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Yamamoto, Tomohiko; Iwasaki, Akihisa*; Kawamura, Kazuki*; Matsubara, Shinichiro*; Harada, Hidenori*
Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 8 Pages, 2018/07
To design fast reactor (FR) core components, seismic response must be evaluated in order to ensure structural integrity. Thus, a core seismic analysis method has been developed to evaluate 3D core vibration behavior considering fluid structure interaction and vertical displacements (rising). 1/1.5 scale 37 core element mock-ups hexagonal-matrix experiment was performed to validate the core elements vibration analysis code in three dimensions (REVIAN-3D). Based on the test data, the analysis model newly incorporated to respond to strong excitation was verified.
Yamamoto, Tomohiko; Kitamura, Seiji; Iwasaki, Akihisa*; Matsubara, Shinichiro*; Okamura, Shigeki*
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 10 Pages, 2017/07
To design fast reactor (FR) components, seismic response must be evaluated in order to ensure structural integrity. Therefore, a sophisticated analysis method has to be developed to study the seismic response of FR core. The fast reactors are made of several hundred core assemblies in hexagonal arrangement. When a big earthquake occurs, large horizontal displacement and impact force of each core assembly may cause a trouble for control rod insertability and core assembly intensity. Therefore, a seismic analysis method of fast reactor core considering horizontal nonlinear behavior, such as impact, fluid-structure interaction, etc. is needed. Validation of the core assembly vibration analysis code in three dimension (REVIAN-3D) was conducted by a full scale experiment. In this validation, the vertical behavior (raising displacement) and horizontal behavior (Impact force, horizontal response) of the analysis result agreed very well with the experiments.
Chikazawa, Yoshitaka; Kato, Atsushi*; Yamamoto, Tomohiko; Kubo, Shigenobu; Ohno, Shuji; Iwasaki, Mikinori*; Hara, Hiroyuki*; Shimakawa, Yoshio*; Sakaba, Hiroshi*
Nuclear Technology, 196(1), p.61 - 73, 2016/10
Times Cited Count：1 Percentile：10.71(Nuclear Science & Technology)JSFR adopts double boundary for all sodium components. However, design measures are investigated for the secondary sodium fire inside the reactor building, which might be assumed as design extension conditions (DECs). Candidates of sodium fire measures in the secondary sodium systems such as sodium drain, nitrogen injection, pressure release valve, catch pan and leak sodium drain system have been compared from the view point of safety. Wide range of sodium fires in the steam generator room and air cooler have been analyzed evaluating performances of the candidate sodium fire measures.
Chikazawa, Yoshitaka; Kato, Atsushi; Yamamoto, Tomohiko; Kubo, Shigenobu; Iwasaki, Mikinori*; Hara, Hiroyuki*; Shimakawa, Yoshio*; Sakaba, Hiroshi*
Proceedings of 2014 International Congress on the Advances in Nuclear Power Plants (ICAPP 2014) (CD-ROM), p.523 - 530, 2014/04
JSFR adopts double boundary for all sodium components. However, design measures are investigated for the secondary sodium fire inside the reactor building, which might be assumed as design extension conditions (DECs). Candidates of sodium fire measures in the secondary sodium systems such as sodium drain, nitrogen injection, pressure release valve, catch pan and leak sodium drain system have been compared from the view point of safety. Wide range of sodium fires in the steam generator room have been analyzed evaluating performances of the candidate sodium fire measures.
Yamamoto, Tomohiko; Kato, Atsushi; Chikazawa, Yoshitaka; Ito, Kei; Iwasaki, Mikinori*; Akiyama, Yo*; Oya, Takeaki*
Proceedings of 2013 International Congress on Advances in Nuclear Power Plants (ICAPP 2013) (USB Flash Drive), 9 Pages, 2013/04
The Japan sodium-cooled fast reactor (JSFR) is planning to adopt a steel-plate reinforced concrete (SC) structure reactor building and an advanced seismic isolation system for reactor building. In the response of Fukushima Dai-ichi Nuclear Power Plant (Fukushima I NPP) accident, the evaluation and countermeasure study of earthquake, and other external hazards on JSFR has been analyzed based on 2010 JSFR design. This paper describes the detail of evaluation and countermeasure of earthquake, tsunami and other external hazards to JSFR reactor building.
Kato, Atsushi; Chikazawa, Yoshitaka; Yamamoto, Tomohiko; Ohno, Shuji; Kubo, Shigenobu; Sakaba, Hiroshi*; Akiyama, Yo*; Iwasaki, Mikinori*
Proceedings of 2013 International Congress on Advances in Nuclear Power Plants (ICAPP 2013) (USB Flash Drive), 9 Pages, 2013/04
After the accident of TEPCO's Fukushima Dai-ichi Nuclear Power Plant, evaluations of severe events beyond the design basis on a NPP are focused. As one of those activities, wide range of sodium combustion and hydrogen generation potentials have been analyzed to investigate potential consequences on SCCV. Structural and boundary integrity of SCCV have been evaluated from sodium combustion analyses for pressure and temperature loads. Hydrogen generation has also been evaluated as potential loads of SCCV.
