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Ota, Hirokazu*; Ogata, Takanari*; Yamano, Hidemasa; Futagami, Satoshi; Shimada, Sadae*; Yamada, Yumi*
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 8 Pages, 2023/05
Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi
Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 8 Pages, 2023/04
Madokoro, Hiroshi; Sato, Ikken
Nuclear Engineering and Design, 376, p.111123_1 - 111123_15, 2021/05
Times Cited Count：6 Percentile：72.21(Nuclear Science & Technology)Ishida, Shinya; Kawada, Kenichi; Fukano, Yoshitaka
Mechanical Engineering Journal (Internet), 7(3), p.19-00523_1 - 19-00523_17, 2020/06
The Phenomena Identification and Ranking Table (PIRT) approach was applied to the validation of SAS4A code in order to indicate the reliability of SAS4A code sufficiently and objectively. Based on this approach, issue and objective were clarified, plant design and scenario were defined, FOM and key phenomena were selected, and the code validation test matrix was completed with the results of investigation about analysis models and test cases. The results of the test analysis corresponding to this matrix show that the SAS4A models required for the IP evaluation were sufficiently validated. Furthermore, the validation with this matrix is highly reliable, since this matrix represents the comprehensive validation that also considers the relation between physical phenomena. In this study, the reliability and validity of SAS4A code were significantly enhanced by using PIRT approach to the sufficient level for CDA analyses in SFR.
Ohshima, Hiroyuki; Kamide, Hideki
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.2095 - 2107, 2019/08
Development of a sodium-cooled fast reactor has been implemented in Japan from the viewpoint of severe accident countermeasures. This paper describes the progress of research and development related to safety enhancement and the severe accident countermeasures. A volcanic PRA methodology was developed for the proper consideration of external hazards. Water and sodium experiments were carried out for the decay heat removal in a core disruptive accident (CDA), and also thermal hydraulic interactions between the core and upper plenum where dipped heat exchanger was operated. In order to elucidate the behavior of molten fuel during CDA, basic experiments of core melt fragmentation in deep and shallow sodium pools were carried out. X-ray visualization showed the liquid column of molten steel was intensively fragmented nearly simultaneously with a rapid expansion of sodium vapor.
Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji; Ganovichev, D. A.*; Baklanov, V. V.*
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 9 Pages, 2019/05
In order to ensure In-Vessel Retention (IVR) of molten-core in Core Disruptive Accident (CDA), we are investigating the possibility of the molten-core discharge through the control rod guide tube (CRGT) to prevent energetics due to exceeding the prompt criticality. Internal structures of the CRGT, such as a sodium-flow regulator when the CRGT is connected to the high-pressure plenum, may disturb the discharge of molten-core from the core region. Based on above background, an experimental program to clarify characteristics of molten-core discharge through the CRGT has been commenced as one of subjects under a joint study with National Nuclear Center of the Republic of Kazakhstan (NNC-RK) named EAGLE-3 project. An experiment using molten-alumina as fuel simulant and sodium was conducted at the out-of-pile test facility owned by NNC-RK to investigate sodium cooling effect around the sodium flow regulator on its destruction. The experimental result represented that void development at the initiation of molten-alumina discharge eliminated liquid-phase sodium from the discharge path and this also eliminated sodium cooling effect around the sodium flow regulator. As a result, early destruction of the sodium flow regulator and massive discharge of molten alumina occurred in turn.
Ishida, Shinya; Kawada, Kenichi; Fukano, Yoshitaka
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 10 Pages, 2019/05
Core Disruptive Accident (CDA) has been considered as one of the important safety issues in the severe accident evaluation of Sodium-cooled Fast Reactor (SFR), and SAS4A code is developed for Initiating Phase (IP) of CDA. Phenomena Identification and Ranking Table (PIRT) approach was applied to the validation of SAS4A code in order to enhance its reliability in this study. SAS4A was validated in the following steps: (1) selection of the figure of merit (FOM) corresponding to Unprotected Loss Of Flow (ULOF) which is one of the most important and typical events in CDA, (2) identification of the phenomena involved in ULOF, (3) ranking the important phenomena, (4) development of the code validation test matrix, and (5) test analyses for validation corresponding to the test matrix. The reliability and validity of SAS4A code were significantly enhanced by this validation with PIRT approach.
