International Thermonuclear Experimental Reactor Magnet Research Articles

Conceptual Design of Isolated Power for Quench Detection System With Highly Insulating Stability Under Super High Field Magnets Using Wireless Power Transfer Technology

The super high field magnets using superconducting wires have been promisingly applied for thermonuclear fusion power generation such as Korea Superconducting Tokamak Advanced Research (KSTAR) and International Thermonuclear experimental reactor (ITER) systems. The quench detection system (QDS) is essential for high reliability in the super high field magnet facilities. Generally, the quench voltage detection system, which uses resistive voltages, has been adopted. Such an ITER magnet, the maximum operation current and voltage of superconducting keeps over 60kA and 50 kV DC. Generally, in the case of medium power, the power transformer has been adopted for QDS as an isolated power due to stable insulating and durable characteristics. However, unfortunately, the magnetic saturation of electric steel for power transformer remained fragile as well as the volume is increased under super high field magnet. From this reason, authors suggest the electron quench detection device (EQDD) module. The main functions of EQDD are to convert high voltage signals of magnet into voltage-drop, and then send into quench monitoring system through optic cable. The EQDD module needs to isolate power system since EQDD module should be operated under high voltage insulation. From this reason, authors apply the wireless power system for EQDD module as a reasonable option instead of power transformer since the wireless power system can transfer power through any non-metallic media between resonance antenna (Tx) and receiver (Rx) coils. In this study, authors describe the possibility of conceptual design for quench detection system with high insulation under super high field magnet using wireless power transfer technology. As the EQDD including isolated wireless power should be electrically shielded, the shielding effects for Tx coil is investigated under different intervals of shielding barrier. Additionally, to expand the delivery distance between Tx and Rx coils, the operation and thermal characteristics for transfer efficiency are investigated under 370 kHz and 750 kHz RF generators, respectively.

Research progress on the mechanical behavior of the cable in conduit conductor for the international thermonuclear experimental reactor project

The international thermonuclear experimental reactor (ITER) program is of the largest and the most influential international cooperation project, and its purpose is to verify the scientific and technical feasibility of the magnetic constraint nuclear fusion reactor. The high temperature plasma with billion degrees Celsius can be constrained in a magnetic cage, which can provide a strong magnetic field by a Tokamak device wounded by cable in conduit conductor (CICC). In 2014, Dr. Devred A, a chairman of ITER magnet project, pointed out that the ITER CICC is mainly faced with three challenges: (1) During insertion into the jacket assembly, the cable exhibits a tendency to rotate under the action of the pulling force. That may increase the twist pitch, especially for the final one. Cable twist pitches must be controlled to prevent excessive AC losses in the CICC, which is a threaten to the stability of CICC. The elongation of the twist pitch must be settled. (2) The CICC′s current sharing temperature ( T cs) showing degradation after the electro-magnetic (EM) and thermal cycling load, that means the Tokamak can only run thousands of times, less than the original design of 30000 times, raising risks at the fusion project. (3) Find the way to fabricate the high-performance superconducting wire and the low resistance joint. As we all know, the central and the toroidal field solenoid superconducting cable of the Tokamak were made of Nb3Sn superconducting strands, and the superconducting cable will working at the mechanical–thermal–electrical–magnetic fields environment. Previous studies have shown that Nb3Sn superconductivity is sensitive to the mechanical deformation. The critical current of the superconductor wires will show a significant degradation with the deformation under the loads of tension, pressure and twist, which increases the risk of ITER Tokamak directly. Therefore, it′s important for the design of the Tokamak magnet system to investigate the equivalent mechanical parameters of the cable and its mechanical behavior under the action of multiple fields. In this paper, several key mechanical problems such as the equivalent mechanical parameters of the superconducting cable, the untwisting behavior in the process of insertion, and the T cs degradation under the thermo-electromagnetic cyclic loads have been briefly reviewed. Firstly, the stress-strain curve of the triplet was analyzed based on the thin-rod model and the tensile stiffness model of the triplet was established. Secondly, on the basis of the triplet model, a complex model of the tensile stiffness and the equivalent CTE of the 3×3 strand was build, and the derivation process was provided. Thirdly, the untwisting behavior of the cable in CICC fabrication was investigated. The bending stiffness model of the petal under the wrapping was established, and the tensile untwisting model of the superconducting cable was also build. Fourthly, a mechanical model was established for the T cs degradation mechanism of TF and CS CICC conductors under the action of thermo-electromagnetic cyclic load. Based on the compatible relationship between the transverse compressive strain and the axial elongation of the superconducting cable, the impacts of axial compressive stiffness and surface friction coefficient on the axial compression strain of the cable under the thermo-electromagnetic cyclic load were studied. Finally, we summarized the existing problems and the future research points on the basis of the previous research results, which will help our researchers to catch the progress of the ITER program. In the future, the CICCs will also be used in China′s large superconducting magnets.

