Analysis on ex-vessel loss of coolant accident for a water-cooled fusion DEMO reactor

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

http://doi.org/10.1109/SOFE.2015.7482333

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.

Neutronics analysis for fusion DEMO reactor design

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.

Corrosion behavior of F82H exposed to high temperature pressurized water with a rotating apparatus

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

https://doi.org/10.1016/j.jnucmat.2014.07.067

R&D status on water cooled ceramic breeder blanket technology

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

https://doi.org/10.1016/j.fusengdes.2014.01.035

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.

Study of safety features and accident scenarios in a fusion DEMO reactor

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

https://doi.org/10.1016/j.fusengdes.2014.04.062

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.

Analysis of accident scenarios of a water-cooled tokamak DEMO

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

https://conferences.iaea.org/indico/contributionDisplay.py?contribId=397&sessionId=31&confId=46

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.

Rail deployment operation test for ITER blanket handling system with positioning misalignment

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

https://doi.org/10.1016/j.fusengdes.2013.02.086

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.

A Feasible DEMO blanket concept based on water cooled solid breeder

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.

Verification test results of a cutting technique for the ITER blanket cooling pipes

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

https://doi.org/10.1016/j.fusengdes.2012.02.108

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.

Engineering study of gas flow in breeder pebble bed for fusion blanket

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.