Watanabe, Kazuhito; Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Uto, Hiroyasu; Sakamoto, Yoshiteru; Araki, Takao*; Asano, Shiro*; Asano, Kazuhito*
Proceedings of 26th IEEE Symposium on Fusion Engineering (SOFE 2015), 6 Pages, 2016/06
Safety studies of a water-cooled fusion DEMO reactor have been performed. In the event of the blanket cooling pipe break outside the vacuum vessel, i.e. ex-vacuum vessel loss of coolant accident (ex-VV LOCA), the pressurized steam and air may lead to damage reactor building walls which have confinement function, and to release the radioactive materials to the environment. In response to this accident, we proposed three cases of confinement strategies. In each case, the pressure and thermal loads to the confinement boundaries and total mass of tritium released to outside the boundaries were analyzed by accident analysis code MELCOR modified for fusion reactor. These analyses developed design parameters to maintain the integrity of the confinement boundaries.
Ishizawa, Akihiro*; Idomura, Yasuhiro; Imadera, Kenji*; Kasuya, Naohiro*; Kanno, Ryutaro*; Satake, Shinsuke*; Tatsuno, Tomoya*; Nakata, Motoki*; Nunami, Masanori*; Maeyama, Shinya*; et al.
Purazuma, Kaku Yugo Gakkai-Shi, 92(3), p.157 - 210, 2016/03
The high-performance computer system Helios which is located at The Computational Simulation Centre (CSC) in The International Fusion Energy Research Centre (IFERC) started its operation in January 2012 under the Broader Approach (BA) agreement between Japan and the EU. The Helios system has been used for magnetised fusion related simulation studies in the EU and Japan and has kept high average usage rate. As a result, the Helios system has contributed to many research products in a wide range of research areas from core plasma physics to reactor material and reactor engineering. This project review gives a short catalogue of domestic simulation research projects. First, we outline the IFERC-CSC project. After that, shown are objectives of the research projects, numerical schemes used in simulation codes, obtained results and necessary computations in future.
Someya, Yoji; Tobita, Kenji; Tanigawa, Hisashi; Uto, Hiroyasu; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto; Tokunaga, Shinsuke
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
This paper presents neutronics analysis mainly focused on key design issues for self-sufficient tritium production based on the conceptual design study carried out for a fusion DEMO reactor in past several years, which includes new findings regarding design methodology of breeding blanket. Self-sufficient production of tritium is one of the most critical requirements for fusion reactors. We considered a fusion DEMO reactor with a major radius of about 8 m and fusion output of 1.5 GW with breeding blanket consisting of a mixed bed of LiTiO and BeTi pebbles. The net tritium breeding ratio (TBR) was estimated to be 1.15 with a three-dimensional analysis with the MCNP-5 with nuclear library of FENDL-2.1, satisfying a self-sufficient supply of tritium (net TBR1.05). Throughout the research, we found that tritium breeding capability (i.e., local TBR) of breeding blanket is less dependent on the arrangement of cooling pipe in the blanket when the neutron wall loading is lower than about 1.5 MW/m which is met in the DEMO considered. The result suggests that tolerance for the installation of cooling pipes in each blanket module will not be a critical matter. In addition, we found that a gap of about 0.02 m between neighboring blanket modules has little impact on the gross TBR.
Nakajima, Motoki; Hirose, Takanori; Tanigawa, Hisashi; Enoeda, Mikio
Journal of Plasma and Fusion Research SERIES, Vol.11, p.69 - 72, 2015/03
Water-cooled blanket is an attractive concept for its compactness and its compatibility with the conventional technologies for PWR. For blanket application, the structural material is required to be as thin as possible for tritium breeding. On the other hand, it is also required the pressure tightness to withstand 15 MPa of internal pressure. Therefore it is necessary to understand the corrosion mechanism in high temperature pressurized water. The effects of water flow and DO in the test water on corrosion properties were investigated using rotating disk specimen in autoclave. In summary, the weight loss by flowing was occurred except for test with DO 8 ppm, and it was more pronounced at lower DO concentration. Since FeO was observed on the specimen of small weight change, and the iron-poor layer thickness increased with decreasing the specimen weight, it seemed that the formation of FeO was effective for the suppression of weight loss.