Magnet Design Considerations for Fusion Nuclear Science Facility

The Fusion Nuclear Science Facility (FNSF) is a nuclear confinement facility that provides a fusion environment with components of the reactor integrated together to bridge the technical gaps of burning plasma and nuclear science between the International Thermonuclear Experimental Reactor (ITER) and the demonstration power plant (DEMO). Compared with ITER, the FNSF is smaller in size but generates much higher magnetic field, i.e., 30 times higher neutron fluence with three orders of magnitude longer plasma operation at higher operating temperatures for structures surrounding the plasma. Input parameters to the magnet design from system code analysis include magnetic field of 7.5 T at the plasma center with a plasma major radius of 4.8 m and a minor radius of 1.2 m and a peak field of 15.5 T on the toroidal field (TF) coils for the FNSF. Both low-temperature superconductors (LTS) and high-temperature superconductors (HTS) are considered for the FNSF magnet design based on the state-of-the-art fusion magnet technology. The higher magnetic field can be achieved by using the high-performance ternary restacked-rod process Nb 3 Sn strands for TF magnets. The circular cable-in-conduit conductor (CICC) design similar to ITER magnets and a high-aspect-ratio rectangular CICC design are evaluated for FNSF magnets, but low-activation-jacket materials may need to be selected. The conductor design concept and TF coil winding pack composition and dimension based on the horizontal maintenance schemes are discussed. Neutron radiation limits for the LTS and HTS superconductors and electrical insulation materials are also reviewed based on the available materials previously tested. The material radiation limits for FNSF magnets are defined as part of the conceptual design studies for FNSF magnets.

Manufacture and test of prototype water pipe chase barrier in ITER Magnet Feeder system

The Magnet Feeder system in the International Thermonuclear Experimental Reactor (ITER) deploys electrical currents and supercritical helium to the superconducting magnets and the magnet diagnostic signals to the operators. In the current design, the feeders located in the upper L3 level of the Tokamak gallery penetrate the Tokamak coolant water system vault, the biological shield and the cryostat. As a secondary confinement to contain the activated coolant water in the vault in the case of water pipe burst accident, a water barrier is welded between the penetration in the water pipe chase outer wall and the mid-plane of the vacuum jacket of the Feeder Coil Terminal Box (CTB). A thin-wall stainless steel diaphragm with an omega shape profile is welded around the CTB as the water barrier to endure 2bar hydraulic pressure. In addition, the barrier is designed as a flexible compensator to withstand a maximum of 15mm of axial displacement of the CTB in case of helium leak accident without failure. This paper presents the detail configuration, the manufacturing and assembly processes of the water barrier. Test results of the prototype water barrier under simulated accident conditions are also reported. Successful qualification of the design and manufacturing process of the water barrier lays a good foundation for the series production of this subsystem.

Manufacture and testing of the CTB&SBB thermal shield for the ITER magnet feeder system

The system of International Thermonuclear Experimental Reactor (ITER) feeders is responsible for the power, helium cooling, and instrumentation of the magnets of the coil terminal box and S-bend box (CTB&SBB) thermal shield outside the cryostat.An 80-K rectangular Al thermal shield is hung inside the CTB&SBB to reduce the thermal radiation heat loads of 4.5-K helium. The American Society of Interventional Pain Physicians (ASIPP) will supply all the 31 sets of feeders for ITER. A manufactured prototype of CTB&SBB thermal shield is first quality-tested before the commencement of the series production.First, a detailed configuration of the rectangular Al thermal shield is presented in this article. The paper also presents more information on the manufacturing process of the thermal shield, especially the welding process, the procedure for ensuring good weld quality, and the use of a specially designed tool to ensure <5-mm deformation on such a 7.3-m-long thermal shield during welding. In addition, the cold test and results, including the cooling process with 13-bar and 17.5-g/s 80-K He gas, and the temperature distribution on different panels of the thermal shield are presented. The whole process of manufacture and testing lays a good foundation for the series production of the thermal shield.

