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Louie, D. L. Y.*; Aoyagi, Mitsuhiro
SAND2022-14235 (Internet), 29 Pages, 2022/10
This report discusses the progress on the collaboration between Sandia National Laboratories (SNL) and Japan Atomic Energy Agency (JAEA) on the sodium fire research in fiscal year 2022. First, the current sodium pool fire model in MELCOR is discussed. The associated sodium fire input requirements are also presented. The theoretical pool fire model improvement developed at SNL is discussed. A control function model has been developed from this improvement. Then, the validation study of the sodium pool fire model in MELCOR is described. To validate this pool fire model with the enhancement, JAEA F7-1 and F7-2 sodium pool fire experiments are used. The results of the calculation are discussed as well as suggestions for further model improvement. Finally, recommendations are made for new MELCOR simulations for next fiscal year, 2023.
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
Louie, D. L. Y.*; Aoyagi, Mitsuhiro; Uchibori, Akihiro; Takata, Takashi; Luxat, D. L.*
Proceedings of Technical Meeting on State-of-the-art Thermal Hydraulics of Fast Reactors (Internet), 6 Pages, 2022/09
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.
Kijima, Jun; Koyama, Hayato; Owada, Mitsuhiro; Hagiwara, Masayoshi; Aoyagi, Yoshitaka
JAEA-Technology 2022-012, 14 Pages, 2022/07
JAEA-Technology-2022-012.pdf:1.51MB
Steam reforming system has been developed for the treatment of organic wastes which are not suitable materials (halogenated oil) for the incineration due to generation of corrosive compounds and plugging materials. The refractory material is cast inside the main reactor, which is a part of the steam reforming system. Since the surface of this refractory material has deteriorated over time, the main reactor was replaced. If the refractory material surface of the used main reactor can be repaired, the used main reactor can be reused as a spare. The refractory material surface was repaired using two types of repair materials ("S" and "P"). Combustion tests were conducted on samples simulating organic wastes to evaluate each repair material. As a result of the combustion test, it was concluded that the repair of the main reactor was possible to use the repair material "P" because no cracks or flakes were observed.
Imaizumi, Yuya; Aoyagi, Mitsuhiro; Kamiyama, Kenji; Matsuba, Kenichi; Akayev, A. S.*; Mikisha, A. V.*; Baklanov, V. V.*; Vurim, A. D.*
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2022/07
The cooling of the residual core materials after the fuel discharge from the SFR core in the core disruptive accident can significantly affect the distribution fraction of the core materials which is an important factor for the in-vessel retention (IVR). The cooling of the residual core materials is called "in-place cooling". For the evaluation of the in-place cooling, behavior in a SFR core was simulated by SIMMER-III, and method of phenomena identification and ranking table (PIRT) was applied based on the analysis result. Experiment which focuses on the thermal-hydraulic phenomena which were extracted by the PIRT was conducted in the framework of EAGLE-3 project. Continuous oscillation of sodium level which can occur in the phase of in-place cooling of SFRs was observed in the experiment, and analysis by the SIMMER-III was conducted. By investigation of the analysis result, difference between the experiment and analysis results was revealed to be due to remaining and occupation of non-condensable gas above the sodium level which would be unrealistic in the experiment. Gas mixture model between non-condensable gas and sodium vapor was developed to solve this problem, and coincidence between experiment and analysis results was largely improved by this new model.
Louie, D. L. Y.*; Aoyagi, Mitsuhiro
Proceedings of International Topical Meetings on Advances in Thermal Hydraulics (ATH 2022) (Internet), p.316 - 329, 2022/06
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.
Aoyagi, Mitsuhiro; Uchibori, Akihiro; Takata, Takashi
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 13 Pages, 2022/03
Louie, D. L. Y.*; Aoyagi, Mitsuhiro
SAND2021-15469 (Internet), 45 Pages, 2021/12
This report discusses the progress on the collaboration between Sandia National Laboratories (SNL) and Japan Atomic Energy Agency (JAEA) on the sodium fire research in fiscal year 2021. First, the current sodium pool fire model in MELCOR is discussed. The associated sodium fire input requirements are also presented. The theoretical pool fire model improvement developed at SNL is discussed. A control function model has been developed from this improvement. Then, the validation study of the sodium pool fire model in MELCOR is described. To validate this pool fire model with the enhancement, JAEA F7-1 and F7-2 sodium pool fire experiments are used. The results of the calculation are discussed as well as suggestions for further model improvement. Finally, recommendations are made for new MELCOR simulations for next fiscal year, 2021.
