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Wen, J.*; Kamada, Yuto*; Yokoyama, Kosei*; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*; Imaizumi, Yuya; Tagami, Hirotaka; Matsuba, Kenichi; Kamiyama, Kenji
Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 8 Pages, 2024/11
Kondo, Satoru; Tobita, Yoshiharu*; Morita, Koji*; Kamiyama, Kenji; Yamano, Hidemasa; Suzuki, Toru*; Tagami, Hirotaka; Sogabe, Joji; Ishida, Shinya
JAEA-Research 2024-008, 235 Pages, 2024/10
JAEA-Research-2024-008.pdf:4.77MB
The SIMMER-III and SIMMER-IV computer codes, developed at the Japan Atomic Energy Agency are the codes with two- and three-dimensional, multi-field, multi-component fluid-dynamics models, coupled with a space- and time-dependent neutron kinetics model. The codes have been used widely for simulating complex phenomena during core-disruptive accidents in liquid-metal fast reactors. Advanced features of the codes in comparison with the former codes include: stable and robust fluid-dynamics algorithm with up to 8 velocity fields, improved representation of structures and multi-phase flow topology, comprehensive treatment of complex heat and mass transfer processes, accurate analytic equations of state, a stable and efficient neutron flux shape solution method and decay heat model. This report describes the models and methods of SIMMER-III and SIMMER-IV. For those individual models, the details of which have been reported elsewhere, only the outlines of the models are presented. The reports of code verification and validation have been already published.
Yamano, Hidemasa; Futagami, Satoshi; Doda, Norihiro; Tagami, Hirotaka; Uchibori, Akihiro; Ogata, Takanari*; Ota, Hirokazu*
Nihon Kikai Gakkai 2024-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2024/09
Wen, J.*; Kamada, Yuto*; Yokoyama, Kosei*; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*; Imaizumi, Yuya; Tagami, Hirotaka; Matsuba, Kenichi; Kamiyama, Kenji
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 8 Pages, 2024/08
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
Tagami, Hirotaka; Tobita, Yoshiharu
Nuclear Engineering and Technology, 56(3), p.873 - 879, 2024/03
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)A sodium-cooled fast reactor (SFR) core has a potential of prompt criticality due to a change of core material distribution during a severe accident, and the resultant energy release has been one of the safety issues of SFRs. In this study, the safety assessment of an unprotected loss-of-flow (ULOF) in a small SFR (SSFR) has been performed using the SIMMER-IV computer code, which couples the models of space- and time-dependent neutronics and multi-component, multi-field thermal hydraulics in three dimensions. The code, therefore, is applicable to the simulations of transient behaviors of extended disrupted core material motion and its reactivity effects during the transition phase (TP) of ULOF, including a potential of prompt-criticality power excursions driven by fuel compaction. Several conservative assumptions are used in the TP analysis by SIMMER-IV. It was found out that one of the important mechanisms that drives the reactivity-inserting fuel motion was sodium vapor pressure resulted from a fuel-coolant interaction (FCI), which itself was non-energetic local phenomenon. The uncertainties relating to FCI is also evaluated in much conservative way in the sensitivity analysis. From this study, the ULOF characteristics in an SSFR have been understood. Occurrence of recriticality events under conservative assumptions are plausible, but their energy releases are limited.
Tagami, Hirotaka; Ishida, Shinya; Tobita, Yoshiharu
Journal of Nuclear Science and Technology, 60(12), p.1548 - 1562, 2023/12
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)In a design of future Sodium-cooled Fast Reactor, there is a demand for evaluation of sequences and consequences of core disruptive accidents. Future SFRs include a unique core design with axially or horizontally heterogeneous core arrangement having complex fuel isotope distribution. A new model to flexibly represent fuel isotope distribution, called the Pu-vector model, has been developed in this study for inclusion in the SIMMER-III and SIMMER-IV codes (simply called as SIMMER). The model calculates movement of individual fuel isotopes, assuming they always accompany the convecting fuel in the fluid-dynamics model. The accuracy of the Pu-vector model was confirmed by comparing with the standard Monte Carlo static neutronics calculation. The new model can improve some of the limitations in the current SIMMER code, in which the fuel isotopes are represented only by two groups, fertile and fissile fuels. Assignment of a number of fuel isotopes to the two groups requires a detailed examination of different combinations of fuel isotopes to determine an optimized combination. The Pu-vector model can eliminate this complicated procedure to be performed prior to a SIMMER analysis, and more importantly provides accurate spatial distribution of fuel isotopes and thus will improve the applicability of SIMMER to the analyses of future large heterogeneous reactors.
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
Tagami, Hirotaka; Tobita, Yoshiharu
Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 8 Pages, 2023/04
Recently, the safety analysis for a licensing of small nuclear power fast reactor is performed. It is necessary to confirm the effectiveness of the design measure to prevent the CV failure in the licensing procedure. Because the energy generation in TP of ULOF is one of the main factors to affect the integrity of CV, the ULOF behavior is analyzed using SIMMER developed under international cooperation. Although the characteristic of TP in small reactor is a slow and mild event progression due to the negative void reactivity, several conservative assumptions are applied in the analysis. Because the prompt criticality by fuel compaction is mainly driven by a fuel coolant interaction, its impact on energy generation is also investigated by conservatively assuming uncertainties. The obtained results by the analysis using SIMMER are used for the subsequent phase to analyze the mechanical integrities of reactor vessel and CV.
