Equilibrium Field Coils Research Articles

A Global Structural and Electromagnetic Finite Element Model for the Prediction of the Mechanical Behavior of the JT-60SA Superconducting Magnet System

The JT-60SA is a fusion experiment designed to contribute to the early realization of fusion energy by providing support to the operation of ITER; by addressing key physics issues for ITER and DEMO; and by investigating how best to optimize the operation of the next fusion devices that will be built after ITER. It is a combined project of the JA-EU Satellite Tokamak Program under the Broader Approach (BA) Program and JAEA's Program for National Use, and it is to be built in Naka, Japan, using the infrastructure of the existing JT-60U experiment. The superconducting magnet system of JT-60SA consists of a Central Solenoid, six Equilibrium Field coils and eighteen Toroidal Field coils. The systems are connected to each other by means of flexible and kinematic mechanical attachments, with the Toroidal Field magnet acting as the structural backbone of the whole magnet system. This is then supported to the cryostat base of the machine through a series of gravity supports. A detailed finite element model, representing a 40 degree sector of the superconducting magnet system of JT-60SA, has been developed with particular focus on the mechanical connections between the different coil systems. A complete set of analyses were carried out to obtain the electromagnetic force distribution on the three magnet sub-systems during all operational scenarios and consequently to predict the corresponding stresses and deformations. By doing this the integrity of the system and the performance of its bolted and pinned connections were verified against the applicable codes and standards. This paper illustrates the details of the modeling strategy which lead to the production of the finite element model and provides a comprehensive report and a critical analysis of the most relevant results obtained to date.

The Manufacturing of the Superconducting Magnet System for the JT-60SA

JT-60SA is the satellite tokamak for ITER in the Broader Approach agreement. The JT-60SA uses 18 toroidal field coils, a central solenoid with 4 modules, and 6 equilibrium field coils, they are all superconducting coils with forced flow cooled conductors. All detailed designs of these superconducting coils have been completed. The manufacturing of conductors and coils are progressing in Japan and EU. This paper shows the latest manufacturing activities and final design adjusting of its magnet system and their utilities.

Manufacture of the Winding Pack and Development of Key Parts for the JT-60SA Poloidal Field Coils

Manufacture of poloidal field (PF) coil system in JT-60SA is progressing. Recently, fabrication of the winding machines for equilibrium field (EF) coils and central solenoid (CS) were completed and the winding with the superconducting conductor was started for an EF coil (EF4). A few double pancake (DP) coils for EF4 were fabricated, and it was realized that the error of circularity for DP coils became less than the designed value. Design of the conductor end structure was also progressed. This part had the role of restraining the conductor end to prevent its separation from the winding pack. It was confirmed by the structural analysis that the conductor end structure for CS, which receives the large electromagnetic (EM) force, had sufficient mechanical strength. It was also checked by the cold test that this structure satisfied the designed performance for the electrical isolation in addition to the mechanical performance. Regarding the design of the inlet for CS, mechanical reinforcement was considered for its structure. Final design of CS inlet was determined by structural analysis to confirm the sufficient mechanical strength against the vertical compression induced by EM forces.

