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Ueta, Shohei; Sasaki, Koei; Arita, Yuji*
Nihon Genshiryoku Gakkai-Shi ATOMO
, 63(8), p.615 - 620, 2021/08
no abstracts in English
Mitsumoto, Rika; Hazama, Taira; Takahashi, Keita; Kondo, Satoru
JAEA-Technology 2019-020, 167 Pages, 2020/03
JAEA-Technology-2019-020.pdf:21.06MBJAEA-Technology-2019-020-high-resolution1.pdf:47.3MBJAEA-Technology-2019-020-high-resolution2.pdf:34.99MBJAEA-Technology-2019-020-high-resolution3.pdf:48.74MBJAEA-Technology-2019-020-high-resolution4.pdf:47.83MBJAEA-Technology-2019-020-high-resolution5.pdf:18.35MBJAEA-Technology-2019-020-high-resolution6.pdf:49.4MBJAEA-Technology-2019-020-high-resolution7.pdf:39.78MB
The prototype fast breeder reactor Monju has produced valuable technological achievements through design, construction, operation and maintenance over half a century since 1968. This report compiles the reactor technologies developed for Monju, including the areas: history and major achievements, design and construction, commissioning, safety, reactor physics, fuel, systems and components, sodium technology, materials and structures, operation and maintenance, and accidents and failures.
Kato, Masato; McClellan, K.*
Transactions of the American Nuclear Society, 113(1), p.613 - 614, 2015/10
A joint study on advanced oxide fuels is being carried out under the Civil Nuclear Energy Working Group (CNWG) bilateral collaboration between the U.S. Department of Energy and the Japan Atomic Energy Agency. The main goal of this study is to support development and validation of a science-based fuel analysis code for minor actinide (MA) bearing MOX fuel. In analysis and evaluation of fuel performance, it is essential to understand the physical properties of the advanced oxide fuels. Therefore, we are investigating physical properties of (U,Pu)O
, (U,Ce,)O, PuO, CeO and other related compounds to prepare a physical property database and to construct an integrated mechanistic physical property model. In this paper, we describe the derivation of a model to represent heat capacity and thermal conductivity of (U,Pu)O
that is based on the experimental database.
Ono, Ayako; Tanaka, Masaaki; Kobayashi, Jun; Kamide, Hideki
Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 7 Pages, 2014/11
Yoshida, Hiroyuki; Tamai, Hidesada; Onuki, Akira; Takase, Kazuyuki; Akimoto, Hajime
Nuclear Engineering and Technology, 38(2), p.119 - 128, 2006/04
The reduced-moderation water reactor core adopts a hexagonal tight-lattice arrangement. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of the core using an advanced numerical simulation technology. As a part of this technology development, we are developing a two-phase flow simulation code TPFIT with an advanced interface tracking method. The vector and parallelization of the code was conducted to fit the large-scale simulations. The numerical results applied to large-scale water-vapor two-phase flow in tight lattice rod bundles are shown and compared with experimental results. In the results, a tendency of the predicted void fraction distribution in horizontal plane agreed with the measured values including the bridge formation of the liquid at the position of adjacent fuel rods where an interval is the narrowest.
Ide, Shunsuke; JT-60 Team
Plasma Science and Technology, 8(1), p.1 - 4, 2006/01
Times Cited Count：0 Percentile：0.01(Physics, Fluids & Plasmas)Recent progress in development of high performance plasma and efforts to prolong their sustainment towards ITER advanced operations and a steady-state reactor in JT-60U are presented focusing following achievements; 
N=3 sustained for 6.2s (
4.1tR) without NTMs in normal shear, fBS0.45 sustained for 5.8s (2.8tR) under nearly full CD in weak a shear plasma, fBS0.75 sustained for 7.4s (2.7tR) under nearly full CD in a reversed shear plasma. Furthermore, importance of these results and issues in advanced tokamak development will be discussed.
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Akimoto, Hajime
Computational Fluid Dynamics 2004, p.649 - 654, 2006/00
no abstracts in English
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Tamai, Hidesada
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 14 Pages, 2004/10
no abstracts in English
Yoshida, Hiroyuki; Nagayoshi, Takuji*; Ose, Yasuo*; Takase, Kazuyuki; Akimoto, Hajime
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(3), p.233 - 241, 2004/09
When there are no experimental data such as the reduced-moderation water reactor (RMWR), therefore, it is very difficult to obtain highly precise predictions. The RMWR core adopts a hexagonal tight lattice arrangement with about 1 mm gap between adjacent fuel rods. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of RMWR core using advanced numerical simulation technology. As part of this technology development, we are developing advanced interface tracking method to improve conservation of volume of fluid. In this paper, we describe a newly developed interface tracking method and examples of the numerical results. In the present results, the error of volume conservation in the bubbly flow is within 0.6%.