ITER Magnet Feeder: Design, Manufacturing and Integration**supported by the National Special Support for R&D on Science and Technology for ITER (Ministry of Public Security of the People's Republic of China-MPS) (No. 2008GB102000)

The International Thermonuclear Experimental Reactor (ITER) feeder procurement is now well underway. The feeder design has been improved by the feeder teams at the ITER Organization (IO) and the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) in the last 2 years along with analyses and qualification activities. The feeder design is being progressively finalized. In addition, the preparation of qualification and manufacturing are well scheduled at ASIPP. This paper mainly presents the design, the overview of manufacturing and the status of integration on the ITER magnet feeders.

COMPUTATIONAL METHOD FOR ELASTIC–PLASTIC AND ANISOTROPIC SUPERCONDUCTING CABLE UNDER SIMPLE LOAD

International Thermonuclear Experimental Reactor (ITER) Cable-in-conduit Conductor (CICC) presents a typical hierarchical twisted configuration. An unpredictable stress–strain distribution is generated in each-level subcable under experimental loading tests or realistic operating conditions. The mechanical response relates to the threshold of transport currents on CICC which indicates the performance and stability of ITER magnets. To quantitatively assess mechanical deformation we develop an analytical scheme based on the hierarchical approach in the classical wire rope theory, following a successive top-down algorithm from the highest cable level to the lowest. The determination and evaluation of complex contact occurrence among adjacent strands is a key aspect which is formulated via a friction–contact model. Finite element (FE) method is used to extend the analytical approach that treats only elastic and isotropic material to the anisotropy and plasticity. The results show that the interstrand friction acts as a moderation for the axial deformation, and the interaction among the conductors has a stronger impact on the cable deformation than that between the conductor and jacket. It is suggested that under tensile load, the displacement constraints at the end and the strand anisotropy help to reduce the cable longitudinal deformation. It is also recommended that increasing the subsequent pitch is beneficial to reducing the cable deformation level.

Temperature and current margin of ITER central solenoid conductor designs during a 15 MA plasma scenario

The plasma current of the international thermonuclear experimental reactor (ITER) will be induced by current pulses in the central solenoid (CS). The foreseen lifetime and operating conditions, with currents up to 45 kA at 13 T, pose significant mechanical challenges for the brittle Nb3Sn strands. On top of this, inter-strand coupling loss should remain restricted during pulsed operation to ensure an achievable repetition rate of plasma cycles. In order to address the degradation of Nb3Sn cable-in-conduit conductors (CICCs) with electromagnetic and thermal load cycling, four alternative cable designs have recently been tested in both DC and AC modes. However, the actual operating conditions experienced by the CS conductor within the ITER magnet system in terms of time-varying magnetic field and current cannot be fully reproduced in existing experimental facilities. In this paper, we present the results of the numerical assessment of the electromagnetic performance of the four ITER CS conductor designs during a 15 MA plasma scenario. The analysis was carried out with the code JackPot-ACDC including the voltage–current characteristic of all superconducting strands.

Identification of wire breakage status based on the measurement and inversion of magnetic field aiming at applications to cable in conduit conductors of ITER magnets

The wire breakage identification of superconducting multistrands Cable in Conduit Conductors (CICC) is a key point to guarantee the reliability of the magnet system of the International Thermonuclear Experimental Reactor (ITER). In this paper, a Nondestructive Testing (NDT) method based on the measurement of magnetic fields surrounding CICC and the reconstruction of the current values in the superconducting cables from the measured magnetic field signals is presented. The validity and efficiency of the proposed method are demonstrated through numbers of numerical simulations and preliminary experiments.

Electromagnetic and Structural Analysis of ITER Central Solenoid Insert Coil

The United States ITER Project Office (USIPO) is responsible for fabrication of the Central Solenoid (CS) for International Thermonuclear Experimental Reactor (ITER). The CS Insert (CSI) project should provide verification of the conductor performance in relevant conditions of temperature, field, currents and mechanical strain. The USIPO will build the CSI that will be tested at the Central Solenoid Model Coil (CSMC) Test Facility at JAEA, Naka, Japan. This paper presents three-dimensional mathematical model of CSI based on the design. Simulations were performed in order to verify structural integrity and compliance with the Japanese High Gas Pressure Safety Law. Numerical simulations lead to the design of the CSI that produced a required strain levels during operation on the superconductor cable, while simultaneously satisfy the ITER magnet structural design criteria.

Design of the ITER TF Magnet Feeder Systems

The International Thermonuclear Experimental Reactor (ITER) superconducting magnet system requires current supplies from 10 kA to 68 kA. The components relating the electrical power and cryogens through the warm-cold barrier to the ITER magnets are the so-called feeders. In total there are 31 such feeders for the ITER magnet system, 9 of them necessary for the toroidal field magnet system are called TF feeders. Each feeder consists of three main units: the In-Cryostat Feeder (ICF), the Cryostat Feed-Through (CFT) and the S-Bend Box (SBB)/Coil Terminal Box (CTB) assembly. This paper presents the latest design of the TF feeder components.