Quench detection of fast plasma events for the JT-60SA central solenoid

The JT-60 is planned to be modified to a full-superconducting tokamak referred to as the JT-60 Super Advanced (JT-60SA). The maximum temperature of the magnet during its quench might reach the temperature of higher than several hundreds Kelvin that will damage the superconducting magnet itself. The high precision quench detection system, therefore, is one of the key technologies in the superconducting magnet protection system. The pick-up coil method, which is using voltage taps to detect the normal voltage, is used for the quench detection of the JT-60SA superconducting magnet system. The disk-shaped pick-up coils are inserted in the central solenoid (CS) module to compensate the inductive voltage. In the previous study, the quench detection system requires a large number of pick-up coils. The reliability of quench detection system would be higher by simplifying the detection system such as reducing the number of pick-up coils. Simplifying the quench detection system is also important to reduce the total cost of the protection system. Hence the design method is improved by increasing optimizing parameters. The improved design method can reduce the number of pick-up coils without reducing the sensitivity of detection; consequently the protection system can be designed with higher reliability and lower cost. The applicability of the disk-shaped pick-up coil for quench detection system is evaluated by the two dimensional analysis. In the previous study, however, the analysis model only took into account the CS, EF (equilibrium field) coils and plasma. Therefore, applicability of the disk-shaped pick-up coil for the quench detection system remains open question because the fast plasma events, such as disruption, mini collapse and ELM (edge localized mode), directly influences on the voltage of pick-up coil making the quench signal undetectable. Consequently, a new analysis model proposed in the present paper was designed to avoid this difficulty by introducing the passive coil series such as vacuum vessel and stabilizer. The influence of fast plasma events is absorbed by passive coil series like real system, and the evaluation of applicability can be examined in detail. The analysis results show that the disk-shaped pick-up coil is applicable whenever the standard operation, disruption, mini collapse and ELM.

Stability Margin of NbTi CIC Conductor for JT-60SA Equilibrium Field Coil

The equilibrium field (EF) coil conductors are designed with the NbTi cable in conduit (CIC) conductor because the maximum magnetic field of each conductor is up to 4.8 T for the EF-L conductor and 6.2 T for the EF-H conductor. The first EF conductor was manufactured and processed into the performance verification test sample. Then the performance verification tests had been operated before beginning the mass production. The current sharing temperature (Tcs) measurement tests and the quench tests were operated as the performance verification test to evaluate the Tcs and the minimum quench energy (MQE). The results of the Tcs measurement test indicate that both the Tcs of the EF-H conductor and the EF-L conductor are almost same as the expected value estimated by the strand characteristics. Additionally, the Tcs results satisfied the requirements of the EF coils, therefore the conductor design and the fabrication process are going well. The results of the quench test indicate that the EF conductors have enough stability marginuch as more than several hundred mJ/cc at the temperature margin of more than 0.5 K and the current of 20 kA. In addition, the supercritical helium (SHe) mass flow seldom influenced on the MQE. Thus the EF coil can be operated in stable condition even if the SHe mass flow will be decreased.

A global structural and electromagnetic Fe model for the prediction of the mechanical behavior of the JT-60SA superconducting magnet system

The JT-60SA is a fusion experiment designed to contribute to the early realization of fusion energy by addressing key physics issues for ITER and by investigating how best to optimize the operation of the fusion power plants. The superconducting magnet system of JT-60SA consists of three sub-systems: a Central Solenoid split in four segments, a set of six Equilibrium Field coils and a set of eighteen Toroidal Field coils. Such systems are connected each other by means of a series of flexible and kinematic mechanical attachments, with the Toroidal Field magnet acting as the backbone of the whole magnet system. A detailed finite element model, representing a 40 degrees sector of the superconducting magnet system of JT-60SA, has been developed and a complete set of analyses were carried out to obtain the distribution of magnetic forces acting on the three magnet sub-systems, to predict the corresponding deformations, to estimate the stress levels in the magnets’ mechanical structures and to verify the integrity of the system and its fatigue life, in agreement with the applicable Codes and Standards and in consideration of the expected lifespan of the machine. This paper illustrates the details of the modeling strategy which lead to the production of the finite element model and provides a comprehensive report and a critical analysis of the most relevant results obtained in the calculations carried out so far.

