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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.
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; Sonehara, Masateru; Aoyagi, Mitsuhiro; Takata, Takashi*; Ohshima, Hiroyuki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04
A new computational code, SPECTRA, has been developed for integrated analysis of in- and ex-vessel phenomena during severe accidents in sodium-cooled fast reactors. The in-vessel thermal hydraulics module includes coupled analytical models for multidimensional multifluid model considering compressibility and relocation of a molten core. A lumped mass model is employed for computing behavior of ex-vessel compressible multicomponent gas including aerosols. This model is coupled with the models for ex-vessel phenomena such as sodium fire. Loss of reactor level event starting from leakage of sodium coolant was computed. Basic capability to evaluate severe accident progress was demonstrated through this analysis.
Takata, Takashi; Aoyagi, Mitsuhiro; Sonehara, Masateru
IAEA-TECDOC-1972, p.224 - 234, 2021/08
Sodium fire is one of the key issues for plant safety of sodium-cooled fast reactor (SFR) regardless of its size. In general, a concrete structure, which includes free and bonging water inside, is used in a reactor building. Accordingly, water vapor will release from the concrete during sodium fire incident due to temperature increase resulting in a hydrogengeneration even in a dry air condition. The probability of hydrogen generation will increase in accordance with a decrease of a dimension of compartment that corresponds to a small and medium sized or modular reactor (SMR). A numerical investigation of a small leakage sodium pool fire has been carried out by changing a dimension of compartment. Furthermore, numerical challenges to enhance a prediction accuracy of hydrogen generation during sodium fire has also been discussed in the paper.
Sonehara, Masateru; Uchibori, Akihiro; Aoyagi, Mitsuhiro; Kawada, Kenichi; Takata, Takashi; Ohshima, Hiroyuki
Dai-25-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 3 Pages, 2021/07
In sodium-cooled fast reactors (SFRs), it has been pointed out that molten fuel may be discharged from the core during a severe accident (SA) accompanied by core damage, and may solidify into debri particles with diameters ranging from several millimeters to several hundred micrometers due to interaction with the sodium coolant and accumulate at the bottom of the reactor vessel. Therefore, it is necessary to understand the behavior of such debri particles appropriately to evaluate the SA event progression. To meet these requirements, a molten fuel behavior analysis code using dissipative particle dynamics (DPD), a kind of particle method, has been developed as a part of the SPECTRA code, tool for consistent analysis of in-vessel and ex-vessel events in sodium fast reactor accidents. In this study, it was found that the new analyses code can reproduce sedimentation behavior of particles by adding a new stress term in the shear direction.
Sonehara, Masateru; Aoyagi, Mitsuhiro; Uchibori, Akihiro; Takata, Takashi; Ohshima, Hiroyuki; Clark, A. J.*; Louie, D. L. Y.*
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 4 Pages, 2020/08
In order to investigate the multi-dimensional effects of sodium combustion, a benchmark analysis of the SNL Surtsey spray combustion experiment (SNL T3 experiments) using AQUA-SF and SPHINCS is conducted in JAEA. As a best estimate analysis, the spray burning duration is adjusted in the computation in order to take into account the temporary suppression of the spray combustion observed in the experiment. Furthermore, droplet size of SPHINCS and AQUA-SF are optimized to represent the T3 experimental results. The best estimate of AQUA-SF results in the droplet diameter of 2.5 mm, which agrees quite well with the spatial temperature measurements, and the sodium droplet diameter measurement with a high speed camera.
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.
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
Aoyagi, Mitsuhiro; Sonehara, Masateru; Uchibori, Akihiro; Takata, Takashi; Ohshima, Hiroyuki
no journal, ,
Development of the multi-scenario simulation systems for in-vessel/ex-vessel phenomena under severe accident in sodium cooled fast reactor is an important issue. In this study, physical models for the multi-scenario simulation systems were developed, such as a model for melting fuel behavior using the dissipative particle dynamics method.
Uchibori, Akihiro; Aoyagi, Mitsuhiro; Sonehara, Masateru; Takata, Takashi; Ohshima, Hiroyuki
no journal, ,
The multi-level, multi-scenario simulation systems have been developed as a fundamental technology of sodium-cooled fast reactors. In this study, a multi-scenario simulation system for in- and ex-vessel phenomena during a severe accident was newly developed. The validity of the system was confirmed through the analysis of a Loss-Of-Reactor-Level (LORL) event.
