Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Li, J.*; Jang, S.*
Mechanical Engineering Journal (Internet), 7(3), p.19-00548_1 - 19-00548_11, 2020/06
Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was constructed to expand application range of an existing computer code. Applicability of the method was constructed through the numerical analysis of the experiment on water vapor discharging in liquid sodium. To improve the evaluation accuracy for the temperature distribution, a Lagrangian particle model for simulating reacting jet was also developed as an alternative method and its basic function was confirmed.
Aoyagi, Mitsuhiro; Uchibori, Akihiro; Takata, Takashi; Ohshima, Hiroyuki
Transactions of the American Nuclear Society, 122(1), p.862 - 865, 2020/06
Development of a new simulation system SPECTRA has been started to enable a simulation of comprehensive in- and ex-vessel events during a severe accident in a sodium-cooled fast reactor. The simulation system SPECTRA consists of two basic modules of thermal-hydraulics; in-vessel basic module and ex-vessel one, and some sub-modules for specific phenomena or events. A sodium fire models are implemented as one sub-module of the ex-vessel module. The sodium fire models are adapted from existing sodium fire analysis codes AQUA-SF and SPHINCS. As the result of verification test, the results show good agreement with the one of original codes. The validation analysis for single droplet falling and combustion corresponds well with the experimental data reasonably.
Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Ohshima, Hiroyuki
Nippon Kikai Gakkai Rombunshu (Internet), 86(883), p.19-00353_1 - 19-00353_6, 2020/03
Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was constructed to expand application range of an existing computer code. Applicability of the method was constructed through the numerical analysis of the experiment on water vapor discharging in liquid sodium.
Aoyagi, Mitsuhiro; Uchibori, Akihiro; Kikuchi, Shin; Takata, Takashi; Ohno, Shuji; Ohshima, Hiroyuki
Nippon Kikai Gakkai Rombunshu (Internet), 86(883), p.19-00366_1 - 19-00366_8, 2020/03
Sodium fire is one of key issues in sodium-cooled fast reactor plant. JAEA has developed sodium fire analysis codes, such as AQUA-SF and SPHINCS, to evaluate the consequence of sodium fire events. This paper describes the PIRT (Phenomena Identification and Ranking Table) process for sodium fire events. Ranking table for important phenomena and an assessment matrix are completed. As a part of comprehensive validation based on the assessment matrix, experimental analyses using the AQUA-SF and SPHINCS codes for a sodium spray fire experiment Run-E1 show good agreement with the experimental result.
Aoyagi, Mitsuhiro; Uchibori, Akihiro; Kikuchi, Shin; Takata, Takashi; Ohno, Shuji; Ohshima, Hiroyuki
Nuclear Engineering and Design, 353, p.110240_1 - 110240_10, 2019/11
Times Cited Count：0 Percentile：100(Nuclear Science & Technology)JAEA has developed sodium fire analysis codes to evaluate the consequences of sodium fire events. This paper describes a PIRT (Phenomena Identification and Ranking Table) process for such events. Because a sodium fire event involves complex phenomena, the FOMs are specified through a factor analysis. Associated phenomena in a sodium fire event are identified through both element- and sequence-based phenomena analyses. The importance of each phenomenon is evaluated by considering the sequence-based analysis of associated phenomena related to the FOMs. Then, a ranking table is established. An assessment matrix of important phenomena and experiments is completed to confirm the sufficiency of experimental data for the validation of the models in the sodium fire analysis codes. Additional assessments are discussed specifically for the aerosol module and the CFD module in three-dimensional codes from a perspective of careful validation.
Uchibori, Akihiro; Shiina, Yoshimi*; Watanabe, Akira*; Takata, Takashi
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.5315 - 5324, 2019/08
A computational fluid dynamics code SERAPHIM for a sodium-water reaction phenomenon has been developed. An unstructured mesh-based analysis method has been integrated into this code for evaluation in a complex-shaped domain including multiple heat transfer tubes in our recent studies. In this study, a numerical analysis of an experiment on sodium-water reaction in a tube bundle domain was performed to investigate applicability of the unstructured mesh-based SERAPHIM code. The formulated high-temperature region and its temperature level agreed with an experimental result.
Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Li, J.*; Jang, S.*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 6 Pages, 2019/05
A numerical analysis model to predict occurrence of tube failure propagation by overheating rupture in steam generators of sodium-cooled fast reactors was developed. Applicability of the model was demonstrated through the numerical analysis of the experiment on water vapor discharging in liquid sodium. To improve an accuracy of a sodium-side temperature evaluation, a Lagrangian particle model for simulating the reacting jet was also developed. The numerical results by the program unit of this model showed that the discharged gaseous particles spread with a particle-particle and particle-tube interaction.
