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Kenzhina, I.*; Ishitsuka, Etsuo; Ho, H. Q.; Sakamoto, Naoki*; Okumura, Keisuke; Takemoto, Noriyuki; Chikhray, Y.*
Fusion Engineering and Design, 164, p.112181_1 - 112181_5, 2021/03
Tritium release into the primary coolant during operation of the JMTR (Japan Materials Testing Reactor) and the JRR-3M (Japan Research Reactor-3M) had been studied. It is found that the recoil release by 
Li(n
,
)
H reaction, which comes from a chain reaction of beryllium neutron reflectors, is dominant. To prevent tritium recoil release, the surface area of beryllium neutron reflectors needs to be minimum in the core design and/or be shielded with other material. In this paper, as the feasibility study of the tritium recoil barrier for the beryllium neutron reflectors, various materials such as Al, Ti, V, Ni, and Zr were evaluated from the viewpoint of the thickness of barriers, activities after long-term operations, and effects on the reactivities. From the results of evaluations, Al would be a suitable candidate as the tritium recoil barrier for the beryllium neutron reflectors.
Kenzhina, I.*; Ishitsuka, Etsuo; Okumura, Keisuke; Ho, H. Q.; Takemoto, Noriyuki; Chikhray, Y.*
Journal of Nuclear Science and Technology, 58(1), p.1 - 8, 2021/01
Times Cited Count：1 Percentile：7.24(Nuclear Science & Technology)The sources and mechanisms for the tritium release into the primary coolant in the JMTR and the JRR-3M containing beryllium reflectors are evaluated. It is found that the recoil release from chain reaction of 
Be is dominant and its calculation results agree well with trends derived from the measured variation of tritium concentration in the primary coolant. It also indicates that the simple calculation method used in this study for the tritium recoil release from the beryllium reflectors can be utilized for an estimation of the tritium release into the primary coolant for a research and testing reactors containing beryllium reflectors.
Ishitsuka, Etsuo; Sakamoto, Naoki*
Physical Sciences and Technology, 6(2), p.60 - 63, 2019/12
Tritium release into the primary coolant of the research and test reactors during operation had been studied, and it is found that the recoil release from chain reaction of Be is dominant. To reduce tritium concentration of the primary coolant, feasibility study of the tritium recoil barrier for the beryllium neutron reflectors was carried out, and the tritium recoils of various materials were calculated by PHITS. From these calculation results, it is clear that the thickness of tritium recoil barrier depends on the material and 20
40 
m is required for three orders reduction.
Watanabe, Masao; Nojiri, Hiroyuki*; Ito, Shinichi*; Kawamura, Seiko; Kihara, Takumi*; Masuda, Takatsugu*; Sahara, Takuro*; Soda, Minoru*; Takahashi, Ryuta
JPS Conference Proceedings (Internet), 25, p.011024_1 - 011024_5, 2019/03
Recently, neutron scattering experiments have been rapidly progressed under high magnetic field. In the J-PARC, proto-type compact pulse magnet system with the power supply, the coil and the sample stick has been developed. Basic specifications of the power supply are as follows; maximum charged voltage with capacitor is 2 kV, maximum current is 8 kA, repetition rate is a pulse per several minutes and pulse duration is several msec. Maximum magnetic field in the coil is more than 30 Tesla. The sample stick is designed for Orange-Cryostat. In this presentation, We report the details of the pulsed magnet system and the performance of it on neutron scattering experiments at MLF beam line (HRC).
Ishitsuka, Etsuo; Kenzhina, I.*; Okumura, Keisuke; Ho, H. Q.; Takemoto, Noriyuki; Chikhray, Y.*
JAEA-Technology 2018-010, 33 Pages, 2018/11
JAEA-Technology-2018-010.pdf:2.58MB
As a part of study on the mechanism of tritium release to the primary coolant in research and testing reactors, tritium recoil release rate from Li and U impurities in the neutron reflector made by beryllium, aluminum and graphite were calculated by PHITS code. On the other hand, the tritium production from Li and U impurities in beryllium neutron reflectors for JMTR and JRR-3M were calculated by MCNP6 and ORIGEN2 code. By using both results, the amount of recoiled tritium from beryllium neutron reflectors were estimated. It is clear that the amount of recoiled tritium from Li and U impurities in beryllium neutron reflectors are negligible, and 2 and 5 orders smaller than that from beryllium itself, respectively.
