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Ono, Ayako; Yamashita, Susumu; Sakashita, Hiroto*; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 12 Pages, 2022/09
Japan Atomic Energy Agency is developing the computational fluid dynamics code, JUPITER, based on the volume of fluid (VOF) method to analyze detailed thermal-hydraulics in a reactor. The detailed numerical simulation of boiling from a heating surface needs a substantial computational cost to resolve the microscale thermal-hydraulic phenomena such as the bubble generation from a cavity and evaporation of a micro-layer. This study developed the simplified boiling model from the heating surface to reduce the computational cost, which will apply to the detailed simulation code based on the surface tracking method such as JUPITER. We applied the simplified boiling model to JUPITER, and compared the simulation results with the experimental data of the vertical heating surface in the forced convection. We confirmed the degree of their reproducibility, and the issues to be modified were extracted.
Yamano, Hidemasa; Takai, Toshihide; Emura, Yuki; Fukuyama, Hiroyuki*; Higashi, Hideo*; Nishi, Tsuyoshi*; Ota, Hiromichi*; Morita, Koji*; Nakamura, Kinya*; Fukai, Hirofumi*; et al.
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 12 Pages, 2022/09
This paper describes the project overview and progress of experimental and analytical studies conducted until 2020. Specific results in this paper are the measurement of the eutectic reaction rates and the validation of physical model describing the eutectic reaction in the analysis code through the numerical analysis of the B
C-SS eutectic reaction rate experiments in which a BC pellet was placed in a SS crucible.
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.
Hamase, Erina; Miyake, Yasuhiro*; Imai, Yasutomo*; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 12 Pages, 2022/09
To enhance the safety of sodium-cooled fast reactors, the natural circulation (NC) decay heat removal systems with a dipped-type direct heat exchanger (D-DHX) have been investigated. During the D-DHX operation, since the core-plenum interaction occurs, development of the reactor vessel model including the more model by using a computational fluid dynamics code (RV-CFD) is required. Previously, the CFD model based on the subchannel analysis was developed. In this study, to achieve much lower computational cost maintaining the prediction accuracy, the coarse-mesh subchannel CFD (CMSC) model has been developed and was incorporated into the core of RV-CFD. As a result of PLANDTL-1 test analysis, the RV-CFD with the CMSC model can reproduce the radial heat transfer under NC conditions.