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Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji; Akaev, A.*; Vurim, A.*; Baklanov, V.*
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 12 Pages, 2022/09
The In-Vessel Retention (IVR) of molten-core in Core Disruptive Accidents (CDAs) is of prime importance in enhancing the safety of sodium-cooled fast reactors. One of the main subjects in ensuring IVR is to design the Control Rod Guide Tube (CRGT) which allows effective discharge of molten core materials from the core region. The effectiveness of the CRGT design is assessed through CDA analyses, and it is reasonable for these analyses to develop a computer code collaborated with experimental researches. Thus, experiments addressing the discharge behavior of the molten-core materials through the CRGT have proceeded as one of the subjects in the collaboration research named the EAGLE-3 project, and the obtained experimental results are reflected in the development of the SIMMER code. In this project, a series of out-of-pile tests using molten-alumina as the fuel simulant was conducted to understand the discharge behavior of molten-core materials through the CRGT. In this study, in order to investigate the effect of an internal structure in the CRGT on the discharge behavior of the molten-core materials, the data of an out-of-pile test in which the molten-alumina penetrated to a duct with the internal structure were analyzed. In addition, the post-test analysis using the SIMMER code was conducted and the results were compared with the test results.
Matsushita, Hatsuki*; Kobayashi, Ren*; Sakai, Takaaki*; Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 9 Pages, 2022/09
During core disruptive accidents in sodium-cooled fast reactors, the molten core material flows through flow channels, such as the control rod guide tubes, into the core inlet plenum under the core region. The molten core material can be cooled and solidified while impinging on a horizontal plate of the inlet plenum in a sodium coolant. However, the solidification and cooling behaviors of molten core materials impinged on a horizontal structure have not been sufficiently studied thus far. Notably, this is an important phenomenon that needs to be elucidated from the perspective of improving the safety of sodium-cooled fast reactors. Accordingly, a series of experiments on discharging a simulated molten core material (alumina: Al
O
) into a sodium coolant on a horizontal structure was conducted at the experimental facility of the National Nuclear Center of the Republic of Kazakhstan. In this study, analyses on the sodium experiments using SIMMER-III as the fast reactor safety evaluation code were performed. The analysis methods were validated by comparing the results and experiment data. In addition, the cooling and solidification behaviors during jet impingement were evaluated. The results indicated that the molten core material exhibited fragmentation owing to the impingement on the horizontal plate and was, therefore, scattered toward the periphery. Furthermore, the simulated molten core material was evaluated to be cooled by sodium and subsequently solidified.
Igarashi, Kai*; Onuki, Ryoji*; Sakai, Takaaki*; Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08
Kondo, Satoru; Tobita, Yoshiharu
JAEA-Research 2019-009, 382 Pages, 2020/03
JAEA-Research-2019-009.pdf:22.82MB
The SIMMER-III computer code, developed at the Japan Atomic Energy Agency (JAEA, the former Power Reactor and Nuclear Fuel Development Corporation), is a two-dimensional, multi-velocity-field, multi-component fluid-dynamics code, coupled with a space- and time-dependent neutron kinetics model. The code is being used widely for simulating complex phenomena during core-disruptive accidents (CDAs) in liquid-metal fast reactors (LMFRs). In parallel to the code development, a comprehensive assessment program was performed in two phases: Phase 1 for verifying individual fluid-dynamics models; and Phase 2 for validating its applicability to integral phenomena important to evaluating LMFR CDAs. The SIMMERIII assessment program was participated by European research and development organizations, and the achievement of Phase 1 was compiled and synthesized in 1996. This report has been edited by revising and reproducing the original 1996 informal report, which compiled the achievement of Phase 1 assessment. A total of 34 test problems were studied in the areas: fluid convection, interfacial area and momentum exchange, heat transfer, melting and freezing, and vaporization and condensation. The problems identified have been reflected to the Phase 2 assessment and later model development and improvement. Although the revisions were made in the light of knowledge base obtained later, the original individual contributions by the participants, both positive and negative, are retained except for editorial changes.
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.
C for severe accident simulationsLiu, X.*; Morita, Koji*; Yamano, Hidemasa
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 12 Pages, 2018/10
On the basis of experimental results, growth of the eutectic material is modeled by the parabolic rate law. Heat and mass transfer processes are also modeled considering both the equilibrium and non-equilibrium phase changes of eutectic material. Thermophysical properties of eutectic material obtained from the experimental measurements are also included in the analytic thermophysical property model and analytic equation-of-state model.