Fabrication and tests of EF conductors for JT-60SA

The conductors for plasma equilibrium field (EF) coils of JT-60SA are NbTi cable-in-conduit (CIC) conductor with stainless steel 316L jacket. The production of superconductors for actual EF coils started from February 2010. Nine superconductors with 444 m in length were produced up to July 2010. More than 300 welding of jackets were performed. Six nonconformities were found by inspections as go gauge, visual inspection and X-ray test. In order to shorten the manufacturing time schedule, helium leak test was conducted at once after connecting the long length jacket not just after the welding. The maximum force to pull the cable into jacket was about 7.6 kN on average. The mass flow rates of 9 conductors showed almost same values indicating that there are no blockages in the conductors. The measured current sharing temperature agreed with the expectation values from strand performance indicating that no degradation was caused by production process. The coupling time constants of conductors ranged from 80 to 90 ms which are much smaller than the design value of 200 ms.

Development of effective power supply using electric double layer capacitor for static magnetic field coils in fusion plasma experiments

A cost-effective power supply for static magnetic field coils used in fusion plasma experiments has been developed by application of an electric double layer capacitor (EDLC). A prototype EDLC power supply system was constructed in the form of a series LCR circuit. Coil current of 100 A with flat-top longer than 1 s was successfully supplied to an equilibrium field coil of a fusion plasma experimental apparatus by a single EDLC module with capacitance of 30 F. The present EDLC power supply has revealed sufficient performance for plasma confinement experiments whose discharge duration times are an order of several seconds.

ＪＴ‐６０ＳＡ真空容器の設計と試作試験

JT-60 is planned to be upgraded to a JT-60SA superconducting tokamak machine. This project is the JA-EU satellite tokamak program under both broader approach and Japanese domestic programs. The JT-60SA tokamak is composed of the following main components: vacuum vessel (VV), thermal shield, superconducting coils (toroidal field coil, equilibrium field coil, and central solenoid), cryostat, and heating facilities. The VV has a D-shaped poloidal cross section and a double-wall structure to ensure high rigidity and toroidal one-turn resistance simultaneously. The material of the VV is 316L stainless steel with a low cobalt content of <0.05 wt%. Before the start of manufacturing the VV, fundamental welding R&D was performed to study the manufacturing procedures. The manufacturing procedures were successfully established with a trial manufacturing of 20 degree upper half of the VV. Based on the results, the actual VV manufacturing has started in November 2009.

Recent Progress of the Design Activity for the Poloidal Field Coil System in JT-60SA

PA (procurement arrangement) for poloidal field (PF) coil system, which consists of the central solenoid (CS) and the equilibrium field (EF) coils, was agreed between Japan and EU. During this activity, design of PF coils system was continued to be modified. For CS, material for the jacket of this conductor was changed into stainless steel (316LN) to make providing easier. In the modified material, maximum stress at the jacket was kept within the allowable limit. Accompanying this modification, the amount of pre-compress had to be re-estimated. Therefore, it was clarified that designs of pre-compression and tie plates need not to be major modification. For EF coils, positions and the number of turns were modified since the progress of the research for the plasma operations required in JT-60SA. Due to this optimization, total amount of superconducting material was reduced. The detail designs of PF coils were also performed to reduce the materials of supports and to evaluate the mechanical strength considering the various events. Thickness of clamp plate of the EF coil which received relatively small electromagnetic force was able to be reduced. Regarding the design of support legs with flexible plate, deformation of toroidal field (TF) coil was considered that should be included the evaluation of stress at this parts because this parts are directly attached on the TF coil case. Therefore, the revised designs of supports with sufficient mechanical strength were obtained for EF1 and EF4.

Construction of the Jacketing Facility and First Production Results of Superconductor for JT-60SA

In order to fabricate the central solenoid (CS) and equilibrium field (EF) coil conductors for JT-60 Super Advanced (JT-60SA), the jacketing facility of 680 m was constructed at Naka site of Japan Atomic Energy Agency. The dummy copper conductors of 10 m and superconductors of 7 m were fabricated. The attachment sleeve to connect the cable and winch wire to pull cable through a 551 m jacket in maximum was developed. The sleeve attained the 30 kN of pulling force. In order to investigate the strand damage by the edge of central spiral through compaction and bending process, the measurements of strand critical current in prototype EF conductor with bending curvature radius of 150 mm was conducted. The degradation of was not observed.

