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Sn strandKajitani, Hideki; Ishiyama, Atsushi*; Agatsuma, Ko*; Murakami, Haruyuki; Hemmi, Tsutomu; Koizumi, Norikiyo
Teion Kogaku, 50(12), p.608 - 615, 2015/12
A cable-in-conduit (CIC) conductor using NbSn strand is applied to an ITER TF coil. The NbSn strand in the conductor is periodically bent due to electromagnetic force, which causes degradation of performance. This degradation should be evaluated to predict conductor critical current performance. In a past study, a numerical simulation model was developed to evaluate the superconductivity of a periodically bent single strand. However, this model is not suitable for application to strands in the conductor because of the extensive calculation time. The author thus developed a new analytical model with a much shorter calculation time to evaluate the performance of periodically bent strand. This new model uses the classical model concept of a high transverse resistance model (HTRM). The calculated results show good agreement with the test results of a periodically bent NbSn strand. This indicates that a more practical solution can be achieved when evaluating the performance of periodically bent strands. Thus, the model developed in this study can be applied to evaluate the performance of conductors incorporating many strands.
Saito, Toru; Okubo, Toshikazu*; Izumi, Keisuke*; Okawa, Yoshinao*; Kobayashi, Norihiro*; Yamazaki, Toru; Kawano, Katsumi; Isono, Takaaki
Teion Kogaku, 50(8), p.400 - 408, 2015/08
Aramid fiber-reinforced plastic (AFRP) has been developed as a structural material that has the advantages of light weight and high strength. In this study, tensile tests were carried out to measure the tensile properties of AFRP rod on the market for reinforcement of concrete at room temperature, 77 K and 4.2 K. Especially at cryogenic temperatures, it is difficult to perform a tensile test of the bar because the specimen slips through the jig grip. To prevent the rod from slipping, tensile tests were carried out with some filling conditions. The applicable and appropriate tensile test conditions were established by modifying the jig grip, treating the surface of the rod and using cryogenic epoxy infill to grip the rod. They were more than 1100 MPa. Additionally, the AFRP rod included a temperature dependence in which the Young's modulus increased as the test temperature decreased. It was confirmed that the Young's modulus increased because aramid fiber was more dominant than epoxy.
Nakajima, Hideo; Shimamoto, Susumu*; Iguchi, Masahide; Hamada, Kazuya; Okuno, Kiyoshi; Takahashi, Yoshikazu
Teion Kogaku, 48(10), p.508 - 516, 2013/10
JAEA is procuring both structural materials and structural design of Toroidal Field (TF) coil and Central Solenoid (CS) for ITER. Although 316LN is used in the most parts of the superconducting magnets system, the cryogenic stainless steels, JJ1 and JK2LB, which were newly developed by JAEA and Japanese steel companies, are used in the highest stress area of TF coil case and whole CS conductor jackets, respectively. These two materials became commercially available based on demonstration of productivity and weldability of materials, and evaluations of 4 K mechanical properties of trial products including welded parts. In order to simplify quality control in mass production, JAEA has used materials specified in the material section of "Codes for Fusion Facilities - Rules on Superconducting Magnet Structure (2008)" issued by the Japan Society of Mechanical Engineers (JSME). The design of structural materials, production technology and quality control are described in this paper.
Shimamoto, Susumu*; Nakajima, Hideo; Takahashi, Yoshikazu
Teion Kogaku, 48(2), p.60 - 67, 2013/03
JAEA started development of cryogenic structural material for Tokomak fusion reactor 30 years ago. Because, there was no specialized steel and mechanical data at 4K, JAEA settled target of mechanical characteristics which should satisfy requirements for coil structure at 4K and equipped evaluation facilities at 4K such as tensile test, fatigue test and so on. On the other hand JAEA initiated collaboration with steel industries in order to realize new cryogenic structural material and carried out mechanical evaluation at 4K on numerous samples which were supplied from industries. JAEA contributed standardization of these testing methods at 4K specified in the Japanese industrial standards (JIS). JAEA also supported to establish a construction code for structure of superconducting coil for fusion facility at the Japan Society of Mechanical Engineer (JSME), which is used in manufacture of the ITER toroidal field coil. This paper describes history over 30 years on the material development.
