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Kizu, Kaname; Tsuchiya, Katsuhiko; Yoshida, Kiyoshi; Edaya, Masahiro; Ichige, Toshikatsu*; Tamai, Hiroshi; Matsukawa, Makoto; della Corte, A.*; Di Zenobio, A.*; Muzzi, L.*; et al.
IEEE Transactions on Applied Superconductivity, 18(2), p.212 - 215, 2008/06
Times Cited Count：19 Percentile：67.07(Engineering, Electrical & Electronic)The maximum magnetic field and maximum current of CS and EF coils is 9 T, 20 kA and 6.2 T, 21 kA, respectively. The conductor for CS is Nb
Sn CIC conductor with JK2LB conduit. On the other hand, EF coil conductors are NbTi CIC conductor with SS316LN conduit. In order to reduce the pressure drop and to raise the temperature margin against large AC loss and nuclear heating, central spiral is introduced inside cable. The Tcs margin and stability analyses of the CS and EF coils are performed by using the one-dimensional fluid analysis code with transient heat loads. These coils have enough high Tcs and stability margin against the operational scenario.
Takahashi, Yoshikazu; Yoshida, Kiyoshi; Nabara, Yoshihiro; Edaya, Masahiro*; Bessette, D.*; Shatil, N.*; Mitchell, N.*
IEEE Transactions on Applied Superconductivity, 17(2), p.2426 - 2429, 2007/06
Times Cited Count：14 Percentile：58.29(Engineering, Electrical & Electronic)The ITER TF coils consists of 18 D-shape coils. The operating current, the maximum field and the stored magnetic energy are 68 kA, 11.8 T and 41 GJ, respectively. A NbSn cable-in-conduit conductor with a central channel is used, with a cooling length of 380 m. An accurate prediction of the coil performance requires, in addition to assessments of the superconductor behavior, a thermohydraulic analysis of the supercritical He. The overall thermohydraulic conditions were simulated by the full-scale quasi three dimensional code VINCENTA. Analysis of stability and quench was carried out using one dimensional Gandalf electric and thermohydraulic code. An interface was written between these codes. The stability margin against the mechanical disturbance and due to a plasma disruption was estimated. In the quench analysis, the temperature rise during the fast discharge was calculated. According to these results, it is confirmed that the TF coils will be operated with the designed performance.
Takahashi, Yoshikazu; Yoshida, Kiyoshi; Nabara, Yoshihiro; Edaya, Masahiro*; Mitchell, N.*
IEEE Transactions on Applied Superconductivity, 16(2), p.783 - 786, 2006/06
Times Cited Count：9 Percentile：46.66(Engineering, Electrical & Electronic)To investigate the conductor behavior during a quench, quench tests of Center Solenoid (CS) insert coils were carried out with various initial conditions in DC and pulse modes. The conductor has very similar configuration and parameters. The inductive heater, attached at the center of the length, initiated an artificial quench in DC mode. A quench has also occurred during the pulse operation with the ramping rate of 0.4-2 T/s. Simulations of electric, thermal and hydraulic behaviors of the conductor during the quench tests in both modes were carried out by using the thermohydraulic simulation code. The experimental results were compared with the simulation and good agreement was obtained. These results are described and the implication for quench detection in ITER is discussed in this paper. The voltage tap method will be used for the quench detection for the CS, and the sensitivity of the detection and the maximum temperature of the conductor during a quench are described. It is shown that the detection system could be designed with high enough detection sensitivity.
Takahashi, Yoshikazu; Yoshida, Kiyoshi; Nabara, Yoshihiro; Edaya, Masahiro*
no journal, ,
Quench tests of Center Solenoid (CS) insert in both DC and pulse modes have been carried out with various initial conditions in the Center Solenoid Model Coil Facility, in order to investigate the CS conductor behavior during a quench. The conductor of the coil is cable-in-conduit conductors with NbSn strands and has similar design and configuration to the conductor for the ITER Central Solenoid. The length of the conductor of the CS insert is about 140 m. Simulations of electric, thermal and hydraulic behavior of the conductor during quench tests in the pulse mode were carried out by using the "Gandalf" simulation code. The quench currents obtained by the simulation agree to the experimental results within 10%. The implication for quench detection in ITER is also discussed and it is confirmed that the maximum temperature of the ITER-CS conductor during a quench is 132K and lower than the design criteria (
150 K) with enough margin. The ITER-CS will not have serious damages during a quench.
Takahashi, Yoshikazu; Yoshida, Kiyoshi; Nabara, Yoshihiro; Edaya, Masahiro*
no journal, ,
no abstracts in English
Takahashi, Yoshikazu; Yoshida, Kiyoshi; Nabara, Yoshihiro; Edaya, Masahiro*
no journal, ,
The ITER TF coil system consists of 18 D-shape coils. The operating current and the maximum field are 68 kA and 11.8 T, respectively. A NbSn cable-in-conduit conductor is used, with a cooling length of about 380 m. The overall thermohydraulic conditions for the analysis were simulated by the full-scale quasi 3-D code (VINCENTA). Analysis of quench of the TF conductor was carried out by using the 1-D electric and thermohydraulic simulation code (Gandalf). An interface is written between these codes. The temperature and voltage distributions during a quench were estimated. The 2-D analysis code (ANSYS) was used to simulate the heat transfer from the conductor to the radial plates during a quench. The results show that the maximum temperature in the conductor during quench is about 150K, if a quench can be detected with a voltage of 0.2 V. These results conclude that the TF coils will be discharged without any damage, even if an unexpected quench occurs.
Kizu, Kaname; Tsuchiya, Katsuhiko; Yoshida, Kiyoshi; Edaya, Masahiro; Tamai, Hiroshi; Matsukawa, Makoto
no journal, ,
no abstracts in English
Tsuchiya, Katsuhiko; Edaya, Masahiro; Suzuki, Yutaka; Kizu, Kaname; Yoshida, Kiyoshi; Matsukawa, Makoto
no journal, ,
no abstracts in English
Kizu, Kaname; Yoshida, Kiyoshi; Edaya, Masahiro; Tsuchiya, Katsuhiko; Matsukawa, Makoto; Ichige, Toshikatsu*
no journal, ,
AC losses are generated in the CS and EF coils by the change of magnetic field and operational current caused by plasma disruption. It is required that the temperature margin for current sharing temperature has to be larger than 1 K even though the event of disruption. Thus, the heat evolution by disruption was evaluated as follows. At first, time evolution of current in coil, vacuum vessel, baffle plate, control coil and plasma was investigated by circuit analysis. Then, time evolution of magnetic field in each coil was calculated by the current pattern. Finally, AC losses were evaluated by the field pattern. It was found that the temperature increase of innermost turn in EF 1 coil becomes about 0.5 K. This indicates that the 1 K temperature margin is attained taking into account the temperature increase by standard plasma operational scenario.
Tsuchiya, Katsuhiko; Edaya, Masahiro*; Kizu, Kaname; Yoshida, Kiyoshi; Matsukawa, Makoto
no journal, ,
no abstracts in English
