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Watanabe, Hitoshi; Nakano, Masanao; Fujita, Hiroki; Kono, Takahiko; Inoue, Kazumi; Yoshii, Hideki*; Otani, Kazunori*; Hiyama, Yoshinori*; Goto, Ichiro*; Kibe, Satoshi*; et al.
JAEA-Review 2014-040, 115 Pages, 2015/01
JAEA-Review-2014-040.pdf:4.26MB
Based on the regulations (the safety regulation of Tokai reprocessing plant, the safety regulation of nuclear fuel material usage facilities, the radiation safety rule, the regulation about prevention from radiation hazards due to radioisotopes, which are related with the nuclear regulatory acts, the local agreement concerning with safety and environment conservation around nuclear facilities, the water pollution control law, and bylaw of Ibaraki prefecture), the effluent control of liquid waste discharged from the Nuclear Fuel Cycle Engineering Laboratories of Japan Atomic Energy Agency has been performed. This report describes the effluent control results of the liquid waste in the fiscal year 2013. In this period, the concentrations and the quantities of the radioactivity in liquid waste discharged from the reprocessing plant, the plutonium fuel fabrication facilities, and the other nuclear fuel material usage facilities were much lower than the limits authorized by the above regulations.
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
Times Cited Count：21 Percentile：84.18(Nuclear Science & Technology)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.
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
Times Cited Count：35 Percentile：92.09(Nuclear Science & Technology)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.
Tobita, Kenji; Nishio, Satoshi*; Enoeda, Mikio; Nakamura, Hirofumi; Hayashi, Takumi; Asakura, Nobuyuki; Uto, Hiroyasu; Tanigawa, Hiroyasu; Nishitani, Takeo; Isono, Takaaki; et al.
JAEA-Research 2010-019, 194 Pages, 2010/08
JAEA-Research-2010-019-01.pdf:48.47MBJAEA-Research-2010-019-02.pdf:19.4MB
This report describes the results of the conceptual design study of the SlimCS fusion DEMO reactor aiming at demonstrating fusion power production in a plant scale and allowing to assess the economic prospects of a fusion power plant. The design study has focused on a compact and low aspect ratio tokamak reactor concept with a reduced-sized central solenoid, which is novel compared with previous tokamak reactor concept such as SSTR (Steady State Tokamak Reactor). The reactor has the main parameters of a major radius of 5.5 m, aspect ratio of 2.6, elongation of 2.0, normalized beta of 4.3, fusion out put of 2.95 GW and average neutron wall load of 3 MW/m
. This report covers various aspects of design study including systemic design, physics design, torus configuration, blanket, superconducting magnet, maintenance and building, which were carried out increase the engineering feasibility of the concept.
S
Nakao, Hironori*; Owada, Kenji; Shimomura, Susumu*; Ochiai, Akira*; Namikawa, Kazumichi*; Mizuki, Junichiro; Mimura, Hidekazu*; Yamauchi, Kazuto*; Murakami, Yoichi*
AIP Conference Proceedings 1234, p.935 - 938, 2010/06
Times Cited Count：2 Percentile：66.93(Physics, Applied)Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Kawashima, Hisato; Kurita, Genichi; Tanigawa, Hiroyasu; Nakamura, Hirofumi; Honda, Mitsuru; Saito, Ai*; Sato, Satoshi; et al.
Nuclear Fusion, 49(7), p.075029_1 - 075029_10, 2009/07
Times Cited Count：137 Percentile：97.72(Physics, Fluids & Plasmas)Recent design study on SlimCS focused mainly on the torus configuration including blanket, divertor, materials and maintenance scheme. For vertical stability of elongated plasma and high beta access, a sector-wide conducting shell is arranged in between replaceable and permanent blanket. The reactor adopts pressurized-water-cooled solid breeding blanket. Compared with the previous advanced concept with supercritical water, the design options satisfying tritium self-sufficiency are relatively scarce. Considered divertor technology and materials, an allowable heat load to the divertor plate should be 8 MW/m or lower, which can be a critical constraint for determining a handling power of DEMO (a combination of alpha heating power and external input power for current drive).
TiO
pebble bed for a test blanket module in JAEATanigawa, Hisashi; Hoshino, Tsuyoshi; Kawamura, Yoshinori; Nakamichi, Masaru; Ochiai, Kentaro; Akiba, Masato; Ando, Masami; Enoeda, Mikio; Ezato, Koichiro; Hayashi, Kimio; et al.