Sn superconducting strands for ITER toroidal field coilsNabara, Yoshihiro; Nunoya, Yoshihiko; Isono, Takaaki; Hamada, Kazuya; Uno, Yasuhiro; Takahashi, Yoshikazu; Nakajima, Hideo; Tsuzuku, Seiji*; Tagawa, Kohei*; Miyashita, Katsumi*; et al.
Teion Kogaku, 47(3), p.140 - 146, 2012/03
no abstracts in English
Koizumi, Norikiyo; Matsui, Kunihiro; Shimizu, Tatsuya; Nakajima, Hideo; Iijima, Ami*; Makino, Yoshinobu*
Teion Kogaku, 47(3), p.186 - 192, 2012/03
In the ITER TF coil, cover plates (CP) are welded to teeth of a radial plate (RP) to fix a conductor in the groove of the RP. Though total length of welds is approximately 1.5 km and height and width of an RP are 14 m and 9 m, respectively, the welding deformation better than 1 mm in local distortion and several milli-meters in in-plane deformation is required. Therefore, laser welding is used for the CP welding to reduce welding deformation as possible. However, a gap of a welding joint is expected to be 0.5 mm at the maximum. Therefore, at first, the laser welding technique to enable welding on the gap of 0.5 mm width is developed in this study. Using this technology, CP welding trial using an RP mock was successfully performed. The achieved flatness is 0.6 mm. In addition, the welding deformation of a full-scale RP is estimated by the analysis using the inherent strain. The results show that the flatness of 1 mm is achievable and profile of 5 mm can be achieved. Since the in-plane deformation can be corrected by welding somewhere to originate artificial welding deformation, it is confident that the required tolerance of several milli-meters in in-plan profile is achievable.
Koizumi, Norikiyo; Nunoya, Yoshihiko; Nakajima, Hideo
Teion Kogaku, 47(3), p.135 - 139, 2012/03
no abstracts in English
Iguchi, Masahide; Chida, Yutaka; Nakajima, Hideo; Ogawa, Tsuyoshi*; Katayama, Yoshinori*; Ogata, Hiroshige*; Minemura, Toshiyuki*; Miyabe, Keisuke*; Tokai, Daisuke*; Niimi, Kenichiro*
Teion Kogaku, 47(3), p.193 - 199, 2012/03
Japan Atomic Energy Agency (JAEA) has conducted qualification and rationalization activities in Japan in order to rationalize manufacturing procedure of ITER Toroidal Field (TF) coil structures. The activities included qualification of structural materials and qualification of welding procedure according to Japan Society of Mechanical Engineers (JSME) code constituted for fusion devices, demonstration of the manufacturing method and procedures through full-scale segments of TF coil structure. From results of these activities, JAEA confirmed applicability of JSME code to actual series TF coil structures as quality control method hence the quality of structural materials and weld joints of Gas Tungsten Arc Welding (GTAW) were satisfied ITER requirement. In addition, JAEA obtained knowledge of welding deformation of actual TF coil structures. This paper describes results of these qualification and development activities for TF coil structure.
Matsui, Kunihiro; Koizumi, Norikiyo; Hemmi, Tsutomu; Nakajima, Hideo; Kimura, Satoshi*; Shimada, Mamoru*
Teion Kogaku, 47(3), p.160 - 165, 2012/03
no abstracts in English
Hamada, Kazuya; Takahashi, Yoshikazu; Nabara, Yoshihiro; Kawano, Katsumi; Ebisawa, Noboru; Oshikiri, Masayuki; Tsutsumi, Fumiaki; Saito, Toru*; Nakajima, Hideo; Matsuda, Hidemitsu*; et al.
