Evaluation of bending strain in NbSn 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

https://doi.org/10.1109/TASC.2017.2660064

Mechanical properties of welded joint at cryogenic temperature for manufacturing of ITER TF Coil Structure

Sakurai, Takeru; Iguchi, Masahide; Nakahira, Masataka; Saito, Toru; Koizumi, Norikiyo

IEEE Transactions on Applied Superconductivity, 26(4), p.4204705_1 - 4204705_5, 2016/06

http://dx.doi.org/10.1109/TASC.2016.2539686

Japan Atomic Energy Agency (JAEA) has responsibility to procure 9 Toroidal Field (TF) coils and 19 TF coil structures for ITER project. A TF coil structure consists of the main body structure having D-shape with 13.6 m of height and 9 m of width in which a superconducting winding pack is enclosed and the components to connect adjacent TF coils or other surrounding components. TF coil structures are manufactured from austenitic stainless steel having high tensile strength and fracture toughness at cryogenic temperature (4K) in order to ensure the huge electromagnetic force. As structural materials, austenitic stainless steel having high Manganese and Nitrogen which was named as JJ1 and high Nitrogen containing 316LN stainless steel are applied. These materials are welded each other by Tungsten Inert Gas (TIG) welding with FMYJJ1, which had been developed for welding material based on JJ1. The cryogenic mechanical properties of welded joints which have over 200 mm of actual thickness are limited due to less demonstration of such a heavy thick joints. In addition, destructive test specimens cannot be taken from actual TF coil structure. Hence, it is necessary to confirm actual thickness of welded joint performance by actual welding conditions mock-up. JAEA manufactured some welded joint mock-ups having the same welding thickness and combination of base materials as actual TF coil structure by applying actual welding conditions. JAEA measured mechanical properties of tensile and fracture toughness in liquid Helium environment by using test specimens taken from these welded joint mock-ups. This study reports these mechanical test results of welded joints at cryogenic temperature. "The view and opinions expressed herein do not necessarily reflect those of the ITER Organization."

Development of evaluation procedure for critical current of periodically bent NbSn strand

Kajitani, Hideki; Ishiyama, Atsushi*; Agatsuma, Ko*; Murakami, Haruyuki; Hemmi, Tsutomu; Koizumi, Norikiyo

Teion Kogaku, 50(12), p.608 - 615, 2015/12

https://doi.org/10.2221/jcsj.50.608

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.

Behavior of NbSn 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

https://doi.org/10.1109/TASC.2014.2360562

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

https://doi.org/10.2207/qjjws.33.126

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.

Development of ITER superconducting coil in Japan

Koizumi, Norikiyo; Nunoya, Yoshihiko

FSST News, (143), p.6 - 10, 2014/10

no abstracts in English

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.

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

https://doi.org/10.2207/jjws.83.497

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.

Optimization of heat treatment of Japanese NbSn 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

https://doi.org/10.1109/TASC.2013.2286675

no abstracts in English

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.