Phan, L. H. S.*; Ohara, Yohei*; Kawata, Ryo*; Liu, X.*; Liu, W.*; Morita, Koji*; Guo, L.*; Kamiyama, Kenji; Tagami, Hirotaka
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 12 Pages, 2018/10
Self-leveling behavior of core fuel debris beds is one of the key phenomena for the safety assessment of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). The SIMMER code has been developed for CDA analysis of SFRs, and the code has been successfully applied to numerical simulations for key thermal-hydraulic phenomena involved in CDAs as well as reactor safety assessment. However, in SIMMER's fluid-dynamics model, it is always difficult to represent the strong interactions between solid particles as well as the discrete particle characteristics. To solve this problem, a new method has been developed by combining the multi-fluid model of the SIMMER code with the discrete element method (DEM) for the solid phase to reasonably simulate the particle behaviors as well as the fluid-particle interactions in multi-phase flows. In this study, in order to validate the multi-fluid model of the SIMMER code coupled with DEM, numerical simulations were performed on a series of self-leveling experiments using a gas injection method in cylindrical particle beds. The effects of friction coefficient on the simulation results were investigated by sensitivity analysis. Though more extensive validations are needed, the reasonable agreement between simulation results and corresponding experimental data preliminarily demonstrates the potential ability of the present method in simulating the self-leveling behaviors of debris bed. It is expected that the SIMMER code coupled with DEM is a prospective computational tool for analysis of safety issues related to solid particle debris bed in SFRs.
Tagami, Hirotaka; Cheng, S.*; Tobita, Yoshiharu; Morita, Koji*
Nuclear Engineering and Design, 328, p.95 - 106, 2018/03
Times Cited Count：11 Percentile：70.37(Nuclear Science & Technology)Shamsuzzaman, M.*; Horie, Tatsuro*; Fuke, Fusata*; Kamiyama, Motoki*; Morioka, Toru*; Matsumoto, Tatsuya*; Morita, Koji*; Tagami, Hirotaka; Suzuki, Toru*; Tobita, Yoshiharu
Annals of Nuclear Energy, 111, p.474 - 486, 2018/01
Times Cited Count：18 Percentile：83.50(Nuclear Science & Technology)Tobita, Yoshiharu; Kamiyama, Kenji; Tagami, Hirotaka; Matsuba, Kenichi; Suzuki, Toru; Isozaki, Mikio; Yamano, Hidemasa; Morita, Koji*; Guo, L.*; Zhang, B.*
Journal of Nuclear Science and Technology, 53(5), p.698 - 706, 2016/05
Times Cited Count：29 Percentile：92.90(Nuclear Science & Technology)The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior.
Suzuki, Toru; Tobita, Yoshiharu; Kawada, Kenichi; Tagami, Hirotaka; Sogabe, Joji; Matsuba, Kenichi; Ito, Kei; Ohshima, Hiroyuki
Nuclear Engineering and Technology, 47(3), p.240 - 252, 2015/04
Times Cited Count：28 Percentile：89.76(Nuclear Science & Technology)Tagami, Hirotaka; Tobita, Yoshiharu
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 12 Pages, 2014/12
Cheng, S.; Tagami, Hirotaka; Yamano, Hidemasa; Suzuki, Toru; Tobita, Yoshiharu; Takeda, Shohei*; Nishi, Shimpei*; Nishikido, Tatsuya*; Zhang, B.*; Matsumoto, Tatsuya*; et al.
Journal of Nuclear Science and Technology, 51(9), p.1096 - 1106, 2014/09
Times Cited Count：26 Percentile：87.71(Nuclear Science & Technology)Cheng, S.; Tagami, Hirotaka; Yamano, Hidemasa; Suzuki, Toru; Tobita, Yoshiharu; Nakamura, Yuya*; Takeda, Shohei*; Nishi, Shimpei*; Zhang, B.*; Matsumoto, Tatsuya*; et al.
Mechanical Engineering Journal (Internet), 1(4), p.TEP0022_1 - TEP0022_16, 2014/08
Suzuki, Toru; Tobita, Yoshiharu; Kawada, Kenichi; Tagami, Hirotaka; Sogabe, Joji; Ito, Kei
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 10 Pages, 2014/07
Tagami, Hirotaka; Cheng, S.; Tobita, Yoshiharu; Guo, L.*; Zhang, B.*; Morita, Koji*
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 8 Pages, 2014/07
The object of this study is to develop new analytical methods to simulate unique phenomena in self-leveling behavior and implement it to SFR safety analysis code. The new methods are developed with assuming that the debris bed behaves as Bingham fluid from this feature. They are categorized into two main parts. The first part is particle interaction models to model the effect of particle-particle collisions. The second part is a large deformation method, which simulates Bingham fluid characteristic of debris bed. An experimental study of self-leveling behavior is analyzed to validate the new methods. The assessment results show that these methods provide a basis to develop analytical methods of self-leveling behavior of debris bed in the safety assessment of SFRs.
Cheng, S.; Tagami, Hirotaka; Yamano, Hidemasa; Suzuki, Toru; Tobita, Yoshiharu; Zhang, B.*; Matsumoto, Tatsuya*; Morita, Koji*
Annals of Nuclear Energy, 63, p.188 - 198, 2014/01
Times Cited Count：33 Percentile：91.75(Nuclear Science & Technology)