Onoda, Yuichi; Matsuba, Kenichi; Tobita, Yoshiharu; Suzuki, Toru
Mechanical Engineering Journal (Internet), 4(3), p.16-00597_1 - 16-00597_14, 2017/06
Toyooka, Junichi; Kamiyama, Kenji; Tobita, Yoshiharu; Suzuki, Toru
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11
Onoda, Yuichi; Matsuba, Kenichi; Tobita, Yoshiharu; Suzuki, Toru
Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 10 Pages, 2016/06
Guo, L.*; Morita, Koji*; Tobita, Yoshiharu
Journal of Nuclear Science and Technology, 53(2), p.271 - 280, 2016/02
Times Cited Count：7 Percentile：55.03(Nuclear Science & Technology)Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu
Annals of Nuclear Energy, 85, p.740 - 752, 2015/11
Times Cited Count：26 Percentile：90.65(Nuclear Science & Technology)Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 9 Pages, 2015/05
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：27 Percentile：91.4(Nuclear Science & Technology)Cheng, S.; Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu
Science and Technology of Nuclear Installations, 2015, p.964327_1 - 964327_14, 2015/00
Times Cited Count：6 Percentile：45.92(Nuclear Science & Technology)JNC TN9400 2000-038, 98 Pages, 2000/04
JNC-TN9400-2000-038.pdf:7.49MB
As an effort in the feasibility study on commercialized Fast Breeder Reactor cycle systems, an evaluation of the measures to prevent the energetic re-criticality in sodium-cooled large MOX core, which is one of the candidates for the commercialized reactor, has been performed. The core disruptive accident analysis of Demonstration FBR showed that the fuel compaction of the molten fuel by radial motion in a large molten core pool had a potential to drive the severe super-prompt re-criticality phenomena in ULOF sequence. ln order to prevent occurrence of the energetic re-criticality, a subassembly with an inner duct and the removal of a part of LAB are suggested based on CMR (Controlled Material Relocation) concept. The objective of this study is the comparison of the effectiveness of CMR among these measures by the analysis using SIMMER-III. The molten fuel in the subassembly with inner duct flows out faster than that from other measures. The subassembly with inner duct will work effectively in preventing energetic re-criticality. Though the molten fuel in the subassembly without a part of LAB flows out a little slower, it is still one of the promising measures. However, the UAB should be also removed from the same pin to prevent the fuel re-entries into the core region due to the pressurization by FCl below the core, unless it disturbs the core performance. The effect of the axial fuel length of the center pin to CMR behavior is small, compared to the effect of the existence of UAB.
Suzuki, Toru; Tobita, Yoshiharu
JNC TN9400 2000-019, 35 Pages, 2000/03
JNC-TN9400-2000-019.pdf:1.79MB
The transition phase analysis code SIMMER-III has been developed to appropriately evaluate the core disruptive accident in a fast breeder reactor. The momentum exchange model used in the fluid dynamics portion of the code uses the conventional correlation based on ordinary flows such as air-water flows. lt has already been confirmed that this code can represent the experimental results of ordinary flows. However, more detailed research is needed to confirm that this code is applicable to two-phase flow with large liquid density, which would be formed in an actual molten core pool. In addition, since the shapes of bubbles affect their drag in the bubbly flow where the liquid and the gas form continuous and dispersed phases, respectively, it is necessary to take this effect of bubble shape into account to improve SIMMER-III's analytical precision. ln this study, using experimental results obtained through a joint research program with Kyoto university, the momentum exchange model of SIMMER-III is assessed with regard to the bubbly flow regime of two-phase flow with large liquid density, on which experimental data and information on bubble shapes had been lacking. This study suggests that the original SIMMER-III can appropriately represent the characteristics of bubbly flows containing ellipsoidal bubbles with relatively small gas flux. Moreover, this study shows that the precision of SIMMER-III for bubbly flows containing cap bubbles with relatively large gas flux is much improved by using Kataoka-lshii's correlation to determine the drag coefficient of bubbles in the momentum exchange model.
Morita, Koji; Tobita, Yoshiharu; kondo, Satoru; E.A.Fischer*
JNC TN9400 2000-005, 57 Pages, 1999/05
JNC-TN9400-2000-005.pdf:2.92MB
An improved analytic equation-of-state (EOS) model using flexible thermodynamic functions is developed for a reactor safety analysis code, SIMMER-III. The present EOS model is designed to have adequate accuracy in describing thermodynamic properties of reactor-core materials over wide temperature and pressure ranges and to consistently satisfy basic thermodynamic relationships without deterioration of the computing efficiency. The fluid-dynamic algorithm for pressure iteration consistently coupled with the EOS model is also described in the present report. The EOS data of the basic core materials, uranium dioxide, mixed-oxide fuel, stainless steel, and sodium, are developed up to the critical point by compiling the most up-to-date and reliable sources using basic thermodynamic relationships. The thermodynamic consistency and accuracy of the evaluated EOS data are also discussed by comparison with the available sources.
Morita, Koji; Tobita, Yoshiharu; kondo, Satoru; E.A.Fischer*
JNC TN9400 2000-004, 38 Pages, 1999/05
JNC-TN9400-2000-004.pdf:1.11MB
An analytic thermophysical property model using general function forms is developed for a reactor safety analysis code, SIMMER-III. The function forms arc designed to represent correct behavior of properties of reactor-core materials over wide temperature ranges, especially for the thermal conductivity and the viscosity near the critical point. The most up-to-date and reliable sources for uranium dioxide, mixed-oxide fuel, stainless stee1, and sodium available at present are used to determine parameters in the proposed functions. This model is also designed to be consistent with a SIMMER-III model on thermodynamic properties and equations of state for reactor-corc materials.