Kanai, Akihiko*; Kasada, Ryuta*; Nakajima, Motoki; Hirose, Takanori; Tanigawa, Hisashi; Enoeda, Mikio; Konishi, Satoshi*
Journal of Nuclear Materials, 455(1-3), p.431 - 435, 2014/12
Hirose, Takanori; Nozawa, Takashi; Stoller, R. E.*; Hamaguchi, Dai; Sakasegawa, Hideo; Tanigawa, Hisashi; Tanigawa, Hiroyasu; Enoeda, Mikio; Kato, Yutai*; Snead, L. L.*
Fusion Engineering and Design, 89(7-8), p.1595 - 1599, 2014/10
The material properties, focusing on the properties used for design analysis were re-assessed and newly investigated for various heats including F82H-IEA. Moreover, irradiation effects on those properties were studied in this work. As for thermal properties, thermal conductivity that has significant impacts on the thermo-hydraulic properties of the blanket was investigated on several heats of F82H including F82H-IEA. According to the measurements, the thermal conductivity falls in the range 28.31.1 W/m/K at 293 K. Although this is comparable with that of the other ferritic/martensitic steels, it is 20% lower than the published value for F82H-IEA. The re-assessment on the published value revealed that the thermal diffusivity was over-estimated. As for irradiation effects on the physical properties, electric resistivity was measured after irradiation up to 6 dpa at 573 K and 673 K. The reduction of resistivity in F82H and its welds were 3% and 6%, respectively.
Enoeda, Mikio; Tanigawa, Hisashi; Hirose, Takanori; Nakajima, Motoki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko; et al.
Fusion Engineering and Design, 89(7-8), p.1131 - 1136, 2014/10
The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. Regarding the fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. Also the assembling of the complete box structure of the TBM mockup and planning of the pressurization testing was studied. The development of advanced breeder and multiplier pebbles for higher chemical stability was performed for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium simulation technology, investigation of the TBM neutronics measurement technology and the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed.
Sato, Satoshi; Tanigawa, Hisashi; Hirose, Takanori; Enoeda, Mikio; Ochiai, Kentaro; Konno, Chikara
Fusion Engineering and Design, 89(9-10), p.1984 - 1988, 2014/10
In order to evaluate nuclear properties of the ITER JA WCCB-TBM (Water Cooled Ceramic Breeder Test Blanket Module) and ensure that the design conforms to the nuclear regulation for licensing, nuclear analyses have been performed for the WCCB-TBM including flame, shield, pipe-forest, bio-shield and AEU (Ancillary Equipment Unit). Nuclear analyses are performed with the Monte Carlo code MCNP5.14, activation code ACT-4 and Fusion Evaluated Nuclear Data Library FENDL-2.1. MCNP geometry input data of the TBM is created from CAD data with the automatic conversion code GEOMIT, and other geometry input data is created by manually. By adopting the dog-leg gaps, decay -ray dose rate can be drastically reduced and hands-on access is possible for shield. Detailed calculation results will be presented in this symposium.
Nakamura, Makoto; Tobita, Kenji; Gulden, W.*; Watanabe, Kazuhito*; Someya, Yoji; Tanigawa, Hisashi; Sakamoto, Yoshiteru; Araki, Takao*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.
Fusion Engineering and Design, 89(9-10), p.2028 - 2032, 2014/10
After the Fukushima Dai-ichi nuclear accident, a social need for assuring safety of fusion energy has grown gradually in the Japanese (JA) fusion research community. DEMO safety research has been launched as a part of BA DEMO Design Activities (BA-DDA). This paper reports progress in the fusion DEMO safety research conducted under BA-DDA. Safety requirements and evaluation guidelines have been, first of all, established based on those established in the Japanese ITER site invitation activities. The amounts of radioactive source terms and energies that can mobilize such source terms have been assessed for a reference DEMO, in which the blanket technology is based on the Japanese fusion technology R&D programme. Reference event sequences expected in DEMO have been analyzed based on the master logic diagram and functional FMEA techniques. Accident initiators of particular importance in DEMO have been selected based on the event sequence analysis.
Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Gulden, W.*; Sakamoto, Yoshiteru; Araki, Takao*; Watanabe, Kazuhito*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.
