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Journal Articles

Evaluation of thermal strain induced in components of Nb$$_{3}$$Sn strand during cooling

Suwa, Tomone*; Hemmi, Tsutomu*; Saito, Toru*; Takahashi, Yoshikazu*; Koizumi, Norikiyo*; Luzin, V.*; Suzuki, Hiroshi; Harjo, S.

IEEE Transactions on Applied Superconductivity, 28(3), p.6001104_1 - 6001104_4, 2018/04

 Times Cited Count:0 Percentile:100(Engineering, Electrical & Electronic)

Journal Articles

Evaluation of bending strain in Nb$$_{3}$$Sn strands of CIC conductor using neutron diffraction

Hemmi, Tsutomu*; Harjo, S.; Kajitani, Hideki*; Suwa, Tomone*; Saito, Toru*; Aizawa, Kazuya; Osamura, Kozo*; Koizumi, Norikiyo*

IEEE Transactions on Applied Superconductivity, 27(4), p.4200905_1 - 4200905_5, 2017/06

 Times Cited Count:0 Percentile:100(Engineering, Electrical & Electronic)

Journal Articles

Development of evaluation procedure for critical current of periodically bent Nb$$_{3}$$Sn strand

Kajitani, 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 Nb$$_{3}$$Sn strand is applied to an ITER TF coil. The Nb$$_{3}$$Sn 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 Nb$$_{3}$$Sn 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.

Journal Articles

Accuracy of prediction method of cryogenic tensile strength for austenitic stainless steels in ITER toroidal field coil structure

Sakurai, Takeru; Iguchi, Masahide; Nakahira, Masataka; Saito, Toru*; Morimoto, Masaaki*; Inagaki, Takashi*; Hong, Y.-S.*; Matsui, Kunihiro; Hemmi, Tsutomu; Kajitani, Hideki; et al.

Physics Procedia, 67, p.536 - 542, 2015/07

 Times Cited Count:1 Percentile:45.55

Japan Atomic Energy Agency (JAEA) has developed the tensile strength prediction method at liquid helium temperature (4K) using the quadratic curve as a function of the content of carbon and nitrogen in order to establish the rationalized quality control of the austenitic stainless steel used in the ITER superconducting coil operating at 4K. ITER is under construction aiming to verify technical demonstration of a nuclear fusion generation. Toroidal Field Coil (TFC), one of superconducting system in ITER, have been started procurement of materials in 2012. JAEA is producing materials for actual product which are the forged materials with shape of rectangle, round bar, asymmetry and etc. JAEA has responsibility to procure all ITER TFC Structures. In this process, JAEA obtained many tensile strength of both room temperature and 4K about these structural materials, for example, JJ1: High manganese stainless steel for structure (0.03C-12Cr-12Ni-10Mn-5Mo- 0.24N) and 316LN: High nitrogen containing stainless steel (0.2Nitrogen). Based on these data, accuracy of 4K strength prediction method for actual TFC Structure materials was evaluated and reported in this study.

Journal Articles

Behavior of Nb$$_{3}$$Sn cable assembled with conduit for ITER central solenoid

Nabara, Yoshihiro; Suwa, Tomone; Takahashi, Yoshikazu; Hemmi, Tsutomu; Kajitani, Hideki; Ozeki, Hidemasa; Sakurai, Takeru; Iguchi, Masahide; Nunoya, Yoshihiko; Isono, Takaaki; et al.

IEEE Transactions on Applied Superconductivity, 25(3), p.4200305_1 - 4200305_5, 2015/06

 Times Cited Count:0 Percentile:100(Engineering, Electrical & Electronic)

Journal Articles

Development of laser welding technology for fully austenite stainless steel

Takano, Katsutoshi; Koizumi, Norikiyo; Serizawa, Hisashi*; Tsubota, Shuho*; Makino, Yoshinobu*

Yosetsu Gakkai Rombunshu (Internet), 33(2), p.126 - 132, 2015/06

A radial plates (RP), which is used in Toroidal field (TF) coil in ITER, is significantly large, such as 13 m height and 9 m wide, but thin, such as 10 cm thick, and are made of full-austenite stainless steel. Even though they are very large structures, high manufacturing tolerances are required. In addition, it is required that each RP is fabricated every three weeks. Therefore, the authors develop efficient manufacturing methods of RP. The laser welding is selected as a welding method of RP. But the development of the high power laser welding technology is necessary to avoid hot cracking of the materials used for RP, namely full austenite stainless steel with high nitrogen content. The authors carried out trial aiming at an application of the laser welding to RP. As a result, it is effective to optimize the angle of inclination of the weld head. It also seems sensitivity of hot cracking can be less by optimizing the chemical composition of materials to use for RP. It was therefore demonstrated that the application of the laser welding technology in the full austenite stainless steel.

