Highly Efficient Liquid Hydrogen Storage System by Magnetic Levitation Using HTS Coils

Highly efficient liquid hydrogen storage system is studied with magnetic levitation using high-temperature superconducting (HTS) coils. The system also has high safety in case of emergency, such as an earthquake, with a seismic isolation to absorb vibrations provided by HTS levitation coils setup on the ground side. In such an emergency case, it is considered that a large amount of ac losses are generated in HTS coils, and the winding temperature may rise to lead to a coil quench. In this study, the self-oscillation-type heat pipe (OHP), whose thermal transport property is much greater than that of solid thermal conduction, is used to cool the coil windings. As a result, an HTS coil equipped with an OHP cooling system can be realized, supporting both low heat loads in the usual operation and high heat loads in an emergency

Measurement of Decay Time Constant of Shielding Current in ITER-TF Joint Samples

Joint sample tests have been carried out as a qualification test for ITER Toroidal Field (TF) coils. The joint sample comprises two short TF conductors that have "twin-box" joint terminals at both ends. The lower joint is a testing part that is a full size joint of the TF coils. Hall probes are attached on the lower joint box at around the center of the external field coil of the test facility. The magnetic field induced by shielding currents in the joint can be estimated from the difference between the measured magnetic field strength and the magnetic field generated by the external field coil. The magnetic field by the shielding currents during shut-off of the external field coil from -1.0 T is evaluated for six samples. The decay time constants of the shielding currents are gradually elongated with decrease of the shielding currents in all the samples. In comparison with simulation results, it is considered that the main shielding current flows in superconducting cables in the two conductors with crossing the jointed plane and that the joint resistance is decreased at low total current

Evaluation of ITER TF Coil Joint Performance

To evaluate the ITER TF joint performance, the joint test sample, which consists of two short TF conductors and has full size joint, shall be tested using NIFS test facility under the condition of current of 68 kA and external field of 2 T. For high accuracy, the issue of voltage difference between cable and jacket had been anticipated in the evaluation of joint resistance. If a voltage difference exist between them, it is difficult to measure real joint resistance using voltage taps on the jacket. Therefore, the author first calculated the position where voltage of cable and jacket become equipotential and then decided the voltage tap position where the influence of voltage drop could be avoided. Thus, a high accuracy measurement of joint resistance could be achieved and the joint resistance was accurately evaluated as around 1 n Ω , which is well below the ITER requirement of 3 n Ω

Test of ITER-TF Joint Samples With NIFS Test Facilities

Qualification tests of the ITER toroidal field (TF) conductor joints have been carried out by testing joint samples with test facilities in the National Institute for Fusion Science, NINS, Toki, Japan. The joint sample consists of two short TF conductors with the length of 1535 mm, which is restricted by the test facility with 9-T split coils and 100-kA current leads. The sample current is supplied from a dc 75-kA power supply. Each conductor has two joint boxes at both terminals. The lower joint is a testing part that is a full-size joint of the TF coil. The joint resistance of the lower joint is estimated from the increase of the average voltage drop among the six taps on the conductor against the currents. Five joint samples were tested until 2016, and all the samples satisfied the requirement of the joint resistance at less than 3 nΩ. The method of the measurement and the results are summarized, and the voltage distribution among the voltage taps is discussed

Development of FAIR conductor and HTS coil for fusion experimental device

This study is aimed at the development of high-temperature superconducting (HTS) magnets for application in a fusion experimental device next to the Large Helical Device (LHD). By applying the features of an HTS, high current density and high stability can be balanced. As a candidate conductor, REBCO tapes and pure aluminum sheets are laminated and placed in the groove of an aluminum alloy jacket with a circular cross-section, after joining a lid to the jacket using friction stir welding, and twisting the conductor to homogenize its electrical and mechanical properties. The FAIR conductor derives its name from the processes and materials used in its development: Friction stir welding, an Aluminum alloy jacket, Indirect cooling, and REBCO tapes. Initially, the degradation of the critical current of the FAIR conductor is observed, which was eventually resolved. The development status of the FAIR conductor has been reported

Commissioning Test Results of Variable Temperature Helium Refrigerator/Liquefier for NIFS Superconducting Magnet Test Facility

The superconducting magnet test facility in the National Institute for Fusion Science has been upgraded for excitation tests at a wide temperature range and a higher magnetic field of 13 T. As part of the upgrade, the helium refrigerator/liquefier that operated for 24 years was replaced with a variable-temperature helium refrigerator/liquefier. The required liquefaction rate is 250 L/h, and the required refrigeration capacity is 600 W at 4.5 K, same as the previous one. In addition, it has a new feature that can supply helium gas of a wide temperature range. The typical design cooling capacity is 1 kW under the condition of 20-K supply/30-K return and 1.5 kW under the condition of 40-K supply/50-K return. After the replacement, a series of commissioning tests were performed under the various operational conditions. From the results, the satisfactory thermodynamic performance was confirmed. In the future, it is expected that the substantial progress will be made in the development of large-scale superconducting magnets with advanced superconductors such as high-temperature superconductors and MgB2. The design of the variable-temperature helium refrigerator/liquefier and the results of the commissioning tests are reported in detail

Results of All ITER TF Full-Size Joint Sample Tests in Japan

Nine toroidal field (TF) coils have been developed in Japan for the international thermonuclear experimental reactor (ITER). The joint resistance of TF coil should satisfy the requirement of smaller than 3 nano-ohm at 2 T of external magnetic field and 68 kA of transport current. Full-size joint sample (FSJS) tests were performed for joint development and for TF coil manufacture, as part of the process control. 11 FSJS tests are conducted in total. FSJS tests were conducted with assistance from a test faculty in the National Institute for Fusion Science as reported in a previous paper. All FSJS tests successfully satisfied the requirement of resistance less than 3 nΩ at 2 T. Additionally, the TF coil joints are subjected to cyclic electromagnetic force and warm-up/cool-down during the ITER operation. The authors investigated the joint performance for the abovementioned influence. The results showed no degradation in the joint resistance. Thus, the TF joint developed in Japan was qualified successfully

Improvement of Ic degradation of HTS Conductor (FAIR Conductor) and FAIR Coil Structure for Fusion Device

As a high-temperature superconducting (HTS) conductor with a large current capacity applicable to a nuclear fusion experimental device, REBCO (REBa 2 CuO y ) tapes and high-purity aluminum sheets are alternately laminated, placed in a groove of an aluminum alloy jacket having a circular cross section, and the lid is friction-stir welded. To make the current distribution and mechanical characteristics uniform, the conductor is twisted at the end of the manufacturing process. In the early prototype conductor, when the I c was measured in liquid nitrogen under self-magnetic field conditions, I c degradations were observed from the beginning, and the characteristic difference between the two prototype samples under the same manufacturing conditions were large. Furthermore, I c degradation was progressed by repeating the thermal cycle from room temperature to liquid nitrogen temperature. This I c degradation did not occur uniformly in the longitudinal direction of the conductor but was caused by local I c degradation occurring at multiple locations. If the conductor was not manufactured uniformly in the longitudinal direction, the difference in thermal shrinkage between the REBCO tape and the aluminum alloy jacket caused local stress concentration in the REBCO tape and buckling occurred. Element experiments to explain this mechanism were conducted to clarify the conditions under which I c degradation due to buckling occurs. Then prototype conductors were tested with improved manufacturing methods, and as a result, I c degradation could be suppressed to 20% or less. We have achieved the prospect of producing a conductor with uniform characteristics in the longitudinal direction