Matsuo, Eiji*; Sasa, Kyohei*; Koyama, Kazuya*; Yamano, Hidemasa; Kubo, Shigenobu; Hourcade, E.*; Bertrand, F.*; Marie, N.*; Bachrata, A.*; Dirat, J. F.*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 5 Pages, 2019/05
Discharged molten-fuel from the core during Core Disruptive Accident (CDA) could become solidified particle debris by fuel-coolant interaction in the lower sodium plenum, and then the debris could form a bed on a core catcher located at the bottom of the reactor vessel. Coolability evaluations for the debris bed are necessary for the design of the core catcher. The purpose of this study is to evaluate the coolability of the debris bed on the core catcher for the ASTRID design. For this purpose, as a first step, the coolability calculations of the debris beds formed both in short term and later phase have been performed by modeling only the debris bed itself. Thus, details of core catcher design and decay heat removal system are not described in this paper. In all the calculations, coolant temperature around the debris bed is a parameter. The calculation tool is the debris bed module implemented into a one-dimensional plant dynamics code, Super-COPD. The evaluations have shown that the debris beds formed both in short term and later phase are coolable by the design which secures sufficient coolant flow around the core catcher located in the cold pool.
Ohgama, Kazuya; Nakano, Yoshihiro; Oki, Shigeo
Journal of Nuclear Science and Technology, 53(8), p.1155 - 1163, 2016/08
Times Cited Count：1 Percentile：10.71(Nuclear Science & Technology)The power distribution and core characteristics in various configurations of fuel subassemblies with an innerduct structure in the Japan Sodium-cooled Fast Reactor were evaluated using a Monte Carlo code for neutron transport and burnup calculation. The correlation between the fraction of fuel subassemblies facing outward and the degree of power increase at the core center was observed regardless of the compositions. This indicated that the spatial fissile distribution caused by innerduct configurations was the major factor of the difference in the power distribution. A power increase was also found in an off-center region, and it tended to be greater than that at the core center because of the steep gradient of neutron flux intensity. The differences in the worth of control rods caused by innerduct configurations were confirmed.
Suzuki, Toru; Tobita, Yoshiharu; Kawada, Kenichi; Tagami, Hirotaka; Sogabe, Joji; Matsuba, Kenichi; Ito, Kei; Ohshima, Hiroyuki
Nuclear Engineering and Technology, 47(3), p.240 - 252, 2015/04
Times Cited Count：27 Percentile：91.4(Nuclear Science & Technology)Kamiyama, Kenji; Konishi, Kensuke; Sato, Ikken; Toyooka, Junichi; Matsuba, Kenichi; Suzuki, Toru; Tobita, Yoshiharu; Pakhnits, A. V.*; Vityuk, V. A.*; Vurim, A. D.*; et al.
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 8 Pages, 2014/12
Hidaka, Akihide; Asaka, Hideaki; Ueno, Shingo*; Yoshino, T.*; Sugimoto, Jun
JAERI-Research 99-067, p.55 - 0, 1999/12
JAERI-Research-99-067.pdf:2.51MB
no abstracts in English
Hidaka, Akihide; Maruyama, Yu; Ueno, Shingo*; Sugimoto, Jun
JAERI-Conf 99-005, p.49 - 55, 1999/07
no abstracts in English
Katsuta, Hiroji; Noda, Kenji; ; Sugimoto, Masayoshi; Maekawa, Hiroshi; Konishi, Satoshi; Nakamura, Hideo; *; Oyama, Yukio; Jitsukawa, Shiro; et al.
Nihon Genshiryoku Gakkai-Shi, 40(3), p.162 - 191, 1998/00
no abstracts in English
; Nakamura, Hideo; *; Maekawa, Hiroshi; Katsuta, Hiroji; T.Hua*; S.Cevolani*
Eighth Int. Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), 3, p.1260 - 1267, 1997/00
no abstracts in English
JAERI-Conf 96-012, 394 Pages, 1996/08
no abstracts in English
Hidaka, Akihide; Ezzidi, A.*; Sugimoto, Jun
PSA 96: Int. Topical Meeting on Probabilistic Safety Assessment, 3, p.1548 - 1556, 1996/00
no abstracts in English
JAERI-Conf 95-019, 257 Pages, 1995/09
no abstracts in English
Hidaka, Akihide; Soda, Kunihisa; Sugimoto, Jun
Journal of Nuclear Science and Technology, 32(6), p.527 - 538, 1995/06
Times Cited Count：3 Percentile：36.75(Nuclear Science & Technology)no abstracts in English
