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Onoda, Yuichi; Tobita, Yoshiharu; Okano, Yasushi
IAEA-TECDOC-2079, p.215 - 225, 2025/00
The analysis methodologies for the evaluation of unprotected loss of flow accident on sodium-cooled fast reactor in Japan Atomic Energy Agency (JAEA) are briefly explained focusing on the mechanical consequences during expansion phase of the accident. JAEA developed the analysis methodologies for the evaluation of energetics and divided the analysis process into following three: 1) analysis of converting the heat generated into the mechanical energy with SIMMER code, 2) analysis of the structural response of the reactor vessel with AUTODYN code, and 3) analysis of the amount of sodium ejected onto the top shield through the gaps between shield plugs with PLUG code. Pressure-volume relation of the CDA bubble, which is the mixture of gas (fuel, steel vapor and fission gas) and molten core material, obtained by SIMMER calculation is used as the input for structural response analysis with AUTODYN. Pressure history exerted on the lower surface of the top shield obtained by SIMMER calculation is used as the input for PLUG. These analysis codes are validated by simulating the dominant phenomena that significantly affect the results in each calculation. We applied these analysis methodologies developed by JAEA to the reactor case analyses and confirmed their applicability.
Tobita, Yoshiharu; Tagami, Hirotaka; Ishida, Shinya; Onoda, Yuichi; Sogabe, Joji; Okano, Yasushi
IAEA-TECDOC-2079, p.72 - 84, 2025/00
Since the fast reactor core is not in the maximum reactivity configuration, a hypothetical core disruptive accident could lead to the prompt criticality due to a change in the core material configuration, and the resulting energy generation has been one of the issues in fast reactor safety, and therefore appropriate measures are needed to mitigate and contain the effect of energy generated in the accident. In order to assess the effectiveness of these mitigative measures, a set of computer codes to analyze the event progressions and energy generation behavior in the ATWS of fast reactors have been developed, maintained, and improved under international collaboration in JAEA. Since the important physical phenomena, which govern the event progression, vary along with the progression of the accident, the whole accident process is divided into several phases in the analysis of accident, and dedicated analysis methods for each phase are provided to analyze the event progression in each phase. The organization and overview of these analysis methods are described in this paper. As a representative example of the validation approaches in applying these analysis methods to the reactor safety assessment in licensing procedure in Japan, the validation studies to confirm the applicability to reactor analysis of the SIMMER code for analyzing core material movement and reactor power, which is important to analyze the energy generation in the accident, are presented in the paper. The validation studies of the SIMMER code confirmed the applicability of SIMMER to the reactor analysis, while the critical phenomena that the effect of their uncertainty in the reactor analysis should be checked were also recognized.
Sonehara, Masateru; Okano, Yasushi; Uchibori, Akihiro; Oki, Hiroshi*
Journal of Nuclear Science and Technology, 12 Pages, 2024/12
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)For sodium-cooled fast reactors, understanding sodium combustion behaviour is crucial for managing sodium leakage accidents. In this study, we perform benchmark analyses of the Sandia National Laboratories (SNL) T3 experiment using the multi-dimensional thermal hydraulic code AQUA-SF. Conducted in an enclosed space with a large vessel volume of 100 m
and a sodium mass flow rate of 1 kg/s, the experiment highlighted the multi-dimensional effects of local temperature increase shortly after sodium injection. This study aims to extend the capabilities of AQUA-SF by focusing on the simulation of these multi-dimensional temperature variations, in particular the formation of high temperature regions at the bottom of the vessel. The proposed models include the temporary stopping of sodium droplet ignition and spray combustion of sodium splash on the floor. Furthermore, it has been shown that additional heat source near the floor is essential to enhance the reproduction of the high temperature region at the bottom. Therefore, case studies including sensitivity analyses of spray cone angle and prolonged combustion of droplets on the floor are conducted. This comprehensive approach provides valuable insights into the dynamics of sodium combustion and safety measures in sodium-cooled fast reactors.
Sogabe, Joji; Ishida, Shinya; Tagami, Hirotaka; Okano, Yasushi; Kamiyama, Kenji; Onoda, Yuichi; Matsuba, Kenichi; Yamano, Hidemasa; Kubo, Shigenobu; Kubota, Ryuzaburo*; et al.