Journal Articles

Welding joint design of ITER toroidal field coil structure under cryogenic environment

Iguchi, Masahide; Sakurai, Takeru; Nakahira, Masataka; Koizumi, Norikiyo; Nakajima, Hideo

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05

Application of partial penetration welding (PPW) to ITER Toroidal Field Coil structure has been proposed because of limited accessability for weld due to complex geometry and low stress and low importance components. In order to obtain fatigue crack growth (FCG) behavior of PPW joint in cryogenic environment, Japan Atomic Energy Agency performed FCG test at 4K by using Compact Tension (CT) specimens having as-weld notch of PPW. These CT specimens were made from mockups having one of actual joint shape of PPW, double J-groove. As the result of this test, it was observed that crack propagated in weld metal having inclination from as-weld notch. Moreover it was shown that FCG rate of as-weld CT specimens had high FCG rate region in early stage of crack propagation due to residual stress distribution. In addition, application method of this FCG rate to designing of PPW joint was proposed and verified in this study.

Journal Articles

Development of ITER superconducting coil in Japan

Koizumi, Norikiyo; Nunoya, Yoshihiko

FSST News, (143), p.6 - 10, 2014/10

no abstracts in English

Journal Articles

Effects of high-pressure annealing on critical current density in 122 type iron pnictide wires

Pyon, S.*; Tsuchiya, Yuji*; Inoue, Hiroshi*; Koizumi, Norikiyo; Kajitani, Hideki; Tamegai, Tsuyoshi*

Physica C, 504, p.69 - 72, 2014/09

 Times Cited Count:3 Percentile:80.67(Physics, Applied)

no abstracts in English

Journal Articles

Manufacturing technology and material properties of high nitrogen austenitic stainless steel forgings for ITER TF coil cases

Oshikawa, Takumi*; Funakoshi, Yoshihiko*; Imaoka, Hiroshi*; Yoshikawa, Kohei*; Maari, Yasutaka*; Iguchi, Masahide; Sakurai, Takeru; Nakahira, Masataka; Koizumi, Norikiyo; Nakajima, Hideo

Proceedings of 19th International Forgemasters Meeting (IFM 2014), p.254 - 259, 2014/09

ITER is a large-scale experiment that aims to demonstrate that it is possible to produce commercial energy from fusion. ITER Toroidal Field Coil Case (hereinafter referred to as "ITER TFCC") is one of the important components of ITER. The ITER TFCC materials are made of high nitrogen austenitic stainless steel and having various configurations. The ITER TFCC material which manufactured by JCFC has a complex configuration with heaver thickness than other materials. It is difficult to form near net shape to delivery configuration by ordinary open die forging method such as upset and stretching, because the ITER TFCC materials manufactured by JCFC have a complex configuration. Therefore ingot weight and lead time of machining increase when ITER TFCC materials are forged by ordinary open die forging method. Moreover, in order to get good attenuation at Ultrasonic examination, it is necessarily to make fine and uniform grain of the material. However, it is impossible to control grain size of austenitic stainless steel by heat treatment. The grain becomes fine and uniform by only forging process with suitable condition. Therefore, JCFC has studied suitable forging method to become near net shape to delivery configuration and also to get fine grain of center of the material. Based on these result, ITER TFCC materials were manufactured. This innovative forging process led to reduce the weight of ingot compared with general forging. And it had good Ultrasonic attenuation. It was confirmed that the results of material test and nondestructive examination satisfied the requirements of Japan domestic agency (hereinafter referred to as "JADA"). Moreover, the test coupons were taken from center of thick part of product and used for various tests. As the result of tests, it was confirmed that results of material test satisfied the requirements of JADA. It is clear that this innovative forging method is very suitable process for manufacturing of ITER TFCC materials.

Journal Articles

Nuclear energy (Technical topic); Development of ITER toroidal field (TF) coil

Hemmi, Tsutomu; Kajitani, Hideki; Takano, Katsutoshi; Matsui, Kunihiro; Koizumi, Norikiyo

Yosetsu Gakkai-Shi, 83(6), p.497 - 502, 2014/09

JAEA, serving as the Japan Domestic Agency (JADA) in the ITER project, is responsible for the procurement of 9 TF coils. In the TF coil, the radial plate (RP) structure is selected to improve electrical and mechanical reliability of the electrical insulation. Since the superconductor is degraded by the bending strain of 0.1% after the reaction heat-treatment, the conductor is inserted into the RP after winding to D-shape and the heat-treatment. To insert the conductor into the RP, the winding and RP groove length must be controlled with accuracy of 0.02% (7 mm on the 1 turn of 34 m). Accordingly, the targets for solving this issue are as follows: (1) Development of manufacturing procedure of the RP; (2) Development of winding head to achieve highly accurate winding; (3) Estimation of the conductor elongation after the heat-treatment. Therefore, JAEA can establish manufacturing plan for the TF coil as a result of the R&D for these targets.