Low temperature physical properties of a high Mn austenitic steel JK2LB

High manganese austenitic stainless steel JK2LB is developed by the Japan Atomic Energy Agency for applications as a conduit material for superconducting cable-in-conduit conductors for the magnets of international thermonuclear experimental reactor (ITER). The low temperature physical property data of this material are very important to ITER magnet design. Therefore in this paper, our measurements of the physical properties including room temperature Young’s modulus and thermal expansion, magnetization, thermal conductivity, specific heat and resistivity at temperatures from room temperature down to 2 K are reported. We found that JK2LB is antiferromagnetic at low temperatures with a Néel temperature of 240 K. This is consistent with a prediction based on the chemical composition of the austenite stainless steel. The antiferromagnetic phase transition is also evident in the resistivity vs. T curve. Nevertheless, no anomalies are observable in its specific heat and thermal conductivity from 2 K to 300 K. The thermal expansion of this steel between 10 K and 300 K is about 0.22%. Its Young’s modulus, specific heat and thermal conductivity are comparable to that of 316LN stainless steel.

Experimental Studies of Transverse Stress Effects on the Critical Current of a Sub-Sized ${\rm Nb}_{3}{\rm Sn}$ Superconducting Cable

Large superconducting magnets will play a central role in the success of the International Thermonuclear Experimental Reactor (ITER) and for the future of fusion energy. Cable-in-conduit conductors (CICC) will be used for the ITER magnets. As a CICC is energized, electromagnetic forces accumulate across the conductor, pressing strands transversely against one side of the conduit. We have developed a device to study the effect of transverse stress on a sub-sized cable using a mechanical load that simulates the Lorentz loads in the ITER Central Solenoid conductor. The test sample is a single turn (about 110 mm diameter) circular cable composed of 36 superconducting strands (cabling pattern of 3 x 3 x 4 ). The transverse stress is applied to the cable using a conical wedge that converts a vertical force into a radial (transverse) force. The vertical force is provided by a linear actuator. The stress on the cable is measured using two independent techniques: (1) strain gages applied directly to the ring where the cable is located and (2) a load cell located outside the cryostat. The device was successfully tested using a 20T, 190 mm bore Bitter magnet at National High Magnetic Field Laboratory (NHMFL), FSU, and a first series of tests were carried out showing a degradation of the critical current as a function of transverse stress. The maximum transverse stress was 100 MPa, for which the cable critical current degraded to roughly 30% of its initial value.

Conceptual design of a High Temperature Superconductor current feeder system for ITER

The International Thermonuclear Experimental Reactor (ITER) project envisages a techno-economically feasible solution of its current feeder system in order to reduce the overall cryogenic requirements and operational costs. Since the ITER magnet system has a long stand-by time with respect to its operation duty cycle, it is essential to optimize the operational costs of the current feeder system taking into consideration both, the full current and stand-by modes. The present HTS technology has reached the maturity that HTS conductors are applicable for the current feeder system of ITER. The replacement of the actually planned conventional current leads by HTS current leads would provide considerable savings in the refrigeration investment and operational costs. Another option is the substitution of the water cooled high current aluminum feeders by HTS feeders, so called HTS bus bars. In this paper, the different design options of Bi-2223/Ag HTS based bus bars as prototype unit modules for ITER are discussed. The performance of different cooling schemes for HTS bus bars is studied and the design related critical issues e.g. metallic transition (65 K −300 K) and bending of bus bar, AC loss, thermal loss and reliability of the cooling system are investigated.

Development of fine filament NbTi superconducting strands for magnet systems of fusion reactors

Bochvar Institute (VNIINM) has developed and produced fine filament NbTi superconducting strands of several layouts for International Thermonuclear Experimental Reactor (ITER) poloidal field (PF) coils. The strands, having outer diameter of 0.7 mm and filament diameter of 6.5 μm, have been manufactured both by single and double stacking. Strand layouts distinguish by Cu/Sc ratio and a filament zone arrangement in cross-sections, one of them includes CuNi resistive barriers to reduce AC losses in operating condition of ITER magnet system. The results of comparative investigations on critical currents, critical current density, AC losses, critical temperature of the strands are presented. Two strands of these layouts are supposed to be used in the experimental cable for the model coil of SULTAN facility.