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
In the frame of France-Japan collaboration, the calculational methodologies were defined and assessed, and the phenomenology and the severe accident consequences were investigated in a pool-type sodium-cooled fast reactor.
Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi
Journal of Nuclear Science and Technology, 61(7), p.935 - 957, 2024/07
Times Cited Count：2 Percentile：46.61(Nuclear Science & Technology)Ishida, Shinya; Uchibori, Akihiro; Okano, Yasushi
Dai-28-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2024/06
no abstracts in English
Uchibori, Akihiro; Okano, Yasushi
Isotope News, (793), p.32 - 35, 2024/06
The design of a containment vessel in a sodium-cooled fast reactor was optimized from simulation on the hypothetical severe accident including sodium leakage and combustion. The simulation method is one of the base technologies of the design optimization system, ARKADIA. The simulation was performed on the different design conditions including volume of the containment vessel and the safety equipment as optimization parameters. The iterative simulation successfully found that the safety under this accident was kept even in the downsized containment vessel by selecting an effective safety equipment. This study demonstrated that the developed method has basic capability for design optimization in ARKADIA.
Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi
Journal of Nuclear Science and Technology, 61(5), p.582 - 594, 2024/05
Times Cited Count：1 Percentile：25.62(Nuclear Science & Technology)Tagami, Hirotaka; Ishida, Shinya; Okano, Yasushi; Yamano, Hidemasa; Kubo, Shigenobu; Payot, F.*; Saas, L.*; Trotignon, L.*; Gubernatis, P.*; Dufour, E.*; et al.
Proceedings of 11th European Review Meeting on Severe Accident Research Conference (ERMSAR 2024) (Internet), 12 Pages, 2024/05
JAEA has been developing the SIMMER-V code in collaboration with CEA to perform severe accident (SA) simulations of future sodium-cooled fast reactors (SFRs) including a unique core design with large-scale heterogeneous cores. An SA sequence in SFRs has been analyzed by: the SAS4A code for the Initiation Phase (IP), in which fuel pin disruption and vertical fuel dispersion occur in individual fuel subassemblies; and the two-dimensional SIMMER-III or three-dimensional SIMMER-IV code for the Transition Phase (TP), in which core disruption extends to the whole core. The joint development of SIMMER-V is of limited scope but aims at significantly expanding the code applicability by providing flexible interfaces to couple a SIMMER-V calculation with other computational domains or other codes, and by adding new advanced physical models such as a detailed fuel pin model and a model of flexible treatment of fuel isotopic composition. The former tasks are conducted by CEA the latter tasks by JAEA. In parallel to the code development, verification and validation of the new models and methods have been performed. This paper describes the objectives and overall framework of SIMMER-V code development program, representative new elements, and recent development progress.
Ishida, Shinya; Tagami, Hirotaka; Okano, Yasushi; Yamano, Hidemasa; Kubo, Shigenobu; Tobita, Yoshiharu
Proceedings of 11th European Review Meeting on Severe Accident Research Conference (ERMSAR 2024) (Internet), 10 Pages, 2024/05
Aoyagi, Mitsuhiro; Makino, Toru*; Oki, Hiroshi*; Uchibori, Akihiro; Okano, Yasushi
Mechanical Engineering Journal (Internet), 11(2), p.23-00459_1 - 23-00459_12, 2024/04
Tanaka, Masaaki; Enuma, Yasuhiro; Okano, Yasushi; Uchibori, Akihiro; Yokoyama, Kenji; Seki, Akiyuki; Wakai, Takashi; Asayama, Tai
Mechanical Engineering Journal (Internet), 11(2), p.23-00424_1 - 23-00424_13, 2024/04
The outline and development status of element functions and design optimization process in ARKADIA to transform advanced nuclear reactor design to meet expectations of a safe, economic, and sustainable carbon-free energy source are introduced. It is also briefly explained that ARKADIA will realize Artificial Intelligence (AI)-aided integrated numerical analysis to offer the best possible solutions for the design and operation of a nuclear plant including optimization of safety equipment, and merge state-of-the-art numerical simulation technologies and a knowledge base that stores data and insights from past nuclear reactor development projects and R&Ds with AI technologies.
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：7 Percentile：76.96(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：2 Percentile：46.61(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：2 Percentile：37.97(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
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.