Journal Articles

Optimization of heat treatment of Japanese Nb$$_3$$Sn conductors for toroidal field coils in ITER

Nabara, Yoshihiro; Hemmi, Tsutomu; Kajitani, Hideki; Ozeki, Hidemasa; Suwa, Tomone; Iguchi, Masahide; Nunoya, Yoshihiko; Isono, Takaaki; Matsui, Kunihiro; Koizumi, Norikiyo; et al.

IEEE Transactions on Applied Superconductivity, 24(3), p.6000605_1 - 6000605_5, 2014/06

no abstracts in English

Journal Articles

Investigation of strand bending in the He-inlet during reaction heat treatment for ITER TF Coils

Hemmi, Tsutomu; Matsui, Kunihiro; Kajitani, Hideki; Okuno, Kiyoshi; Koizumi, Norikiyo; Ishimi, Akihiro; Katsuyama, Kozo

IEEE Transactions on Applied Superconductivity, 24(3), p.4802704_1 - 4802704_4, 2014/06

 Times Cited Count:0 Percentile:100(Engineering, Electrical & Electronic)

Japan Atomic Energy Agency (JAEA), as Japan Domestic Agency, has responsibility to procure nine ITER Toroidal Field (TF) coils. The TF coil winding consists of a Nb$$_{3}$$Sn Cable-In-Conduit conductor, a pair of joints and a He-inlet. The current capacity of 68 kA is required at the magnetic field of 7 T around the He-inlet region in the TF coil winding. During reaction heat-treatment, the compressive residual strain in Nb$$_{3}$$Sn cable is induced by the difference in the thermal expansion coefficients between the Nb$$_{3}$$Sn cable and stainless steel jacket. The strands bending in the Nb$$_{3}$$Sn cable of the He-inlet is anticipated since there is the compressive residual strain and a gap between the Nb$$_{3}$$Sn cable and the He-inlet to introduce SHE flow. If the strand is bent, the variation of mechanical behaviors, such as the elongation of He-inlet during the reaction heat-treatment and the thermally induced residual strain on the jacket around the He-inlet, are expected. To investigate the strands bending in the Nb$$_{3}$$Sn cable of the He-inlet, the following items are performed; (1) elongation measurement during reaction heat-treatment, (2) residual longitudinal strain measurement using strain gauges by sample cuttings, (3) nondestructive inspection on the cable and strands using high resolution X-ray CT, Detail of test results and investigation of the strands bending in the Nb$$_{3}$$Sn cable of the He-inlet are reported and discussed.

Journal Articles

Progress of manufacturing trials for the ITER toroidal field coil structures

Iguchi, Masahide; Morimoto, Masaaki; Chida, Yutaka*; Hemmi, Tsutomu; Nakajima, Hideo; Nakahira, Masataka; Koizumi, Norikiyo; Yamamoto, Akio*; Miyake, Takashi*; Sawa, Naoki*

IEEE Transactions on Applied Superconductivity, 24(3), p.3801004_1 - 3801004_4, 2014/06

 Times Cited Count:6 Percentile:56.74(Engineering, Electrical & Electronic)

no abstracts in English

Journal Articles

Fabrication of super electromagnetic coil support using HIP diffusion bonding

Takahashi, Masakazu*; Masuo, Hiroshige*; Takano, Katsutoshi; Koizumi, Norikiyo

Proceedings of 11th International Conference on Hot Isostatic Pressing (HIP 2014), 4 Pages, 2014/06

As part of the research and development of Tokamak-type nuclear fusion reactors, super electromagnetic coils are required to control the plasma reaction. The structure that supports the plasmas chamber is maintained at cryogenic temperatures in liquid helium and thus must be able to withstand the extremes of this type of environment. The most promising material for this support is, SUS316LNH and at present, the only fabrication method being employed is traditional machining from solid materials. This creates a large amount of waste material with extremely long fabrication time, due to the large size of the support at 13 m in height and 8 m in width. Therefore, it is thought that by combining machining with the HIP diffusion bonding process, both waste and fabrication time can be reduced. Although this method is still under development, it is believed that a reduction of about 50% in wasted material and an about 40% in machining time can be achieved.