Sonehara, Masateru; Aoyagi, Mitsuhiro; Uchibori, Akihiro; Takata, Takashi; Ohshima, Hiroyuki; Clark, A. J.*; Denman, M. R.*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
In order to investigate the effect of sodium combustion, Sandia National Laboratories (SNL) and Japan Atomic Energy Agency (JAEA) have exchanged information of sodium combustion modelling and related experimental data in the framework of Civil Nuclear Energy Research and Development Working Group (CNWG). The benchmark analysis of the SNL T3 sodium spray combustion experiment and sensitivity study have been carried out using the AQUA-SF code in this paper. The sensitivity analysis clarifies the influencing factors of the multi-dimensional analysis such as turbulence models, radiation heat transfer model from sodium droplets, and momentum exchange between gas and droplets. The result shows that the turbulence effect, radiation from droplets and gas temperature increase at spray burning area affect sodium spray burning rate significantly.
Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Kurihara, Akikazu; Hamada, Hirotsugu; Ohshima, Hiroyuki
Journal of Nuclear Science and Technology, 56(2), p.201 - 209, 2019/02
Times Cited Count：0 Percentile：100(Nuclear Science & Technology)Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was developed to expand application range of an existing computer code. Applicability of the method was demonstrated through the numerical analysis of the experiment on water vapor discharging in liquid sodium.
Uchibori, Akihiro; Watanabe, Akira*; Takata, Takashi; Ohshima, Hiroyuki
Nuclear Technology, 205(1-2), p.119 - 127, 2019/01
Times Cited Count：0 Percentile：100(Nuclear Science & Technology)To evaluate a sodium-water reaction phenomenon in a steam generator of sodium-cooled fast reactors, a computational fluid dynamics code SERAPHIM, in which a compressible multicomponent multiphase flow with sodium-water chemical reaction is computed, has been developed. The original SERAPHIM code is based on the difference method. In this study, unstructured mesh-based numerical method was developed to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Numerical analysis of an underexpanded jet experiment was performed as part of validation of the unstructured mesh-based numerical method. The calculated pressure profile and location of the Mach disk showed good agreement with the experimental data. Applicability of the numerical method for the actual situation was confirmed through the analysis of water vapor discharging into liquid sodium.
Sonehara, Masateru; Aoyagi, Mitsuhiro; Uchibori, Akihiro; Takata, Takashi; Ohshima, Hiroyuki; Clark, A. J.*; Denman, M. R.*
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 5 Pages, 2018/11
no abstracts in English
Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Ohshima, Hiroyuki
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 6 Pages, 2018/11
Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was constructed to expand application range of an existing computer code. Applicability of the method was constructed through the numerical analysis of the experiment on water vapor discharging in liquid sodium.
Li, J.*; Jang, S.*; Yamaguchi, Akira*; Uchibori, Akihiro; Takata, Takashi; Ohshima, Hiroyuki
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 4 Pages, 2018/11
The sodium-water reaction model is developed in particle methods. Two chemical reaction model, called surface reaction model and gas-phase reaction model are developed in the particle method. Validation on the case of vapor injection into liquid water is conducted and good consistency of jet velocity evolution along jet trajectory is obtained. Finally, sodium-water chemical reaction in a configuration of multiple tube bundles is simulated. Jet velocity, water vapor fraction and temperature are investigated and reasonable results are observed, which presents promising future of this methodology.
Uchibori, Akihiro; Takata, Takashi; Yanagisawa, Hideki*; Li, J.*; Jang, S.*
Proceedings of 2018 ANS Winter Meeting and Nuclear Technology Expo; Embedded Topical International Topical Meeting on Advances in Thermal Hydraulics (ATH 2018) (USB Flash Drive), p.1289 - 1294, 2018/11
Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was constructed to expand application range of an existing computer code. Applicability of the method was constructed through the numerical analysis of the experiment on water vapor discharging in liquid sodium. To improve the evaluation accuracy for the temperature distribution, a Lagrangian particle model for simulating reacting jet was also developed as an alternative method and its basic function was confirmed.