Stability and Quench Test for NbTi CIC Conductor of JT-60SA Equilibrium Field Coil

The EF coil conductors of JT-60SA are designed with the NbTi cable in conduit conductor because the maximum magnetic field goes up to 6.2 T. The prototype NbTi conductor was developed and tested to confirm the capability of the real conductor. The prototype conductor was proven to have enough Tcs margin under the operating conditions investigated in previous test. In this study, the quench test was operated to measure the stability margin and the normal state propagation. Firstly, the MQE under the Tcs margin of 0.2 K was found to be 80 mJ/cc-strand which is almost the same as that of the previous CIC conductor to result in enough stability margins on EF coil operation. Secondly, the propagation velocity to upstream direction and downstream direction were found to be about 0.47 m/s and 0.54 m/s, respectively. Finally, the quench analysis was conducted to calculate the maximum temperature during quench. The analysis results showed that the maximum temperature reached about 70 K, which is within the permissible value of 150 K for the EF coils.

Development of JT-60SA superconducting magnet system

The upgrade of JT-60U magnet system to superconducting coils (JT-60SA) is progressing by both parties of Japanese government and European commission in the framework of the Broader Approach agreement. The magnet system for JT-60SA consists of 18 Toroidal Field (TF) coils, a Central Solenoid (CS) with four modules, six Equilibrium Field (EF) coils. The TF coil case encloses the winding pack and is the main structural component of the magnet system. The CS consists of four independent winding pack modules, which is support from the bottom of the TF coils. The six EF coils are attached to the TF coil cases through supports with flexible plates allowing radial displacements. The construction of CS and EF coils was started in 2008 by Japan Atomic Energy Agency (JAEA). The construction of TF coils will be started in 2009 by Fusion for Energy and European voluntary contributors. This paper introduces the design of the JT-60SA superconducting magnet system and activities for development of magnet components in Japan.

Upgrading the NIFS superconductor test facility for JT-60SA cable-in-conduit conductors

The superconductor test facility at the National Institute for Fusion Science (NIFS) was upgraded to test cable-in-conduit (CIC) conductors for the JT-60SA equilibrium field (EF) coil. Supercritical helium (SHe) lines were assembled with transfer tubes and a heat exchanger. The CIC conductor was covered with a thermal insulation vessel, filled with gas helium at atmospheric pressure. The temperature of the conductor was varied using a film heater attached to an inlet pipe. Critical current (Ic) and current sharing temperature (Tcs) measurements of the prototype CIC conductor were carried out successfully in the upgraded test facility. During the measurements, the conductor temperature was precisely controlled.

JT-60SA用超伝導マグネット

The upgrading of the JT-60U to the JT-60SA has started in a joint effort by the Japanese government (JA) and European commission (EU) under the framework of the Broader Approach (BA) Agreement. The superconducting magnet system for the JT-60SA consists of 18 toroidal field (TF) coils, a central solenoid (CS) with four modules, and six equilibrium field (EF) coils. The TF case encloses the winding pack and is the main structural component of the magnet system. The CS consists of independent winding pack modules, which are hung from the top of the TF coils through its pre-load structure. The six EF coils are attached to the TF coil case through supports with flexible plates that allow radial displacement. The CS modules operate at high field and use a Nb3Sn superconductor. The TF coils and EF coils use NbTi superconductors. This paper describes the technical requirements, operational interface and detailed manufacturing design outline of the superconducting magnet system for the JT-60SA.