Sakurai, Takeru; Iguchi, Masahide; Nakahira, Masataka; Inagaki, Takashi; Matsui, Kunihiro; Koizumi, Norikiyo
Fusion Engineering and Design, 109-111(Part B), p.1592 - 1597, 2016/11
Times Cited Count：8 Percentile：53.18(Nuclear Science & Technology)Japan Atomic Energy Agency (JAEA) has responsibility to procure 9 Toroidal Field (TF) coils and 19 TF coil structures for ITER. A TF coil structure consists of the main body structure having a D-shape with 16.5 m in height and 9m in width in which superconducting winding is stored and the components to connect adjacent TF coil or other ITER devices. TF coil structures are required the very tight tolerance which is less than 2 mm for the final dimension, which is quite challenging considering large size of TF coil structure. To achieve this tolerance, extra material will be put on the each material, and machining must be performed after welding. It is important to figure out detail welding deformation and reducing the machining process to optimize manufacturing. JAEA performed an additional manufacturing trial of A1 segment which is part of TF coil structure. JAEA adopted balance welding instead of using strong restriction jig welding in additional trial. The angular distortion of previous result was +6.5/+8.9mm, however angular distortions of latest trial were -3.0/+1.6mm (right side) and 0.0/+2.4mm (left side). This progress shows that welding deformation could be controlled closer in the target value (0.0 mm) than previous method applied. Based on latest knowledge, JAEA started actual TF coil structure manufacturing from April 2014. Actual manufacturing is steadily progressing with development process improvement by learning effect and improvement of manufacturing sequence.
Ishitsuka, Etsuo; Kenzhina, I. E.*; Okumura, Keisuke; Takemoto, Noriyuki; Chikhray, Y.*
JAEA-Technology 2016-022, 35 Pages, 2016/10
JAEA-Technology-2016-022.pdf:3.73MB
As a part of study on the mechanism of tritium release to the primary coolant in research and testing reactors, the calculation methods by PHITS code is studied to evaluate the recoil tritium release rate from beryllium core components. Calculations using neutron and triton sources were compared, and it is clear that the tritium release rates in both cases show similar values. However, the calculation speed for the triton source cases is two orders faster than that for the neutron source case. It is also clear that the calculation up to history number per unit volume of 2
10
(cm
) is necessary to determine the recoil tritium release rate of two effective digits precision. Furthermore, the relationship between the beryllium shape and recoil tritium release rate using the triton sources was studied. Recoil tritium release rate showed linear relation to the surface area per volume of beryllium, and the recoil tritium release rate showed about half of the conventional equation value.
Sakurai, Takeru; Iguchi, Masahide; Nakahira, Masataka; Saito, Toru; Koizumi, Norikiyo
IEEE Transactions on Applied Superconductivity, 26(4), p.4204705_1 - 4204705_5, 2016/06
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."
Novello, L.*; Cara, P.*; Coletti, A.*; Gaio, E.*; Maistrello, A.*; Matsukawa, Makoto; Philipps, G.*; Tomarchio, V.*; Yamauchi, Kunihito
IEEE Transactions on Applied Superconductivity, 26(2), p.4700507_1 - 4700507_7, 2016/03
Times Cited Count：9 Percentile：42.32(Engineering, Electrical & Electronic)Sukegawa, Atsuhiko; Murakami, Haruyuki; Matsunaga, Go; Sakurai, Shinji; Takechi, Manabu; Yoshida, Kiyoshi; Ikeda, Yoshitaka
Fusion Engineering and Design, 98-99, p.2076 - 2079, 2015/10
Times Cited Count：3 Percentile：22.12(Nuclear Science & Technology)The JT-60SA project is a EU - JA satellite tokamak under Broader Approach in support of the ITER project. In-vessel coils are designed and assembled by JA. The resin-insulator is required to have a heat resistance against the baking temperature of vacuum vessel of 200
C (40000 hour). Thus the assessment of the heat load is fundamental for the design of the coils. However, the estimation of the lifetime of resin-insulator under the high-temperature region has not been examined. In the present study, the estimation of the lifetime of seven candidate resin-insulators such as epoxy resin and cyanate-ester resin under the 220C temperature region have been performed for the current coils design. Weight reduction of the seven candidate insulators was measured at different heating times under 180C, 200C and 220C environment using three thermostatic ovens, respectively. The reduction of the insulators has been used as input for Weibull-analysis towards Arrhenius-plot. Lifetime of the resins has been estimated for the first time at the high temperature region by the plot. Lifetime of the resin-insulators have been evaluated and discussed as well as the available temperature of the in-vessel coils.
Isono, Takaaki; Kawano, Katsumi; Ozeki, Hidemasa; Kajitani, Hideki; Koizumi, Norikiyo; Okuno, Kiyoshi; Minato, Tsuneaki*; Nishimiya, Hikaru*; Watabe, Yuki*; Sakamoto, Hiroo*; et al.