Department of Nuclear Energy System
JAERI-Review 2003-004, 236 Pages, 2003/03
JAERI-Review-2003-004.pdf:16.34MB
This report summarizes the research and development activities in the Department of Nuclear Energy System during the fiscal year of 2001 (April 1, 2001 - March 31, 2002). The Department has been organized from April 1998. The main research activity is aimed to build the basis of the development of future nuclear energy systems. The research activities of the Department cover basic nuclear data evaluation, conceptual design of a reduced-moderation water reactor, reactor physics experiments and development of the reactor analysis codes, experiment and analysis of thermal-hydrodynamics, energy system analysis and assessment, development of advanced materials for a reactor, lifetime reliability assessment on structural material, development of advanced nuclear fuel, design of a marine reactor and the research for a nuclear ship system. The maintenance and operation of reactor engineering facilities belonging to the Department are undertaken. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report.
Department of Nuclear Energy System
JAERI-Review 2002-005, 280 Pages, 2002/03
JAERI-Review-2002-005.pdf:18.05MB
no abstracts in English
Kiuchi, Kiyoshi; Ioka, Ikuo; Tachibana, Katsumi; Suzuki, Tomio; Fukaya, Kiyoshi*; Inohara, Yasuto*; Kambara, Shozo; Kuroda, Yuji*; Miyamoto, Satoshi*; Ogura, Kazutomo*
JAERI-Research 2002-008, 63 Pages, 2002/03
JAERI-Research-2002-008.pdf:7.85MB
no abstracts in English
Sakurai, Fumio; Horiguchi, Yoji; Kobayashi, Shinsho; Takayanagi, Masaji
Physica B; Condensed Matter, 311(1-2), p.7 - 13, 2002/01
Times Cited Count：9 Percentile：45.19(Physics, Condensed Matter)no abstracts in English
Okubo, Tsutomu; Takeda, Renzo*; Iwamura, Takamichi; Yamamoto, Kazuhiko*; Okada, Hiroyuki*
Proceedings of International Conference on Back-End of the Fuel Cycle: From Research to Solutions (GLOBAL 2001) (CD-ROM), 7 Pages, 2001/09
An advanced water-cooled reactor concept named the Reduced-Moderation Water Reactor (RMWR) has been proposed to attain a high conversion ratio more than 1.0 and to achieve the negative void reactivity coefficient. At present, several types of design concepts satisfying both the design targets have been proposed based on the evaluation for the fuel without fission products and minor actinides. In this paper, the feasibility of the RMWR core is investigated and confirmed for the plutonium multiple recycling under advanced reprocessing schemes with low decontamination factors as proposed for the FBR fuel cycle.
Ura, Tamaki*; Takamasa, Tomoji*; Nishimura, Hajime*; Aoki, Taro*; Ueno, Michio*; Maeda, Toshio*; Nakamura, Masato*; Shimazu, Shunsuke*; Tokunaga, Sango*; Shibata, Yozo*; et al.
JAERI-Tech 2001-049, 154 Pages, 2001/07
JAERI-Tech-2001-049.pdf:11.24MB
JAERI has studied on design and operation of a nuclear powered submersible research vessel, which will navigate under sea in the Arctic Ocean, as a part of the design study of advanced marine reactors. This report describes operation conditions and an operating system of the vessel those were discussed by the specialists of hull design, sound positioning, ship motions and oceanography, etc. The design conditions on ship motions for submersible vessels were surveyed considering regulations in our country, and ship motions were evaluated assuming the observation activities in the Arctic Ocean. A submarine transponder system and an on ice communication buoy system were examined as a positioning and communication system supposing the activity under ice. Procedures to secure safety of nuclear powered submersible research vessel were discussed based on the investigation of accidents. These results were reflected to the concept of the nuclear powered submersible research vessel, and subjects fto be settled in the next step were clarified.
Department of Nuclear Energy System
JAERI-Review 2001-010, 322 Pages, 2001/03
JAERI-Review-2001-010.pdf:27.21MB
no abstracts in English
Department of Nuclear Energy System
JAERI-Review 99-031, p.320 - 0, 2000/01
JAERI-Review-99-031.pdf:16.21MB
no abstracts in English
Iwamura, Takamichi; Okubo, Tsutomu; Usui, Shuji*; Shimada, Shoichiro*; Nabeshima, Kunihiko; *; Kakuta, Tsunemi; Ishikawa, Nobuyuki; Suzudo, Tomoaki; Nakatsuka, Toru; et al.
JAERI-Review 99-017, 60 Pages, 1999/08
JAERI-Review-99-017.pdf:2.94MB
no abstracts in English
JAERI-Review 98-022, 265 Pages, 1998/11
JAERI-Review-98-022.pdf:12.29MB
no abstracts in English
JAERI-Review 97-011, 338 Pages, 1997/10
JAERI-Review-97-011.pdf:11.71MB
no abstracts in English