Plasma and Fusion Research (Internet), 9, p.1405139_1 - 1405139_11, 2014/10
Key aspects of the safety study of a water-cooled fusion DEMO reactor is reported. Safety requirements, dose target, DEMO plant model and confinement strategy of the safety study are briefly introduced. The internal hazard of a water-cooled DEMO, i.e. radioactive inventories, stored energies that can mobilize these inventories and accident initiators and scenarios, are evaluated. It is pointed out that the enthalpy in the first wall/blanket cooling loops, the decay heat and the energy potentially released by the Be-steam chemical reaction are of special concern for the water-cooled DEMO. An ex-vessel loss-of-coolant of the first wall/blanket cooling loop is also quantitatively analyzed. The integrity of the building against the ex-VV LOCA is discussed.
Nakamura, Makoto; Ibano, Kenzo*; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Gulden, W.*; Ogawa, Yuichi*
Proceedings of 25th IAEA Fusion Energy Conference (FEC 2014) (CD-ROM), 8 Pages, 2014/10
Of late in Japan, a design study has been undertaken of a tokamak fusion DEMO with pressurized water coolant and solid pebble bed breeding blanket, but safety characteristics of this type of DEMO have not been well examined. In this paper, thermohydraulics analysis of in-vessel and ex-vessel loss-of-coolant accidents of a water-cooled tokamak DEMO is reported. Safety characteristics of water-cooled DEMO, particularly possible loads onto confinement barriers, are discussed based on the thermohydraulics analysis results. Measures to reduce such loads are also proposed.
Hirose, Takanori; Someya, Yoji; Tanigawa, Hisashi; Suzuki, Satoshi
Yosetsu Gakkai-Shi, 83(1), p.70 - 77, 2014/01
no abstracts in English
Takeda, Nobukazu; Aburadani, Atsushi; Tanigawa, Hisashi; Shigematsu, Soichiro; Kozaka, Hiroshi; Murakami, Shin; Kakudate, Satoshi; Nakahira, Masataka; Tesini, A.*
Fusion Engineering and Design, 88(9-10), p.2186 - 2189, 2013/10
R&D for rail deployment equipment was performed for the ITER blanket remote handling system. The target torque for the automatic operation was investigated. The result shows that the 20% of the rated torque is adequate as the torque limitation for the automatic operation. A schedule for the procurement of the blanket remote handling system, which will be delivered to the ITER in 2020, was also shown.
Aburadani, Atsushi; Takeda, Nobukazu; Shigematsu, Soichiro; Murakami, Shin; Tanigawa, Hisashi; Kakudate, Satoshi; Nakahira, Masataka*; Hamilton, D.*; Tesini, A.*
Fusion Engineering and Design, 88(9-10), p.1978 - 1981, 2013/10
no abstracts in English
Someya, Yoji; Tobita, Kenji; Uto, Hiroyasu; Hoshino, Kazuo; Asakura, Nobuyuki; Nakamura, Makoto; Tanigawa, Hisashi; Enoeda, Mikio; Tanigawa, Hiroyasu; Nakamichi, Masaru; et al.
Proceedings of 24th IAEA Fusion Energy Conference (FEC 2012) (CD-ROM), 8 Pages, 2013/03
This paper presents the conceptual design of a blanket with simplified structure whose interior consists of the mixture of breeder and multiplier pebble bed, cooling tubes and support for them only. Neutronics calculation indicated that the blanket satisfies a self-sufficient production of tritium. An important finding is that little decrease is seen in TBR even when the gap between neighboring blanket modules is as wide as 0.03 m. This means that blanket modules can be arranged with such a significant clearance gap without sacrifice of tritium production. On the other hand, the thickness of blanket housing is important from the viewpoint of safety. The blanket housing may rupture when the cooling pipe in the blanket is tearing, because thickness of structure materials is thin as 22 mm. This thickness is expected to maintain to 8 MPa in the steam pressure. Finally, the blanket housing, and aspect ratio of blanket shape is proposed in consideration of TBR, and engineering problem such as maintenance and manufacture are discussed.
