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Sun, G.*; Zhan, Y.*; Okawa, Tomio*; Aoyagi, Mitsuhiro; Uchibori, Akihiro; Okano, Yasushi
Experimental Thermal and Fluid Science, 151, p.111095_1 - 111095_15, 2024/02
Times Cited Count：1 Percentile：0.01(Thermodynamics)Uchibori, Akihiro; Doda, Norihiro; Aoyagi, Mitsuhiro; Sonehara, Masateru; Sogabe, Joji; Okano, Yasushi; Takata, Takashi*; Tanaka, Masaaki; Enuma, Yasuhiro; Wakai, Takashi; et al.
Nuclear Engineering and Design, 413, p.112492_1 - 112492_10, 2023/11
Times Cited Count：1 Percentile：72.91(Nuclear Science & Technology)The ARKAIDA has been developed to realize automatic optimization of plant design from safety evaluation for the advanced reactors represented by a sodium-cooled fast reactor. ARKADIA-Design offers functions to support design optimization both in normal operating conditions and design basis events. The multi-level simulation approach by the coupled analysis such as neutronics, core deformation, core thermal hydraulics was developed as one of the main technologies. On the other hand, ARKAIDA-Safety aims for safety evaluation considering severe accidents. As a key technology, the numerical methods for in- and ex-vessel coupled phenomena during severe accidents in sodium-cooled fast reactors were tested through a hypothetical severe accident event. Improvement of the ex-vessel model and development of the AI technology to find best design solution have been started.
Ishida, Shinya; Tagami, Hirotaka; Tobita, Yoshiharu; Okano, Yasushi; Yamano, Hidemasa; Kubo, Shigenobu
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
no abstracts in English
Tanaka, Masaaki; Uchibori, Akihiro; Okano, Yasushi; Yokoyama, Kenji; Uwaba, Tomoyuki; Enuma, Yasuhiro; Wakai, Takashi; Asayama, Tai
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
The book, JSME Series in Thermal and Nuclear Power Generation Vol.3 Sodium-cooled Fast Reactor, was published as a 30th anniversary memorial project of Power & Energy Systems Division. This paper describes an introduction of the book on a part of key technologies regarding safety assessment, thermal-hydraulics, neutronics, and fuel and material development. This introductory paper also provides an overview of an integrated evaluation system named ARKADIA to offer the best possible solutions for challenges arising during the design process, safety assessment, and operation of a nuclear plant over its life cycle, in active use of the R&D efforts and knowledges on thermal-hydraulics and safety assessment with state-of-the-art numerical analysis technologies.
Kosaka, Wataru; Uchibori, Akihiro; Okano, Yasushi; Yanagisawa, Hideki*
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.1150 - 1163, 2023/08
The leakage of pressurized water from a steam generator (SG) and the progress after that are a key issue in the safety assessment or design of a SG in sodium-cooled fast reactor. The analysis code LEAP-III can evaluate a rate of water leakage during the long-term event progress, i.e., from the self-wastage initiated by an occurrence of a microscopic crack in a tube wall to the water leak detection and water/water-vapor blowdown. Since LEAP-III consists of semi-empirical formulae and one-dimensional equations of conservation, it has an advantage in short computation time. Thus, LEAP-III can facilitate the exploration of various new SG designs in the development of innovative reactors. However, there are several problems, such as an excessive conservative result in some case and the need for numerous experiments or preliminary analyses to determine tuning parameters of models in LEAP-III. Hence, we have developed a Lagrangian particle method code, which is characterized by a simpler computational principle and faster calculation. In this study, we have improved the existing particle pair search method for interparticle interaction in this code and developed an alternative model without the pair search. Through the trial analysis simulating in a tube bundle system, it was confirmed that new models reduced the computation time. In addition, it was shown that representative temperatures of the heat-transfer tubes evaluated by this particle method code, which is used to predict the tube failure in LEAP-III, were good agreement with that by SERAPHIM, which is a detailed mechanistic analysis method code.
Li, C.-Y.; Wang, K.*; Uchibori, Akihiro; Okano, Yasushi; Pellegrini, M.*; Erkan, N.*; Takata, Takashi*; Okamoto, Koji*
Applied Sciences (Internet), 13(13), p.7705_1 - 7705_29, 2023/07
Times Cited Count：1 Percentile：0(Chemistry, Multidisciplinary)Tanaka, Masaaki; Enuma, Yasuhiro; Okano, Yasushi; Uchibori, Akihiro; Yokoyama, Kenji; Seki, Akiyuki; Wakai, Takashi; Asayama, Tai
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 11 Pages, 2023/05
Aoyagi, Mitsuhiro; Makino, Toru*; Oki, Hiroshi*; Uchibori, Akihiro; Okano, Yasushi
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 7 Pages, 2023/05
Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
Sun, G.*; Okawa, Tomio*; Aoyagi, Mitsuhiro; Uchibori, Akihiro; Okano, Yasushi
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 8 Pages, 2023/05
When liquid sodium leaks as a liquid jet from piping in a sodium-cooled fast reactor, the jet impinges with structures to produce splashing droplets which can cause significant combustion. According to previous studies on circular nozzles, the amount of splash is affected by the state of the jet at the moment of impingement. In the present work, a nozzle with a sudden contraction in the cross-sectional area was designed to reproduce a supposed pipe leakage, and the breakup behavior of jet discharged from this nozzle was observed. The result shows that the breakup of jet was accelerated until the jet transformed into a particularly stable state when the jet velocity exceeded a certain value. Once the jet has transformed, it will not turn back unless turning down the flow rate to a very low value. The stable jet after the transformation has a longer breakup length than that before the transformation in the same flow rate.
