E) is procuring 27 km of TF conductor, which is of cable-in-conduit type, including 486 strands (2/3 NbTi and 1/3 copper) embedded into a rectangular stainless steel jacket. Since the start of procurement in 2010, the totality of strand has been manufactured, and more than half of the total conductor unit lengths have been accepted by F4E. In the course of this material fabrication, a large quality control (QC) program was established to consolidate the robustness of the quality diagnostic of both productions' follow-up. As for the TF strand production, a comprehensive campaign of critical properties evaluation was led to ensure that the T CS (temperature of current sharing) value in the operation condition is compliant with a 1-K temperature margin in the JT-60SA tokamak. From the large amount of data collected over the two-year production period, a statistic analysis is presented in the paper, together with the methodology that was used to build it. As for TF conductor production, the QC database provided hydraulic properties at low Re. In this framework, a covering QC campaign was led to crosscheck the acceptance data and, at the same time, quoting the reliability of acceptance data exploitation for predictive mass flow analyses in operation. An extra QC, including material analyses and quantitative energy-dispersive X-ray (EDX) analysis of both cable and jacket, was also performed. Highlights are mainly put to the scientific approach of the present topics investigated, but together, some considerations on the risk approach are also presented.

Quench characterization and thermo hydraulic analysis of SST-1 TF magnet busbar

Toroidal field (TF) magnet system of steady-state superconducting tokamak-1 (SST-1) has 16 superconducting coils. TF coils are cooled with forced flow supercritical helium at 0.4MPa, at 4.5K and operate at nominal current of 10,000A. Prior to TF magnet system assembly in SST-1 tokamak, each TF coil was tested individually in a test cryostat. During these tests, TF coil was connected to a pair of conventional helium vapor cooled current leads. The connecting busbar was made from the same base cable-in-conduit-conductor (CICC) of SST-1 superconducting magnet system. Quenches experimentally observed in the busbar sections of the single coil test setups have been analyzed in this paper. A steady state thermo hydraulic analysis of TF magnet busbar in actual SST-1 tokamak assembly has been done. The experimental observations of quench and results of relevant thermo hydraulic analyses have been used to predict the safe operation regime of TF magnet system busbar during actual SST-1 tokamak operational scenarios.

Completion of TF Strand Production and Progress of TF Conductor Manufacture for JT-60SA Project

In the framework of the JT-60SA project, aiming at upgrading the present JT-60U tokamak, Europe, as part of its in-kind contribution within the Broader Approach agreement, will provide the toroidal field (TF) magnet system. For this purpose, Fusion for Energy is committed to procure about 29 km of TF conductor. The TF conductor is cable-in-conduit type and includes 486 strands (2/3 NbTi-1/3 copper) wrapped with a thin stainless steel foil and compacted into a rectangular stainless steel jacket. The procurement is split into two main contracts: one for TF strand manufacturing and the other for TF conductor cabling and jacketing. TF strand manufacture was completed while the TF conductor one is being finished. In the present paper, we draw an overview of both productions emphasizing on the quality control (QC) approaches and on aspects relevant to risk management of the JT-60SA tokamak operation. For the NbTi strand, the complete production overview is provided including extensive statistical considerations on NbTi strand critical performances ( IC, TCS). For the TF conductor, the overview also deals with collected results from acceptance tests and peripheral tests led for consolidating the QC (hydraulic tests, nondestructive examination, full-size sample cold tests in SULTAN facility).

Design and implementation of quench detection instrumentation for TF magnet system of SST-1

Steady State Superconducting Tokamak-1 (SST-1) at Institute for Plasma Research (IPR), India is now in engineering validation phase. The assembled Toroidal Field (TF) magnet system of SST-1 will be operated at 10kA of nominal current at helium cooled condition of 4.5K. A reliable and fail proof quench detection (QD) system is essential for the safety and the investment protection requirements of the magnets. This QD system needs to continuously monitor all the superconducting coils, which include 16 TF magnets, return-loop, bus bars and current leads. In case of any event initiating the normal resistive zone and reaching thermal run-away, the QD system needs to trigger the magnet protection circuits. Precision instrumentation and control system with 204 signal channels had been developed for detection of quench anywhere in the entire TF magnet system. In the present configuration of quench detection scheme, the voltage drop across each double pancake (DP) of each TF coil are compared with its two adjacent DPs for the detection of normal zone and cancelation of inductive couples. Two identical redundant systems with one out of two configurations are successfully commissioned and tested at IPR. This paper describes the design and implementation of the QD system, Installation experience, validation test and initial results from the recent SST-1 magnet system charging.

Thermo hydraulic and quench propagation characteristics of SST-1 TF coil

SST-1 toroidal field (TF) magnet system is comprising of sixteen superconducting modified ‘D’ shaped TF coils. During single coil test campaigns spanning from June 10, 2010 till January 24, 2011; the electromagnetic, thermal hydraulic and mechanical performances of each TF magnet have been qualified at its respective nominal operating current of 10,000A in either two-phase or supercritical helium cooling conditions. During the current charging experiments, few quenches have initiated either as a consequence of irrecoverable normal zones or being induced in some of the TF magnets. Quench evolution in the TF coils have been analyzed in detail in order to understand the thermal hydraulic and quench propagation characteristics of the SST-1 TF magnets. The same were also simulated using 1D code Gandalf. This paper elaborates the details of the analyses and the quench simulation results. A predictive quench propagation analysis of 16 assembled TF magnets system has also been reported in this paper.