Nuclear Fusion, 49(5), p.055021_1 - 055021_6, 2009/05
Times Cited Count：23 Percentile：64.51(Physics, Fluids & Plasmas)This paper presents recent achievements of the research activities for the TBM being developed in JAEA, focusing on the pebble bed of the tritium breeder materials and tritium behaviour. For the breeder material, the chemical stability of LiTiO has been improved by LiO additives. In order to analyze the pebble bed behaviour, thermo-mechanical properties of the LiTiO pebble bed has been experimentally obtained. In order to verify nuclear properties of the pebble bed, the activation foil method has been proposed and a preliminary experiment has been conducted. For the tritium behaviour, the chemical densified coating method has been well developed and tritium recovery system has been modified taking account of the design change of the TBM.
Enoeda, Mikio; Tanigawa, Hisashi; Tsuru, Daigo; Hirose, Takanori; Ezato, Koichiro; Yokoyama, Kenji; Dairaku, Masayuki; Seki, Yohji; Suzuki, Satoshi; Mori, Kensuke*; et al.
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
Kiuchi, Kiyoshi; Ioka, Ikuo; Tanabe, Makoto*; Nanjo, Yoshiyasu*; Ogawa, Hiroaki; Ishijima, Yasuhiro; Tsukatani, Ichiro; Ochiai, Takamasa; Kizaki, Minoru; Kato, Yoshiaki; et al.
JAEA-Research 2006-023, 173 Pages, 2006/03
JAEA-Research-2006-023.pdf:20.51MB
The research concerning new cladding materials for ultra-high burnup of fuel elements with MOX fuels aiming at 100 GWd/t of BWR was pursued for 5 years from 2001 to 2005. On the Phase 1, the modified stainless steel of Fe-25Cr-35Ni-0.2Ti as fuel claddings and Nb-Mo alloy as a liner for inhibiting the pellet- clad interaction were selected as candidate materials, by evaluating fundamental properties required to BWR cladding materials, that are the nuclear economy, radioactivity, mass-transfer, irradiation properties, mechanical properties so on. On the present study, the making process of cladding tubes, lining by diffusion bonding, end plug by laser welding were developed and optimized, by considering the practical use of fuel elements consists of these candidates. The practical applicability was basically examined by irradiation tests using the accelerator of TIARA and the research reactor of JRR-3, for mainly confirming the resistance to IGSCC as one of the current important issues of BWR core materials of low carbon grade stainless steels. Creep and fatigue testing data were also obtained for evaluating the long performance of candidate materials. The behavior as fuel elements was analyzed with the safety calculation code for BWRs. The obtained results were established as a data base system, by considering the applicability to the fuel design and in-pile loop tests.
Maekawa, Fujio; Ochiai, Kentaro; Shibata, Keiichiro*; Kasugai, Yoshimi; Wada, Masayuki*; Morimoto, Yuichi*; Takeuchi, Hiroshi
Fusion Engineering and Design, 58-59, p.595 - 600, 2001/11
Times Cited Count：11 Percentile：61.99(Nuclear Science & Technology)no abstracts in English
Kugo, Teruhiko; Shimada, Shoichiro*; Okubo, Tsutomu; Ochiai, Masaaki
JAERI-Research 98-059, 40 Pages, 1998/10
JAERI-Research-98-059.pdf:1.73MB
no abstracts in English
Okubo, Tsutomu; Kugo, Teruhiko; *; Shimada, Shoichiro*; Ochiai, Masaaki
Proc. of Workshop on Advanced Reactors with Innovative Fuels, p.127 - 137, 1998/10
no abstracts in English
Ishii, Keiichi*; Shinohara, Koji; Takakura, Kosuke; Ochiai, Kentaro; Isobe, Mitsutaka*; Baba, Mamoru*; Kashiwa, Soichiro*; Okamoto, Atsushi*; Kitajima, Sumio*; Sasao, Mamiko*
no journal, ,
no abstracts in English
Enoeda, Mikio; Tanigawa, Hisashi; Hirose, Takanori; Suzuki, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Nakamichi, Masaru; et al.
no journal, ,
ITER Test Blanket Module (TBM) program is the first module scale breeding blanket performance test in the real DT fusion reaction environment which will be realized in ITER. Therefore, it is regarded as one of the most important development step of breeding blanket for DEMO. For the TBM program, Japan has the plan to test Water Cooled Ceramic Breeder (WCCB) TBM as the primary candidate breeding blanket. Keeping consistency with ITER construction activities, TBM design and development activities are being performed. This paper overviews the recent achievements of the development of the module fabrication technology and tritium production technology of the WCCB TBM in Japan.
Kato, Yoshinari; Hashimoto, Makoto; Miyauchi, Hideaki; Yasumune, Takashi; Ogawa, Ryuichiro; Goto, Shingo; Ochiai, Yukihiro*; Matsui, Junki
no journal, ,
no abstracts in English