Status of ITER magnet design and model coils

The International Thermonuclear Experimental Reactor (ITER) Final Design Report (FDR) of July 1998 provided a comprehensive design of the ITER facility. Although this design fulfilled all its initial objectives, including cost, the financial constraints among the ITER Parties led to the decision that a cost reduction of about 50% is required and should be achieved by a reduction in machine size anti more limited performance objectives. The new machine design under consideration is not only smaller in size, but also puts more emphasis on plasma configurations which are more elongated and have higher triangularity. These new requirements have led to some design changes. One major change is the structural arrangement of the magnets, where the toroidal field (TF) coils are no longer bucked on the central solenoid (CS), but wedged on themselves. The ITER magnet R&D program is focused on the manufacture and testing of two model coils. The CS model coil (CSMC) manufacture has been completed and installation for testing is well underway. The TF model coil (TFMC) manufacture is expected to be completed in early 2000. The main results of the model coil tests will be available by the end of the year 2000.

Integrated design of the ITER magnets and their auxiliary systems

The magnet system design for the International Thermonuclear Experimental Reactor (ITER) has reached a high degree of integration to meet performance and operation requirements, including reliability and maintainability, in a cost effective manner. The article identifies the requirements of long inductive burn time, large number of tokamak pulses, operational flexibility for the poloidal field system, magnet reliability and the cost constraints as the main design drivers. Key features of the magnet system which stem from these design drivers are described, together with interfaces and integration aspects of certain auxiliary systems.

Development of the superconductors for ITER magnet system

A review is given of the present status of the development and production of Nb 3Sn and Nb–Ti superconductors for the Model Coils and the real Magnet System of the International Thermonuclear Experimental Reactor (ITER) in the Russian Federation Home Team. It is shown that Nb 3Sn bronze processed superconductors produced for the Model Central Solenoid Coil insert meet the ITER joint Central Team requirements. In particular, the critical current density, measued in non-Cu area is not less than 550 A/mm 2 for 12 T at 4.2 K, the level of hysteresis losses is not in excess of 200 mJ/cm 3, and the Cu-stabilizing shell resistivity ratio of Cr-plated wire is 150. Internal tin Nb 3Sn superconductor development and test results are presented, confirming the possibility of their application for the ITER Magnet System winding. Nb–Ti superconductors for PF coils properties have also been considered. The possibility of Nb 3Sn and Nb–Ti superconductor manufacture with the use of large composite billets up to 300 mm in dia is shown, creating the possibility for large scale industrial production (several tens of tons/year) of these materials for the ITER Magnet System.

Protection measures for selected ITER magnet system off-normal conditions

The International Thermonuclear Experimental Reactor (ITER) magnet systems provide the magnetic field intensity and field geometry to contain and control plasma during the various phases of pulsed operation. During these pulses, the toroidal field (TF) coils operate with a constant current. The central solenoid (CS) and poloidal field (PF) coils, on the other hand, are each independently powered. The maximum terminal voltages during plasma operation and protective discharges are 15 kV for CS and 10 kV for TF and PF. The energy stored in the 20 TF coil system is 103 GJ; in each of the other coils it is approximately 10 GJ or less. This paper describes the protection requirements and selected design concepts being considered for the large superconducting coils for the ITER. Ground faults, short circuits and helium leaks are the major serious accidents to be prevented in the coils. All coils use a solid insulation system to avoid ground faults. The electrical circuits including coil and power supply are grounded through resistors that limit current in the event of a ground fault. In the case of a short circuit within the coil winding, a large energy would be dissipated close to the small shorted volume. The impact of the short circuit can be reduced by using a potential screen. Inside the cryostat, helium leakage is most likely at the electrical insulating breaks in the cryogenic cooling lines between the coils and helium manifolds. A double containment (metallic shield and glass-epoxy) is therefore provided for the insulation breaks to allow for the detection of small leaks and to limit the spread of helium to other locations.

A Numerical Model for the Simulation of Quench in the ITER Magnets

A computational model describing the initiation and evolution of normal zones in the cable-in-conduit superconductors designed for the international thermonuclear experimental reactor (ITER) is presented. Because of the particular geometry of the ITER cables, the model treats separately the helium momenta in the two cooling channels and the temperatures of the cable constituents. The numerical implementation of the model is discussed in conjunction with the selection of a well-suited solution algorithm. In particular, the solution procedure chosen is based on an implicit upwind finite element technique with adaptive time step and mesh size adjustment possibilities. The time step and mesh adaption procedures are described. Examples of application of the model are also reported.