Journal Articles

Comparison of $$J$$c characteristics in PIT wires based on BaFe$$_{2}$$As$$_{2}$$ with different substitutions

Tamegai, Tsuyoshi*; Pyon, S.*; Ding, Q. P.*; Inoue, Hiroshi*; Kobayashi, Hiroki*; Tsuchiya, Yuji*; Sun, Y.*; Kajitani, Hideki; Koizumi, Norikiyo

Journal of Physics; Conference Series, 507(2), p.022041_1 - 022041_4, 2014/05

 Times Cited Count:3 Percentile:17.01

The author studied the effect of chemical compound for superconducting performance ($$J$$c) of iron-based superconductors. As a result, it was revealed that BaFe$$_{2}$$As$$_{2}$$ showed high Jc. And more, BaFe$$_{2}$$As$$_{2}$$ showed higer $$J$$c such as 32,000A/cm$$^{2}$$ (4.2K) by being applied of 120 MPa of hot isostatic pressure (HIP). Thus, higher $$J$$c performance could be achieved.

Journal Articles

Development of manufacturing technology of radial plate in superconducting coil for fusion reactor by diffusion bonding by Hot Isostatic Pressing (HIP)

Takano, Katsutoshi; Koizumi, Norikiyo; Masuo, Hiroshige*; Natsume, Yoshihisa*

Yosetsu Gakkai Rombunshu (Internet), 32(1), p.8 - 14, 2014/03

The authors performed trial manufacture of the RP segments using a diffusion bonding method, namely Hot Isostatic Pressing (HIP). As a result of trials, it was clarified that even when HIPping is applied, the mechanical characteristic of base metal is not deteriorated. The machining period can be reduced by half compared with the traditional manufacturing method. On the other hand, mechanical strength at 4 K is degraded due to weak bonding, that is no grain growth through joint, by HIPping. However, additional test indicates promising possibility of much better joint by higher temperature and joint surface treated HIPpings. These results justified that RP segment manufacturing is not only possible, but it is a technically valid manufacturing method that satisfies all requirements.

Journal Articles

Residual strains in ITER conductors by neutron diffraction

Harjo, S.; Hemmi, Tsutomu; Abe, Jun; Gong, W.; Nunoya, Yoshihiko; Aizawa, Kazuya; Ito, Takayoshi*; Koizumi, Norikiyo; Machiya, Shutaro*; Osamura, Kozo*

Materials Science Forum, 777, p.84 - 91, 2014/02

 Times Cited Count:2 Percentile:18.34

Journal Articles

Evaluation of inter-laminar shear strength of GFRP composed of bonded glass/polyimide tapes and cyanate-ester/epoxy blended resin for ITER TF coils

Hemmi, Tsutomu; Nishimura, Arata*; Matsui, Kunihiro; Koizumi, Norikiyo; Nishijima, Shigehiro*; Shikama, Tatsuo*

AIP Conference Proceedings 1574, p.154 - 161, 2014/01

 Times Cited Count:3 Percentile:11.03

Japan Atomic Energy Agency (JAEA), as Japan Domestic Agency, has responsibility to procure 9 ITER Toroidal Field (TF) coils. The insulation system of the ITER TF coils consists of 3 layers of insulations, which are a conductor insulation, a double-pancake (DP) insulation and a ground insulation, composed of multi-layer glass/polyimide tapes impregnated a resin. The ITER TF coils are required to withstand an irradiation of 10 MGy from $$gamma$$-ray and neutrons since the ITER TF coils is exposed by first neutron ($$>$$ 0.1 MeV) of 10$$^{22}$$ n/m$$^{2}$$ during the operation of 20 years in the ITER. Cyanate-ester/epoxy blended resins and bonded glass/polyimide tapes are developed as insulation materials to realize the required radiation-hardness for the insulation of the ITER TF coils. To evaluate the radiation-hardness of the developed insulation materials, the inter-laminar shear strength of glass-fiber reinforced plastics (GFRP) fabricated using developed insulation materials is measured as one of most important mechanical properties before/after the irradiation in a fission reactor of JRR-3. As a result, it is demonstrated that the GFRPs using the developed insulation materials have a sufficient performance to apply the ITER TF coil insulation.

Journal Articles

Progress of ITER and JT-60SA magnet development in Japan

Koizumi, Norikiyo; Nunoya, Yoshihiko; Yoshida, Kiyoshi; Barabaschi, P.*

Physics Procedia, 58, p.232 - 235, 2014/00

 Times Cited Count:0 Percentile:100

no abstracts in English

340 (Records 1-20 displayed on this page)