EF Coils Research Articles

Completion of JT-60SA construction and contribution to ITER

Construction of the JT-60SA tokamak was completed on schedule in March 2020. Manufacture and assembly of all the main tokamak components satisfied technical requirements, including dimensional accuracy and functional performances. Development of the plasma heating systems and diagnostics have also progressed, including the demonstration of the favourable electron cyclotron range of frequency (ECRF) transmission at multiple frequencies and the achievement of long sustainment of a high-energy intense negative ion beam. Development of all the tokamak operation control systems has been completed, together with an improved plasma equilibrium control scheme suitable for superconducting tokamaks including ITER. For preparation of the tokamak operation, plasma discharge scenarios have been established using this advanced equilibrium controller. Individual commissioning of the cryogenic system and the power supply system confirmed that these systems satisfy design requirements including operational schemes contributing directly to ITER, such as active control of heat load fluctuation of the cryoplant, which is essential for dynamic operation in superconducting tokamaks. The integrated commissioning (IC) is started by vacuum pumping of the vacuum vessel and cryostat, and then moved to cool-down of the tokamak and coil excitation tests. Transition to the super-conducting state was confirmed for all the TF, EF and CS coils. The TF coil current successfully reached 25.7 kA, which is the nominal operating current of the TF coil. For this nominal toroidal field of 2.25 T, ECRF was applied and an ECRF plasma was created. The IC was, however, suspended by an incident of over current of one of the superconducting equilibrium field coil and He leakage caused by insufficient voltage holding capability at a terminal joint of the coil. The unique importance of JT-60SA for H-mode and high-β steady-state plasma research has been confirmed using advanced integrated modellings. These experiences of assembly, IC and plasma operation of JT-60SA contribute to ITER risk mitigation and efficient implementation of ITER operation.

Progress of Manufacturing CS and EF Coils for the JT-60SA Tokamak

Progress of Manufacturing CS and EF Coils for the JT-60SA Tokamak

AC Loss of $\hbox{Nb}_{3}\hbox{Sn}$ Butt Joint for JT-60SA Central Solenoid

The upgrading project of the JT-60 to the JT-60 Super Advanced (JT-60SA) at Japan Atomic Energy Agency has started as a joint effort between Japan and the EU. The coils of the JT60-SA are separately fabricated in Japan and EU. The CS and EF coils are fabricated in Japan. In this paper, we describe the butt joint of the JT-60SA CS and measured the ac loss under a time-varying external magnetic field, and analyzed the ac loss at the butt joint by using the FEM simulation software COMSOL Multiphysics. As a result, the measured ac loss of the butt joint were found to be the smallest of all joints (the CS butt joint, the EF pancake joint, the terminal EF joint and the EF prototype joint), and to strongly depend on the thickness of the copper sleeve in full-size coil joint.

Joint resistance measurements of pancake and terminal joints for JT-60SA EF coils

To evaluate the joint fabrication technology for the JT-60SA EF coils, joint resistance measurements were conducted using a sample consisting of pancake and terminal joints. Both joints are shake-hands lap joints composed of cable-in-conduit conductors and a pure copper saddle-shaped spacer. The measurements demonstrated that both joints fulfilled the design requirement. Considering these measurements, the characteristics of both joints were investigated using analytical models that represent the joints. The analyses indicated that the characteristics of the conductors used in the joints affect the characteristics of the joints.

Feeder components and instrumentation for the JT-60SA magnet system

The modifying of the JT-60U magnet system to the superconducting coils is progressing as a satellite facility for ITER by both parties of Japanese government and European commission in the Broader Approach agreement. The magnet system requires current supplies of 25.7kA for 18 TF coils and of 20kA for 4 CS modules and 6 EF coils. The magnet system generates an average heat load of 3.2kW at 4K to the cryogenic system. The feeder components connected to the power supply provide current supply. The cooling pipes connected to the cryogenic system provide coolant supply. The instrumentation of the JT-60SA magnet system is used for its operation.

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.

AC Loss and Temperature Rise at CIC Conductor Joint in EF Coil of JT-60SA

Japan and EU are promoting the project that the coils in JT-60 installed in Japan Atomic Energy Agency are changed to superconducting coils. We numerically and experimentally evaluate thermal stability of a conductor joint in the planning superconducting coil (EF coil). We measured an AC loss of the prototype joint, and estimated the coupling time constant. The AC loss at the EF coils joint segment was calculated based on the coupling time constant and the magnetic field distribution. Moreover, the temperature rise due to the AC loss was numerically derived. The calculation result showed that the temperature rise was small and the coil was enough stable.

Mechanical Design of JT-60SA Magnet System

Latest design of superconducting magnet system in JT-60SA is presented. This magnet system consists of the TF coils and PF coils (CS and EF coils). For this magnet system, stress analyses are systematically carried out to optimize the structural design. In the analysis of TF coil, it is clarified that jacket of conductor has significant strength under the designed electromagnetic (EM) load. In addition, coil case has margin of mechanical strength. For the PF coils, maximum stress appears at jackets of CS and EF conductor are less than 500 MPa, which is within fatigue limit. The latest design of CS support structure is very strong, because the no gap is made between CS modules even though total repulsive force of CS is maximum with no pre-compression. Therefore, optimization or simplification of structure design is possible for all magnet systems from the latest design in the detail design phase.

STARFIRE, a commercial tokamak power plant design

STARFIRE is a design for a conceptual commercial tokamak electrical power plant based on the deuterium/tritium/lithium fuel cycle. In addition to the goal of being technologically credible, the design incorporates safety and environmental considerations. STARFIRE is considered to be the tenth in a series of commercial fusion power plants. STARFIRE has a 7-m major radius reactor producing 1200 MW of net electrical power from 4000 MW of thermal power, with an average neutron wall load of 3.6 MW/m 2. The aspect ratio is 3.6 and a D-shaped plasma with a height-to-width ratio of 1.6 and average toroidal beta of 0.067 is used. The maximum magnetic field is 11T. Availability goals have been set at 85% for the reactor and 75% for the complete plant including the reactor. The major features for STARFIRE include a steady-state operating mode based on a continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum for impurity control, most superconducting EF coils outside the TF superconducting coils, fully remote maintenance, and a low-activation shield.