Teion Kogaku, 47(3), p.153 - 159, 2012/03
The Japan Atomic Energy Agency (JAEA) has the responsibility to procure 25% of the ITER Toroidal Field coil conductors as the Japanese Domestic Agency (JADA) in the ITER project. The TF conductor is a circular shaped, cable-in-conduit conductor, composed of a cable and a stainless steel conduit (jacket). The outer diameter and maximum length of the TF conductor are 43.7 mm and 760 m, respectively. JAEA has constructed newly conductor manufacturing facility. Prior to starting conductor, JAEA manufactured a 760-m long Cu dummy conductor as process qualification of dummy cable, the jacket sections and fabrication procedures, such as welding, cable insertion, compaction and spooling. Following qualification of all manufacturing processes, JAEA has started to fabricate superconducting conductors for the TF coils.
Matsui, Kunihiro; Hemmi, Tsutomu; Koizumi, Norikiyo; Nakajima, Hideo; Kimura, Satoshi*; Nakamoto, Kazunari*
Teion Kogaku, 47(3), p.166 - 171, 2012/03
no abstracts in English
Takano, Katsutoshi; Koizumi, Norikiyo; Shimizu, Tatsuya; Nakajima, Hideo; Esaki, Koichi*; Nagamoto, Yoshifumi*; Makino, Yoshinobu*
Teion Kogaku, 47(3), p.178 - 185, 2012/03
In the ITER TF coil, the tight tolerances of 1 mm in flatness and a few mm in profile are required in manufacturing a radial plate (RP), although the height and width of an RP are 13 m and 9 m, respectively. In addition, since cover plates (CP) should be fitted to a groove of an RP with tolerance of 0.5 mm, the tight tolerances are also required to a CP. Thus, we can conclude that the manufacturing procedure of the RP and CP has been demonstrated.
Hemmi, Tsutomu; Matsui, Kunihiro; Koizumi, Norikiyo; Nakajima, Hideo; Iijima, Ami*; Sakai, Masahiro*
Teion Kogaku, 47(3), p.172 - 177, 2012/03
The manufacturing process for the ITER Toroidal Field (TF) coil has to be demonstrated for the qualification. Since the impregnation of its insulation system is one of the qualifications, the authors performed impregnation test using cyanate-ester and epoxy blended resin, which is a candidate among resins because of its excellent resistance to radiation. To establish the insulation and impregnation procedure of the TF coil manufacturing, three types of trials were performed. (1) Impregnation tests using an acrylic model to fix the impregnation conditions; (2) Impregnation test using a metallic model to confirm that no void remains in the insulation layer after the curing in the D-shaped configuration; and (3) Insulation and impregnation trials using 1/3 scale double pancake (DP) to establish the insulation and impregnation procedure for the TF coil manufacturing, The procedure of the insulation and impregnation for the ITER TF coil was established from the results of these trials.
Nagamoto, Yoshifumi*; Osemochi, Koichi*; Shimada, Mamoru*; Senda, Ikuo*; Koizumi, Norikiyo; Chida, Yutaka; Iguchi, Masahide; Nakajima, Hideo
Teion Kogaku, 47(3), p.200 - 205, 2012/03
Based on the results of the sub- and full-scale trials, the TF coil and TF coil structure manufacturing procedures were considered. Radial plate (RP) will be manufactured by assembling 10 sets of segments with Laser Beam Welding. Cover plates (CPs) will be manufactured by 3 different methods, depending on their geometry. For Winding pack (WP), winding system to enable to measurement of the conductor length with the accuracy of 0.01% for serial production was designed. Assembling procedure and groove types of narrow gap TIG welding for coil structures were determined. Hereafter, technical improvements will be considered in order to aim for further optimization of manufacturing.
Sn cables for ITER toroidal field coilsIsono, Takaaki; Tsutsumi, Fumiaki; Nunoya, Yoshihiko; Matsui, Kunihiro; Takahashi, Yoshikazu; Nakajima, Hideo; Ishibashi, Tatsuji*; Sato, Go*; Chida, Keiji*; Suzuki, Rikio*; et al.