IEEE Transactions on Applied Superconductivity, 25(3), p.4201004_1 - 4201004_4, 2015/06
Times Cited Count：2 Percentile：13.25(Engineering, Electrical & Electronic)Zito, P.*; Lampasi, A.*; Novello, L.*; Matsukawa, Makoto; Shimada, Katsuhiro; Portesine, M.*; Fasce, F.*; Cinarelli, D.*; Dorronsoro, A.*; Vian, D.*
Proceedings of IEEE 15th International Conference on Environment and Electrical Engineering (IEEE-EEEIC 2015), p.156 - 160, 2015/06
Okuno, Kiyoshi; Nakajima, Hideo; Koizumi, Norikiyo
IEEE Transactions on Applied Superconductivity, 16(2), p.880 - 885, 2006/06
Times Cited Count：11 Percentile：50.47(Engineering, Electrical & Electronic)no abstracts in English
Abe, Kanako*; Nakajima, Hideo; Hamada, Kazuya; Okuno, Kiyoshi; Kakui, Hideo*; Yamaoka, Hiroto*; Maruyama, Naoyuki*
IEEE Transactions on Applied Superconductivity, 16(2), p.807 - 810, 2006/06
Times Cited Count：13 Percentile：54.50(Engineering, Electrical & Electronic)no abstracts in English
Isono, Takaaki; Koizumi, Norikiyo; Okuno, Kiyoshi; Kurihara, Ryoichi; Nishio, Satoshi; Tobita, Kenji
Fusion Engineering and Design, 81(8-14), p.1257 - 1261, 2006/02
Times Cited Count：6 Percentile：39.61(Nuclear Science & Technology)In order to realize an economically competitive power generation system, generation of a higher field is required. Toroidal Field (TF) coils of fusion DEMO plant at JAERI are required to generate magnetic field of 16 to 20 T. To realize this high field, advanced superconducting materials, such as Nb
Al and high temperature superconductor (HTS), are considered. HTS has enough performance in a 20-T field at 4 K, and a forced-cooled type HTS conductor using a silver alloy sheathed Bi-2212 round wire has been proposed. Required areas of superconductor, structure, stabilizer, coolant and insulator in the cross section of coil winding have been calculated. However, there are many technical issues to be solved, such as accurate temperature control during heat treatment in an atmosphere of oxygen. On the other hand, a large coil using NbAl has been developed by JAERI, and major technology to fabricate a 16-T NbAl coil was developed. Validity and issues of grading the winding area are discussed, and there is a possibility to increase a field up to around 17 T using the method.
Hoshino, Katsumichi; Yamamoto, Takumi; Tamai, Hiroshi; Oasa, Kazumi; Kawashima, Hisato; Miura, Yukitoshi; Ogawa, Toshihide; Shoji, Teruaki*; Shibata, Takatoshi; Kikuchi, Kazuo; et al.
Fusion Science and Technology, 49(2), p.139 - 167, 2006/02
Times Cited Count：2 Percentile：16.43(Nuclear Science & Technology)The main results obtained by the various heating and current drive systems, external coil system and divertor bias system are reviewed from the viewpoint of the advanced active control of the tokamak plasma. Also, the features of each system are described. The contribution of the JFT-2M in these areas are summarized.
Nakahira, Masataka; Takeda, Nobukazu
Hozengaku, 4(4), p.47 - 52, 2006/01
The technical structural standard for ITER (International Thermonuclear Experimental Fusion Reactor) should be innovative because of their quite different features of safety and mechanical components from nuclear fission reactors, and the necessity of introducing several new fabrication and examination technologies. Recognizing the international importance of Fusion Standard, Japan and ASME has started the cooperation development of the Fusion Standard. This paper shows the special features of ITER from view points of safety, design and fabrication, and proposes approach for development of the fusion standard.
test results of a D-shaped Nb
Al CICC coil fabricated with a react-and-wind process for the National Centralized TokamakAndo, Toshinari*; Kizu, Kaname; Miura, Yushi*; Tsuchiya, Katsuhiko; Matsukawa, Makoto; Tamai, Hiroshi; Ishida, Shinichi; Koizumi, Norikiyo; Okuno, Kiyoshi
Fusion Engineering and Design, 75-79, p.99 - 103, 2005/11
Times Cited Count：1 Percentile：9.89(Nuclear Science & Technology)no abstracts in English
Hamada, Kazuya; Nakajima, Hideo; Takano, Katsutoshi*; Kudo, Yusuke; Tsutsumi, Fumiaki*; Okuno, Kiyoshi; Jong, C.*
Fusion Engineering and Design, 75-79, p.87 - 91, 2005/11
Times Cited Count：10 Percentile：54.93(Nuclear Science & Technology)no abstracts in English
Iida, Hiromasa; Petrizzi, L.*; Khripunov, V.*; Federici, G.*; Polunovskiy, E.*
Fusion Engineering and Design, 75(1-4), p.133 - 139, 2005/11
The design of the ITER machine was presented in 2001. A nuclear analysis has been performed on ITER by means of the most detailed models and the best assessed nuclear data and codes. As the construction phase of ITER is approaching, the design of the main components has been optimized/finalized and several minor design changes/optimizations have been made, which required refined calculations to confirm that nuclear design requirements are met. Some of the proposed design changes have been made to mitigate critical radiation shielding problems. This paper reviews some of the most recent neutronic work with emphasis on critical nuclear responses in the TF coil inboard legs and vacuum vessel related to design modifications made to the blanket modules and vacuum vessel.
