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Takamatsu, Kuniyoshi; Funatani, Shumpei*
Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 11 Pages, 2024/11
Our research objectives are to develop a VCS that utilizes radiative cooling to passively remove decay heat and residual heat from the RPV during expected and unexpected natural phenomena and accidents. To solve the back pressure problem around the inlet and outlet, it is necessary to minimize reliance on fluid actuation, such as water, air, etc., and to avoid using natural circulation or natural convection as much as possible to improve safety against external hazards. In this presentation, we present the structural concept of the proposed VCS integrated with the reactor building and report the results of the cooling performance evaluation based on the results of experimental and analytical studies conducted to date.
Hayakawa, Satoshi*; Hagiwara, Hiroyuki*; Imamura, Akira*; Onoda, Yuichi; Tanaka, Masaaki; Nakamura, Hironori*
Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 8 Pages, 2024/11
In a sodium-cooled fast reactor, a cover gas region filled with argon gas is located above the sodium pool in the main vessel to prevent the hot sodium from contacting the structures. This region involves heat transportation by natural convection of the cover gas, radiation among liquid surface and structures, and sodium phase change between mist and vapor. In this study, the numerical evaluation method has been developed with a commercial CFD code, Fluent, incorporating the sodium mist transport and growth models, and the radiation scattering model. Simulations of a laboratory scale test with a cylindrical cover gas region was carried out for the validation of the method and showed that the temperature distribution and sodium mist concentration in the cover gas region are in good agreements with the test results. A simulation of a pool-type sodium cooled fast reactor has also conducted and the basic aspect of physical phenomena taking place in the cover gas region were evaluated.
Hamase, Erina; Kawamura, Takumi*; Doda, Norihiro; Tanaka, Masaaki
Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 10 Pages, 2024/11
A plant dynamics analysis code, Super-COPD, is being developed for the design and safety evaluation of sodium-cooled fast reactors. Verification, validation, and uncertainty quantification (VVUQ) are required to ensure the reliability of its analysis results. In this study, to develop the VVUQ method, the uncertainty propagation analysis of input parameters was performed for the loss of flow without scram test in the FFTF, and the process of validation was investigated. In addition, the method of sensitivity analysis was investigated. As a result, the uncertainty of the analysis results was quantified, the applicability of the statistical method was confirmed. The sensitivity analysis using the Sobol' method identified the models that needs to be prioritized for improvement.
Yamasaki, Ryota; Aizawa, Kosuke; Kobayashi, Jun; Kurihara, Akikazu
Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 7 Pages, 2024/11
Emura, Yuki; Matsuba, Kenichi; Kikuchi, Shin; Yamano, Hidemasa
Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 8 Pages, 2024/11
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
Uesawa, Shinichiro; Ono, Ayako; Yamashita, Susumu; Yoshida, Hiroyuki
Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 7 Pages, 2024/11
A conductance-typed wire-mesh sensor (WMS), utilizing the difference in conductivity between gas and liquid phases between the electrodes, is one of the practical measurement techniques of a cross-sectional void fraction distribution in a flow path. In this study, we performed two-phase computational fluid dynamics (CFD) and electrostatic simulations around a WMS for a single spherical bubble and bubbly flow to clarify the systematic error in the WMS. The results for the single bubble indicated that there were systematic errors based on the non-uniform current density distribution around the WMS. The correlation between instantaneous void fractions and WMS signals is not uniquely determined for positions of the single bubble moving across the WMS, even for the same bubble. Moreover, the correlation between the instantaneous void fractions and the WMS signals did not fit in a linear approximation and Maxwell's equation, which traditionally used transformation methods from the WMS signal to the void fraction. The results for the bubbly flow indicated that the WMS had difficulty in quantitative measurements of the instantaneous void fraction because the values had a significant deviation of the void fraction of approximately 
0.2. On the other hand, time-averaged void fraction values had relatively small deviation. Thus, we concluded that the WMS, using existing transformation methods, can measure time-averaged void fractions, but it is difficult to measure quantitatively instantaneous void fractions.
Fukuda, Takanari
Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 10 Pages, 2024/11
Deepening the understanding of the molten core-concrete interaction (MCCI) is of the great importance for the sake of the severe accident managements as well as the fuel debris retrieval. Due to the difficulty to perform the experimental study with the extremely hot corium, the computational fluid dynamics (CFD) is expected to provide physical insights on the thermal-hydraulics taken place in the corium. The particle method are one of the CFDs that have advantages on seamless tracking of the multi-phase multi-component flow, typically involved in the MCCI. However, the adequacy of the modelling methods for the interfacial tension has not yet well investigated, especially for the general multi-phase flow with more than three phases. Hence, in this study, a simple liquid-liquid-gas three phase flow is analyzed with the existing two types of the interfacial tension models: the continuum surface force (CSF) model and the potential model. Through the comparison, it has been implied that the CSF model gives more accurate result with the satisfactory resolution, whereas the stability is strongly dependent on the resolution of the bulk fluid. On the other hand, the potential model outperforms in terms of the stability, presumably because it does not require the numerical estimation of the geometrical information. However the inter-particle potential force seems to induces locally unphysical pressure distribution, which can be especially detrimental on the multiple interface junctions.
Yamano, Hidemasa; Morita, Koji*
Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 9 Pages, 2024/11