Teion Kogaku, 47(3), p.147 - 152, 2012/03
no abstracts in English
Al cable-in-conduit conductorKoizumi, Norikiyo; Matsui, Kunihiro; Hemmi, Tsutomu; Nakajima, Hideo; Takeuchi, Takao*; Banno, Nobuya*; Kikuchi, Akihiro*
Teion Kogaku, 46(8), p.495 - 499, 2011/08
JAEA and NIMS has been collaborating in development of high performance Rapid-Heating-Quenching-Transformation (RHQT) NbAl CIC conductor, aiming at the application of this conductor to DEMO plant. The technical issue of the RHQT NbAl strand in the application to a fusion magnet is stabilization against perturbation. NIMS developed technique to attach copper stabilizer by using electroplating and a sub-scale CIC conductor is developed using this conductor. JAEA performed stability test of the developed CIC conductor to demonstrate efficiency of this copper stabilization technique. The measured stability margin is sufficiently high compared to the one of similar NbTi CIC conductor previously tested by the authors. Then, it can be concluded that the copper stabilizer works efficiently from view point of stability, resulting in solving the remained technical issues in the RHQT CIC conductor. Therefore, we can say that the RHQT NbAl CIC conductor is the most promising candidate for application to magnet in DEMO plant.
Nakazawa, Shinobu*; Teshima, Shotaro*; Arai, Daichi*; Miyagi, Daisuke*; Tsuda, Makoto*; Hamajima, Takataro*; Yagai, Tsuyoshi*; Nunoya, Yoshihiko; Koizumi, Norikiyo; Takahata, Kazuya*; et al.
Teion Kogaku, 46(8), p.474 - 480, 2011/08
It is observed that measured critical currents of a large current CIC conductor sample become lower than expected ones, since unbalanced current distribution is caused through contacting resistances between strands and Cu sleeves in CIC conductor joints. In order to evaluate the contacting length, we identify all strands 3 dimensional positions in the CIC conductor, and then we measure contacting number and lengths of strands which appear on surface of the cable for contacting with the Cu sleeves. It is found that some strands do not appear on the surface of cable and the contacting lengths are widely distributed with large standard deviation. We develop a numerical code which simulates strand positions in the CIC, and then compare the analyzed contacting strand number and contacting length with measured ones. It is found that the both results are in good agreement and hence the code is available for evaluating the contacting parameters. We vary twist pitches of sub-cables to search the contacting parameters and then show that all strands appear on the cable surface and have contacting lengths with small standard deviation. It is found that the twist pitches are a key parameter for optimization of the contacting parameters.
Kamiya, Koji; Murakami, Haruyuki; Kizu, Kaname; Ichige, Toshikatsu; Yoshida, Kiyoshi
Teion Kogaku, 46(1), p.10 - 17, 2011/01
JT-60SA project replaces the JT-60U tokamak with a full superconducting tokamak. The helium refrigerator cools the superconducting coils by circulating 4.4 K, 0.6 MPa supercritical helium in the circulation loop at certain mass flow rate. Since the cooling power of the helium refrigerator is determined by the heat load of the superconducting coils, estimation of the heat generation and required mass flow rate to acquire sufficient temperature margin is of crucial importance. In this paper, optimizing the mass flow rate in the superconducting coils was attempted to satisfy 1 K temperature margin. Then, it is shown that the consequent maximum pressure drop in the circulation loop is 81 kPa to result in minimizing the heat load of the supercritical helium circulation pump.
Tatsumoto, Hideki; Aso, Tomokazu; Kato, Takashi; Otsu, Kiichi
Teion Kogaku, 45(4), p.181 - 190, 2010/04
To mitigate pressure fluctuation caused by the load, a pressure control system is a necessary requirement. Accordingly, a control system was designed and installed, using a heater and an accumulator. Changes in pressure caused by operation of a 120-kW and a 302-kW proton beam were studied. It was confirmed that the pressure control system was effective in mitigating the pressure fluctuation caused by the load. A simulation code was also developed and the pressure rise behavior and the accumulator variation were studied. The simulation results indicated good agreement with the experimental data within 10%. The pressure fluctuation for a 1-MW proton beam was predicted to be 33.9 kPa, which is lower than the allowable pressure rise of 0.1 MPa, and produced an accumulator variation of 11.35 mm. We believe that the pressure control system is effective for use with the operation of a 1-MW proton beam.
Koizumi, Norikiyo
Teion Kogaku, 44(11), p.496 - 497, 2009/11
no abstracts in English