Yamamoto, Tomohiko; Kato, Atsushi; Chikazawa, Yoshitaka; Oya, Takeaki*; Iwasaki, Mikinori*; Hara, Hiroyuki*; Akiyama, Yo*
Proceedings of 2012 International Congress on Advances in Nuclear Power Plants (ICAPP '12) (CD-ROM), p.500 - 508, 2012/06
Japan Sodium-cooled Fast Reactor (JSFR) is planning to adopt the new concepts of reactor building. One is that the steel plate reinforced concrete is adopted for containment vessel and reactor building. The other is the advanced seismic isolation system. This paper describes the detail of new concepts for JSFR reactor building and engineering evaluation of the new concepts.
Hirano, Fumio; Sato, Seichi*; Kozaki, Tamotsu*; Inagaki, Yaohiro*; Iwasaki, Tomohiko*; Oe, Toshiaki*; Kato, Kazuyuki*; Kitayama, Kazumi*; Nagasaki, Shinya*; Niibori, Yuichi*
Journal of Nuclear Science and Technology, 49(3), p.310 - 319, 2012/03
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)The thermal impacts of hull and end piece wastes from the reprocessing of MOX spent fuels burned in LWRs on repository performance were investigated. The heat generation rates in MOX spent fuels and the resulting heat generation rates in hull and end piece wastes change depending on the fuel histories including the burn-up of UO
spent fuels, the cooling period before reprocessing, the storage period of fresh MOX fuels. The heat generation rates of hull and end piece wastes from the reprocessing of MOX spent fuels with any of those histories are significantly larger than those from UO spent fuels with burn-ups of 45 GWd/THM. If a temperature below 80
C is specified for cement-based materials used in waste packages after disposal, the allowable number of canisters containing compacted hull and end pieces in a package for 45 GWd-MOX needs to be limited to a value of 0.7 to 1.6, which is significantly lower than the value of 4.0 for 45 GWd-UO.
Kato, Atsushi; Negishi, Kazuo; Yamamoto, Tomohiko; Akiyama, Yo*; Hara, Hiroyuki*; Iwasaki, Mikinori*
Transactions of 21st International Conference on Structural Mechanics in Reactor Technology (SMiRT-21) (CD-ROM), 8 Pages, 2011/11
Japan Sodium-cooled Fast Reactor (JSFR) adopts a new concept of a containment vessel called steel plate reinforced concrete containment vessel (SCCV). The SCCV is considered to be effective to shorten construction period thanks for elimination of rebar work at a site compared with applying the reinforced concrete CV. Other than this advantage, the SCCV achieves high quality in building structure, since steel structure parts can be fabricated at a factory prior to the site construction. Although the SC structure has been used for the buildings of LWR etc, it is important to investigate its characteristics under high temperature to adopt the SC structure to the JSFR CV. This paper mainly describes the design study and experiments to investigate potential characteristics of the SC structure under hypothetical sodium combustion in the CV.
Nishihara, Kenji; Iwanaga, Kohei*; Tsujimoto, Kazufumi; Kurata, Yuji; Oigawa, Hiroyuki; Iwasaki, Tomohiko*
Journal of Nuclear Science and Technology, 45(8), p.812 - 822, 2008/08
Times Cited Count：47 Percentile：93.66(Nuclear Science & Technology)In the present neutronics design of the Accelerator-driven system (ADS) cooled by lead-bismuth eutectic (LBE), we investigated several methods to reduce the power peak and beam current, and estimated the temperature drops of the cladding tube and beam window. The methods are adjustment of inert matrix ratio in fuel in each burn-up cycle, multi-region design in terms of pin radius or inert matrix content, and modification of the level of the beam window position and the height of the central fuel assemblies. As the result, we optimized the ADS combined with the adjustment of inert matrix ratio in each burn-up cycle, multi-region design in terms of inert matrix content and deepened window level. The maximum temperatures of the optimized ADS at the surface of the cladding tube and the beam window were reduced by 91 and 38 C, respectively.