Tanigawa, Hisashi; Aburadani, Atsushi; Shigematsu, Soichiro; Takeda, Nobukazu; Kakudate, Satoshi; Mori, Seiji*; Jokinen, T.*; Merola, M.*
Fusion Engineering and Design, 87(7-8), p.999 - 1003, 2012/08
This paper presents results of R&D activities where the laser and TIG welding tools were developed to apply the blanket hydraulic connection. The target pipe is 48.26 mm in outer diameter and 2.77 mm-thick. A single path welding without filler materials is required to reduce the weld heat input related to re-weldability. For the laser welding, the focal spot diameter was expanded to increase allowable misalignment. The TIG welding tool was equipped with AVC (Arc Voltage Control) to avoid a torch sticking and to enlarge allowable misalignment. For each tool, the welding conditions were optimized for all position welding to horizontally located pipes. Obtained parameters such as the weld heat input, allowable misalignment, lifetime of the tools and amount of sputter and fume, were comparatively assessed.
Shigematsu, Soichiro; Tanigawa, Hisashi; Aburadani, Atsushi; Takeda, Nobukazu; Kakudate, Satoshi; Mori, Seiji*; Nakahira, Masataka*; Raffray, R.*; Merola, M.*
Fusion Engineering and Design, 87(7-8), p.1218 - 1223, 2012/08
The current design of the ITER blanket system is a modular configuration and a total of 440 blanket modules are to be installed in the ITER vacuum vessel. Each blanket module consists of the first wall (FW) and the shield block (SB). The FW receives a high heat load from the plasma. The SB shields components from the neutrons generated by the nuclear fusion reaction. The FW will be damaged by the heat load and neutrons, so it requires scheduled replacement. For the FW replacement, cutting/welding tools for the cooling pipes must be able to conduct the following operations: access and cut/weld the pipe from the inside of the cooling pipe. The cutting tool for the pipe end is required to cut flat plate circularly from the surface side of the FW. This paper describes the current status of R&D of the cutting tools for maintenance of the cooling pipe of the FW.
Enoeda, Mikio; Tanigawa, Hisashi; Hirose, Takanori; Suzuki, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Nakamichi, Masaru; et al.
Fusion Engineering and Design, 87(7-8), p.1363 - 1369, 2012/08
The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and evaluation toward DEMO blanket, the module fabrication technology development by a candidate structural material, reduced activation martensitic/ferritic steel, F82H, is one of the most critical items from the viewpoint of realization of TBM testing in ITER. Fabrication of a real scale first wall, side walls, a breeder pebble bed box and assembling of the first wall and side walls have succeeded. Recently, the real scale partial mockup of the back wall was fabricated. The fabrication procedure of the back wall, whose thickness is up to 90 mm, was confirmed toward the fabrication of the real scale back wall by F82H. This paper overviews the recent achievements of the development of the WCCB TBM in Japan.
Seki, Yohji; Yoshikawa, Akira; Tanigawa, Hisashi; Hirose, Takanori; Ezato, Koichiro; Enoeda, Mikio; Sakamoto, Kensaku
Dai-17-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.265 - 266, 2012/06
In the case of a water cooled ceramic breeder in a blanket, pebbles of a ceramic tritium breeder are packed in a container constituted by a partition plate. Helium purge gas is applied as a transport fluid in a tritium recovery system. It is of importance to build database of a pressure drop as part of a design of the tritium recovery system. In this experimental study, the pressure drops of He gas through pebble bed were measured within the wide range of a flow rate up to 100 L/min. The results indicate that a laminar flow is dominant and the pressure drop was correctly predicted by the empirical equation within a part of flow rate. Reliability of prediction ability of pressure drop was established by this experiment within the flow rate which is less than 60 L/min. Moreover, this paper describes that slight difference between the experimental result and the empirical equation within a range of flow rate from 60 L/min to 100 L/min.
Uto, Hiroyasu; Someya, Yoji; Tanigawa, Hisashi
Purazuma, Kaku Yugo Gakkai-Shi, 88(5), P. 288, 2012/05
The 15th Workshop on Plasma Physics by Young Scientist was held at Naka Fusion Institute and organized by Division of Advanced Plasma Research and Tokamak System Technology. This workshop was focused on "Research for DEMO". 56 young scientists participated in the workshop, and 8 overview talks and 17 oral talks were given about plasma heating and control techniques. Because the participants had both of oral and poster presentations, the participants understood the study of each other and the discussion lasted during the workshop. On the last day of the workshop, the participants visited the JT-60SA vacuum vessel sector-assembly buildings, TFCs stored in the storage building and JT-60SA superconducting magnet fabrication building.