Sugimoto, Taro*; Saito, Shimpei*; Kaneko, Akiko*; Abe, Yutaka*; Uchibori, Akihiro; Ohshima, Hiroyuki
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 7 Pages, 2018/07
A computational fluid dynamics code for a sodium-water reaction phenomenon in a steam generator of sodium-cooled fast reactors has been developed. In order to provide the data for validation of this code, the visualization experiment on liquid droplet entrainment in the high-pressure air jet submerged in the water pool was carried out. The experiment successfully elucidated the behavior, such as atomization of the relatively large diameter liquid droplet generated from the gas-liquid interface.
Uchibori, Akihiro; Watanabe, Akira*; Takata, Takashi; Ohshima, Hiroyuki
Nippon Kikai Gakkai Rombunshu (Internet), 84(859), p.17-00394_1 - 17-00394_6, 2018/03
For assessment of the wastage environment under tube failure accident in a steam generator of sodium-cooled fast reactors, a mechanistic computer code called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction has been developed. The original SERAPHIM code is based on the finite difference method. In this study, unstructured mesh-based numerical method was developed and introduced into the SERAPHIM code to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Validity of the unstructured mesh-based SERAPHIM code was investigated through the analysis of an underexpanded jet experiment. The calculated pressure profile showed good agreement with the experimental data. Numerical analysis of water vapor discharging into liquid sodium was also performed. It was demonstrated that the proposed numerical method could be applicable to evaluation of the sodium-water reaction phenomenon.
Uchibori, Akihiro; Watanabe, Akira*; Takata, Takashi; Ohshima, Hiroyuki
Journal of Nuclear Science and Technology, 54(10), p.1036 - 1045, 2017/10
Times Cited Count：3 Percentile：48.33(Nuclear Science & Technology)To evaluate a sodium-water reaction phenomenon in a steam generator of sodium-cooled fast reactors, a computational fluid dynamics code SERAPHIM, in which a compressible multicomponent multiphase flow with sodium-water chemical reaction is computed, has been developed. The original SERAPHIM code is based on the difference method. In this study, unstructured mesh-based numerical method was developed to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Numerical analysis of an underexpanded jet experiment was performed as part of validation of the unstructured mesh-based numerical method. The calculated pressure profile showed good agreement with the experimental data. Applicability of the numerical method for the actual situation was confirmed through the analysis of water vapor discharging into liquid sodium. The effect of use of the unstructured mesh was also investigated by the two analyses using structured and unstructured mesh.
Uchibori, Akihiro; Takata, Takashi; Ohshima, Hiroyuki; Watanabe, Akira*
Proceedings of 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17) (USB Flash Drive), 12 Pages, 2017/09
To evaluate a sodium-water reaction phenomenon in a steam generator of sodium-cooled fast reactors, a computational fluid dynamics code SERAPHIM, in which a compressible multicomponent multiphase flow with sodium-water chemical reaction is computed, has been developed. The original SERAPHIM code is based on the difference method. In this study, unstructured mesh-based numerical method was developed to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Numerical analysis of an underexpanded jet experiment was performed as part of validation of the unstructured mesh-based numerical method. The calculated pressure profile and location of the Mach disk showed good agreement with the experimental data. Applicability of the numerical method for the actual situation was confirmed through the analysis of water vapor discharging into liquid sodium.
Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Kurihara, Akikazu; Hamada, Hirotsugu; Ohshima, Hiroyuki
JAEA-Research 2017-007, 61 Pages, 2017/07
JAEA-Research-2017-007.pdf:4.3MB
For safety assessment of a steam generator of sodium-cooled fast reactors, it is necessary to evaluate the possibility of occurring tube failure propagation and of water leak rate under sodium-water reaction accident. In the previous studies, a computer code called LEAP-II calculating a wastage-type failure propagation and the water leak rate during long-time event progress was developed. In this study, a numerical method to evaluate the possibility of occurring overheating rupture was introduced into the LEAP-II code to expand application range of this code. The completed code is called LEAP-III. The test analysis on a tube bundle configuration demonstrated that the overheating rupture model could provide conservative prediction.
Aoyagi, Mitsuhiro; Uchibori, Akihiro; Kikuchi, Shin; Takata, Takashi; Ohno, Shuji; Ohshima, Hiroyuki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06
The present PIRT process is aimed to identify key phenomena involved in sodium fire accidents that involve complex phenomena in sodium-cooled fast reactor plants. In this PIRT process, the figures of merit (FOMs) are specified through factor analysis. Associated phenomena are identified through the element- and sequence-based phenomena analyses. Importance of each associated phenomenon is evaluated by considering the sequence-based analysis of associated phenomena correlated with the FOMs. Then, we complete the ranking table through the factor and phenomenon analyses. An assessment matrix of important phenomena and experiments is completed finally for model validation.