Conceptual Design of Superconducting Magnet System for JT-60SA

The upgrade of JT-60U magnet system to superconducting coils (JT-60SA) has been decided by both parties of Japanese government (JA) and European commission (EU) in the framework of the Broader Approach (BA) agreement. The magnet system for JT-60SA consists of 18 toroidal field (TF) coils, a Central Solenoid (CS) with four modules, seven Equilibrium Field (EF) coils. The TF case encloses the winding pack and is the main structural component of the magnet system. The CS consists of independent winding pack modules, which is hung from the top of the TF coils through its pre-load structure. The seven EF coils are attached to the TF coil cases through supports which include flexible plates allowing radial displacements. The CS modules operate at high field and use Nb3 Sn type superconductor. The TF coils and EF coils use NbTi superconductor. The magnet system has a large heat load from nuclear heating from DD fusion and large AC loss. This paper describes the technical requirements, the operational interface and the outline of conceptual design of the superconducting magnet system for JT-60SA.

Conductor Design of CS and EF Coils for JT-60SA

The conductor for central solenoid (CS) and equilibrium field (EF) coils of JT-60 Super Advanced (JT-60SA) were designed. The conductor for CS is Nb3Sn Cable-In-Conduit (CIC) conductor with JK2LB jacket. EF coil conductors are NbTi CIC conductor with SS316LN jacket. The field change rate (3.9 T/s), faster than ITER generates the large AC loss in conductor. The analyses of current sharing temperature (Tcs)margins for these coils were performed by the one-dimensional fluid analysis code with transient heat loads. The margins of these coils are 1 K for the plasma standard and disruption scenarios. The minimum Tcs margin of CS conductor is 1.2 K at plasma break down (BD). The margin is increased by decreasing the rate of initial magnetization. It is found that the disruption mainly impacts the outer low field EF coil. The disruption decreases the Tcs margin of the coil by >1 K. A coupling time constant of <100 ms, Ni plating, and a central spiral are required for NbTi conductor.

Conceptual Design of Magnet System for JT-60 Super Advanced (JT-60SA)

At Japan Atomic Energy Agency (JAEA), the JT-60 is planned to be modified to a full-superconducting tokamak referred as JT-60 Super Advanced (JT-60SA) as one of the JA-EU broader approach projects. In JT-60SA, the magnet system with 1400 ton consists of 18 toroidal field (TF) coils, 4 stacks of central solenoid (CS) and 7 plasma equilibrium field (EF) coils. The TF coil with using the NbTi cable-in-conduit (CIC) conductor has the maximum magnetic field of 6.4 T in winding, the magnetomotive force of 8.2 Tm and the magnetic stored energy of 1.5 GJ which is 1.5 times larger than present superconducting fusion devices. The conductor for CS which is the Nb3Sn CIC conductor with the maximum magnetic field of 10 T was newly designed to increase the flux swing capability for the ITER plasma simulation and to accommodate the diminution of available space because of the TF coil with NbTi conductor. For CS, the faster maximum varying field of 2.88 T/s is required because of fast plasma initiation scenario of JT-60SA.

Poloidal Field Coil Configuration and Plasma Shaping Capability in NCT

This paper describes the latest design of poloidal field coil configuration in the national centralized tokamak (NCT), that is a new design based on the former superconducting coil tokamak JT-60SC. The most notable design change was made for the outer equilibrium field (EF) coils to increase a plasma shaping parameter <tex xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">$ S(= q_95ast I_ p/ aast B_ t)$</tex> and to make off-axis heating using N-NBI possible. As a result, the maximum plasma shaping parameter of <tex xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">$sim$</tex> 7 was obtained by double null divertor configuration with a lower aspect ratio of A <tex xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">$sim$</tex> 2.6. It was confirmed that two additional EF coils are useful for the plasma squareness control, but only one is finally adopted for ITER plasma simulation and to keep compatibility of off-axis N-NBI heating.

Coupling Effect between Equilibrium Field and Heating Field and Modification of the Power Supply System on SUNIST Spherical Tokamak

Strong inductive coupling between the heating field and equilibrium field is confirmed to be responsible for the poor plasma equilibrium in initial discharges on the SUNIST spherical tokamak. A modification project for the power supply system of equilibrium field coils is successfully performed to increase the duration time of plasma current flattop from much less than 1ms to about 2 ms.