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
Ono, Masahiro*; Uchibori, Akihiro; Okano, Yasushi; Takata, Takashi*
JAEA-Testing 2022-004, 193 Pages, 2023/03
JAEA-Testing-2022-004.pdf:3.31MB
A computer code TRACER (Transport phenomena of Radionuclides for Accident Consequence Evaluation of Reactor) version 2.4.1 has been developed to evaluate species and quantities of fission products (FPs) released into cover gas due to a fuel pin failure in an LMFBR. The TRACER version 2.4.1 includes the models related to NUREG-0772 and also new or modified computational program codes in order to possess a new function shown below, and partial modify of coefficient of FP transition model between coolant and cover gas. This manual includes manual conventions for TRACER Version 2.3, addition of reference such as formula, improvement of explanation of input file creation method, addition of improvement of NUREG-0772 model added to TRACER code, modification of figure of sample analysis performed in appendix. It includes modifications and additions of sample analysis.
Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi
Journal of Nuclear Science and Technology, 13 Pages, 2023/00
Times Cited Count：1 Percentile：0.01(Nuclear Science & Technology)Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi
Journal of Nuclear Science and Technology, 23 Pages, 2023/00
Times Cited Count：1 Percentile：72.91(Nuclear Science & Technology)Aoyagi, Mitsuhiro; Sonehara, Masateru; Ishida, Shinya; Uchibori, Akihiro; Kawada, Kenichi; Okano, Yasushi; Takata, Takashi
Proceedings of Technical Meeting on State-of-the-art Thermal Hydraulics of Fast Reactors (Internet), 3 Pages, 2022/09
Uchibori, Akihiro; Sogabe, Joji; Okano, Yasushi; Takata, Takashi*; Doda, Norihiro; Tanaka, Masaaki; Enuma, Yasuhiro; Wakai, Takashi; Asayama, Tai; Ohshima, Hiroyuki
Proceedings of Technical Meeting on State-of-the-art Thermal Hydraulics of Fast Reactors (Internet), 10 Pages, 2022/09
The ARKAIDA has been developed to realize automatic optimization of plant design from safety evaluation for the advanced reactors represented by a sodium-cooled fast reactor. ARKADIA-Design offers functions to support design optimization both in normal operating conditions and design basis events. The multi-level simulation approach by the coupled analysis such as neutronics, core deformation, core thermal hydraulics was developed as one of the main technologies of the ARKADIA-Design. On the other hand, ARKAIDA-Safety aims for safety evaluation considering severe accidents. As a key technology, the numerical methods for in- and ex-vessel coupled phenomena during severe accidents in sodium-cooled fast reactors were tested through a hypothetical severe accident event.
Sun, G.*; Zhan, Y.*; Okawa, Tomio*; Aoyagi, Mitsuhiro; Uchibori, Akihiro; Okano, Yasushi
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 6 Pages, 2022/08
When a liquid sodium leakage accident occurs in a sodium-cooled fast reactor, the injected sodium collides with structures to produce splashing droplets, which can result in a violent combustion. According to previous studies on circular nozzles, the amount of splash is affected by the state of the jet at the moment of impact. However, the outlet shape of damaged area is hardly to be circular; and meanwhile it influences the flow pattern of jet a lot. Considering about this, in the present work, high-speed cameras were used to observe the jet discharged from oval nozzles vertically downward to investigate the falling process of the jet. The result shows that surface wave appears on the jet and within a certain range of flow velocity it can be observed obviously, meanwhile accelerate the breakup of jet.
Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2022/07
Identifying accident scenarios that could lead to severe accidents and evaluating their frequency of occurrence are essential issues. This study aims to establish the methodology of the dynamic Probabilistic Risk Assessment (PRA) for sodium-cooled fast reactors that can consider the time dependency and the interdependence of each event. Specifically, the Continuous Markov chain Monte Carlo (CMMC) method is newly applied to the SPECTRA code, which analyzes the severe accident conditions of nuclear reactors, to develop an evaluation methodology for typical external hazards. Currently, a fault-tree model of air coolers of decay heat removal system is implemented as the CMMC method, and a series of preliminary analysis of the plant's transient characteristics under the scenario of volcanic ashfall has been conducted.
Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.
Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07
This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.
Okano, Yasushi; Ammirabile, L.*; Sofu, T.*
2018 GIF Symposium Proceedings (Internet), p.253 - 262, 2020/05
GIF ISAM (Integrated Safety Assessment Methodology) includes five analytical tools (i.e. QSR, PIRT, OPT, DPA, PSA) and it is intended that each tool be used to answer specific safety-related questions with different levels of detail during various design stages and the ISAM as a whole offers flexibility and a graded approach to analyse technical issues of complex system architectures. Although each tool can be selected for individual and exclusive use, the full value of the integrated methodology is derived from using all tools, in an iterative fashion and in combination with the others, throughout the design process. The paper describes what is ISAM and pilot examples of individual use of QSR, PIRT and OPT and also combination application of DPA-PSA.