Nishihara, Kenji; Nakayama, Shinichi; Morita, Yasuji; Oigawa, Hiroyuki; Iwasaki, Tomohiko*
Journal of Nuclear Science and Technology, 45(1), p.84 - 97, 2008/01
Times Cited Count：40 Percentile：91.3(Nuclear Science & Technology)We studied how cooling in the predisposal storage period may affect the design of the emplacement area in a repository for radioactive wastes produced by a light-water-reactor nuclear system that uses partitioning and/or transmutation (PT) technology. Three different fuel cycle scenarios involving PT technology were analyzed: (1) partitioning process only (separation of some fission products), (2) transmutation process only (separation and transmutation of minor actinides), and (3) both partitioning and transmutation. The necessary predisposal storage periods for some predefined emplacement configurations were determined through transient thermal analysis. For each scenario, we also estimated the storage capacity required for dry storage. The contributions of PT technology on the storage and disposal were discussed holistically, and we noted that the coupled introduction of partitioning and transmutation processes can bring an appreciable reduction in waste management size.
Yoshikawa, Takamichi*; Iwasaki, Tomohiko*; Wada, Kotaro*; Suyama, Kenya
Proceedings of American Nuclear Society Topical Meeting on Physics of Reactors (PHYSOR 2006) (CD-ROM), 8 Pages, 2006/09
To examine the procedures of the reprocessing, the vitrification and the geologic disposal, precise burn-up calculation for high burn-up and MOX fuels has been performed for not only PWR but also BWR by using SWAT and SWAT2 codes which are the integrated burn-up calculation code systems combined with the burn-up calculation code, ORIGEN2, and the transport calculation code, SRAC (the collision probability method) or MVP (the continuous energy Monte Carlo method), respectively. The calculation results shows that all of the evaluated items (heat generation and concentrations of Mo and Pt) largely increase and those significantly effect to the current procedures of the vitrification and the geologic disposal. The calculation result by SWAT2 confirms that the bundle calculation is required for BWR to be discussed about those effects in details, especially for the MOX fuel.
Igashira, Masayuki*; Shibata, Keiichi; Takano, Hideki*; Yamano, Naoki*; Matsunobu, Hiroyuki*; Kitao, Kensuke*; Katakura, Junichi; Nakagawa, Tsuneo; Hasegawa, Akira; Iwasaki, Tomohiko*; et al.
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(1), p.128 - 139, 2004/03
no abstracts in English
Kiyanagi, Yoshiaki*; Nagamiya, Shoji*; Oyama, Yukio; Ikeda, Yujiro; Oigawa, Hiroyuki; Igashira, Masayuki*; Baba, Mamoru*; Iwasaki, Tomohiko*; Watanabe, Yukinobu*; Ishibashi, Kenji*
Nihon Genshiryoku Gakkai-Shi, 46(3), p.173 - 197, 2004/03
no abstracts in English
Nishihara, Kenji; Iwasaki, Tomohiko*; Udagawa, Yutaka*
Journal of Nuclear Science and Technology, 40(7), p.481 - 492, 2003/07
Times Cited Count：17 Percentile：72.74(Nuclear Science & Technology)A new one-point equation is derived according to the balance of the fission neutrons. The equation has the same form as the conventional equation containing keff. The variables of the equation are the number of the fission neutrons and the delayed neutron precursors, and the coefficients are the multiplication factors of a prompt fission neutron, a delayed neutron and a source neutron. In the equation derived here, all variables and coefficients have each clear physical meaning. Analytic, deterministic and probabilistic calculations of the equation are performed for an accelerator-driven system.
Nishitani, Takeo; Okada, Koichi; Sato, Satoshi; Sasao, Mamiko*; Iwasaki, Tomohiko*; Sugawara, Takanori; Shinto, Katsuhiro*; Kitajima, Sumio*; Nomura, Ken*
no journal, ,
no abstracts in English
Okada, Koichi; Kondo, Keitaro; Sato, Satoshi; Nishitani, Takeo; Nomura, Ken*; Okamoto, Atsushi*; Iwasaki, Tomohiko*; Kitajima, Sumio*; Sasao, Mamiko*
no journal, ,
no abstracts in English
Yoshikawa, Takamichi*; Iwasaki, Tomohiko*; Endo, Hideki*; Suyama, Kenya; Shikoda, Keiji*; Yamada, Kohei*; Hamahata, Yoshiki*; Oeda, Shin*
no journal, ,
In order to develop a fuel design system of BWR based on opened calculation codes, a burnup code system SWAT2 using continuous energy Mote Carlo code was revised to include a function of branch calculation. A tool to evaluate lattice constants used in calculation code adopting modern nodal method was also developed and validated by comparison with CASMO.
Sato, Ikken; Iwasaki, Tomohiko*; Konashi, Kenji*
no journal, ,
no abstracts in English
Sato, Ikken; Iwasaki, Tomohiko*; Konashi, Kenji*
no journal, ,
Concerning safety characteristics of FBR core with hydride control materials, representative accidents for design basis and those beyond the design basis were selected and transient responses were evaluated. Based on this result, it was concluded that no unfavorable response was expected for this option.