Uchibori, Akihiro; Sonehara, Masateru; Aoyagi, Mitsuhiro; Kawada, Kenichi; Takata, Takashi; Ohshima, Hiroyuki
no journal, ,
A computational code, SPECTRA was developed for integrated analysis of in- and ex-vessel phenomena during severe accidents in sodium-cooled fast reactors. SPECTRA consists of in- and ex-vessel modules which have a thermal hydraulics module as a base part. The in-vessel thermal hydraulics module computes complicated multi-dimensional behavior of liquid sodium and gas by using the multi-fluid model considering compressibility. Relocation of a molten core is computed by the dissipative particle dynamics method. A lumped mass model was employed for computation of ex-vessel multi-component gas flow including aerosols. Analytical models for sodium fire, sodium-concrete interaction, and debris-concrete interaction were integrated into the ex-vessel module. Basic capability of SPECTRA was demonstrated through analysis of a loss of reactor level event of a loss of reactor level event.
Aoyagi, Mitsuhiro; Enuma, Yasuhiro; Sonehara, Masateru; Kosaka, Wataru
no journal, ,
no abstracts in English
Uchibori, Akihiro; Aoyagi, Mitsuhiro; Sonehara, Masateru; Sogabe, Joji; Okano, Yasushi; Takata, Takashi*
no journal, ,
The ARKADIA-Safety for safety evaluation and design optimization considering severe accident has been developed. This paper presents validation of sodium fire analysis model and integration of ex-vessel analysis model for improvement of the evaluation method.
Sonehara, Masateru; Aoyagi, Mitsuhiro; Kosaka, Wataru; Uchibori, Akihiro; Okano, Yasushi
no journal, ,
In order to improve user convenience based on the assumption that the development system will be provided to the private sector, we constructed the basic parts of an AI-based design optimization tool, a user interface input GUI tool, and a quality assurance automation tool.
Uchibori, Akihiro; Aoyagi, Mitsuhiro; Sonehara, Masateru; Okano, Yasushi; Takata, Takashi*
no journal, ,
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. A new computational code, SPECTRA, for integrated analysis of in- and ex-vessel phenomena during severe accidents in sodium-cooled fast reactors is a key technology of ARKAIDA-Safety which aims for safety evaluation. The in-vessel thermal hydraulics module in this code includes coupled analytical models for multidimensional multifluid model considering compressibility and relocation of a molten core. A lumped mass model is employed for computing behavior of ex-vessel compressible multicomponent gas including aerosols. This model is coupled with the models for ex-vessel phenomena such as sodium fire. Loss of reactor level event starting from leakage of sodium coolant was computed. Basic capability to evaluate severe accident progress was demonstrated through this analysis.
Okano, Yasushi; Uchibori, Akihiro; Aoyagi, Mitsuhiro; Sonehara, Masateru; Ishida, Shinya; Sogabe, Joji; Takata, Takashi*
no journal, ,
no abstracts in English
Uchibori, Akihiro; Sonehara, Masateru; Aoyagi, Mitsuhiro; Okano, Yasushi; Takata, Takashi
no journal, ,
A new computational code, SPECTRA, has been developed as a base simulation system for integrated safety evaluation of advanced sodium-cooled fast reactors. In this study, a lumped mass module for in-vessel coolant behavior and user convenience improvement tool were constructed. The lumped mass module was validated through the analysis of a compressible gas-liquid two-phase flow experiment by using the computational domain connected with a multidimensional mesh. As user convenience improvement, the AI-driven design tool successfully found an optimum solution in a single parameter problem. The GUI tool supporting creation of a input data and quality assurance of analysis work was also constructed.
Sonehara, Masateru; Okano, Yasushi; Uchibori, Akihiro; Oki, Hiroshi*
no journal, ,
In the development of sodium-cooled fast reactors, it is important to understand the combustion behavior in sodium leak accidents. In this study, the multidimensional thermal fluid analysis code AQUA-SF was applied to the numerical simulation of the large-scale combustion test (so-called the T3 test) with accounting characteristic event progressions (temporary stop of combustion, droplet splash on the floor).