Zhan, Y.*; Sun, G.*; Okawa, Tomio*; Aoyagi, Mitsuhiro; Takata, Takashi
Experimental Thermal and Fluid Science, 126, p.110402_1 - 110402_8, 2021/08
Times Cited Count：4 Percentile：55.67(Thermodynamics)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.
Aoyagi, Mitsuhiro; Takata, Takashi; Uno, Masayoshi*
Nuclear Engineering and Design, 380, p.111258_1 - 111258_11, 2021/08
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Aoyagi, Mitsuhiro; Louie, D. L. Y.*; Uchibori, Akihiro; Takata, Takashi; Luxat, D.*
Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 10 Pages, 2021/08
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.
Louie, D. L. Y.*; Aoyagi, Mitsuhiro
Transactions of the American Nuclear Society, 124(1), p.824 - 827, 2021/06
The Sodium Chemistry (NAC) package in MELCOR has been developed to enhance application to sodium cooled fast reactors. Based on the recommendations in the previous study through the benchmark analyses of the F7-1 pool fire experiment, this study aims to improve the MELCOR models capturing the oxide layer effect, sodium pool spreading and pool-pan heat transfer, respectively. Each of these models enable a better characterization of the all the processes of relevance to sodium pool fires as observed during the F7-1 test. The MELCOR sodium pool fire enhancement has demonstrated the importance of the improved models.
Zhan, Y.*; Kuwata, Yusuke*; Okawa, Tomio*; Aoyagi, Mitsuhiro; Takata, Takashi
Experimental Thermal and Fluid Science, 120, p.110249_1 - 110249_12, 2021/01
Times Cited Count：4 Percentile：67.66(Thermodynamics)Louie, D. L. Y.*; Aoyagi, Mitsuhiro
SAND2021-0136 (Internet), 53 Pages, 2021/01
This report discusses the progress on the collaboration between Sandia National Laboratories (SNL) and Japan Atomic Energy Agency (JAEA) on the sodium fire research in fiscal year 2020. First, the current sodium pool fire model in MELCOR is discussed. The associated sodium fire input requirements are also presented. These input requirements are flexible enough to permit further model development via control functions to enhance the current model without modifying the source code. The theoretical pool fire model improvement developed at SNL is discussed. A control function model has been developed from this improvement. Then, the validation study of the sodium pool fire model in MELCOR is described. To validate this pool fire model with the enhancement, a JAEA F7-1 sodium pool fire experiment is used. The results of the calculation are discussed as well as suggestions for further model improvement. Finally, recommendations are made for new MELCOR simulations for next fiscal year, 2021.
Aoyagi, Mitsuhiro; Uchibori, Akihiro; Takata, Takashi; Louie, D. L. Y.*; Clark, A. J.*
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08
The Sodium Chemistry (NAC) package in MELCOR has been developed to enhance application to sodium cooled fast reactors. The models in the NAC package have been assessed through benchmark analyses of the F7-1 pool fire experiment. This study assesses the capability of the pool fire model in MELCOR and provides recommendations for future model improvements. The MELCOR analysis yields lower values than the experimental data in pool combustion rate and pool, catch pan, and gas temperature during early time. The current heat transfer model for the catch pan is the primary cause of the difference. After sodium discharge stopping, the pool combustion rate and temperature become higher than experimental data. This is caused by absence of a model for pool fire suppression due to the oxide layer buildup on the pool surface. Based on these results, recommendations for future works are needed, such as heat transfer modification for the catch pan and consideration of the effects of the oxide layer.
Uchibori, Akihiro; Aoyagi, Mitsuhiro; Takata, Takashi; Ohshima, Hiroyuki
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 7 Pages, 2020/08
The multi-scenario simulation system named SPECTRA has been developed for integrated analysis of in- and ex-vessel phenomena during a severe accident in sodium-cooled fast reactors. The base module computing ex-vessel compressible gas behavior by a lumped mass model and a sodium-concrete interaction module were verified through the basic analyses individually. A validity of the system including the base module and the individual physical module such as the sodium-concrete interaction module was confirmed through the analysis assuming sodium leakage from a reactor vessel and a primary cooling loop.
