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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.
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
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
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.
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
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.
Odaira, Naoya*; Ito, Daisuke*; Ito, Kei*; Saito, Yasushi*; Imaizumi, Yuya; Matsuba, Kenichi
no journal, ,
Horiguchi, Naoki
no journal, ,
In severe accidents of LWRs, it is expected that radioactive micro-particles are captured by the interfaces of bubbles in a pool, and the capture controls the discharge of radioactive micro-particles to the exterior environment. In recent research on the visualization of bubbles, which include insoluble micro-particles, the accumulation of particles on the interfaces of bubbles is reported. However, numerical studies for severe accident management commonly ignore this accumulation and treat a single particle to move from a gas phase to a liquid phase irreversibly. Therefore, experimental studies for the visualization of the single micro-particle behavior on an interface is required in order to improve severe accident management. In the experiments for this purpose, the manipulation of a micro-particle near an interface in a non-contact manner is required too. Although laser manipulation techniques can not only manipulate a micro-particle in a single liquid phase in a non-contact manner but also visualize its behavior, there is no report related to a micro-particle on an interface. In this study, we developed a visualization method of single micro-particle behavior on an interface using a laser manipulating technique. We prepared an apparatus to manipulate and visualize an insoluble micro-particle in a single water phase using a laser manipulating technique; we developed a method to form an interface using water and silicon oil in the apparatus. We conducted the experiment using the apparatus and confirmed that a micro particle was captured by the interface. Another micro-particle was captured by an interface with small curvature but returned to the original phase after a short time. We considered that the Brownian motion of the particle itself and the curvature of the interface affect the behavior.
Masaki, Naoto*; Kaneko, Akiko*; Horiguchi, Naoki; Yoshida, Hiroyuki
no journal, ,
During a severe accident in nuclear reactors, melted fuel rods and other materials may fall as liquid jets into the pool of cooling water remaining in the lower head of the pressure vessel. From the viewpoint of preventing accident development, the clarification of cooling and solidification mechanisms with atomization in the cooling water is essential. Many researchers performed experimental and analytical work to understand jet behavior and atomization in a deep pool. In the case of insufficient water depth, i.e., shallow water pool, there are a few reports in both experimental and analytical approaches. The reports of shallow pools show that atomization occurs even after the jet has impacted the bottom wall. However, the mechanism has not been understood. In this study, we assumed that instability due to velocity difference affects the atomization in a shallow-water pool and tried to give a velocity difference to the interface as an agitation using a vortex provided by a stirrer to examine whether the agitation affects atomization. We conducted visualization and measurement experiments using a simulated liquid-liquid two-fluid system in the shallow pool. As a result, an agitation using a vortex occurred in the pool. The number of droplets and the rise time of the number changed increasing the stirrer frequency. This result suggests that the agitation contributed to the atomization.
Ito, Kei*; Ezure, Toshiki; Odaira, Naoya*; Ito, Daisuke*; Saito, Yasushi*
no journal, ,
In this study, the authors proposed a phase-change model in which the simulation accuracy and the stability are enhanced by taking into account the precise physical mechanism of phase-change process at gas-liquid interface. Then, the calculation method of temperature gradient near an interface, which determines the calculation of condensation quantity, is examined to enhance the simulation accuracy. Furthermore, the superiority of developed method to conventional methods is confirmed by simulating the basic condensation problem and comparing the simulation results.
Yamano, Hidemasa; Morita, Koji*
no journal, ,
Ota, Yosei*; Kanda, Yuna*; Hisamochi, Rikuya*; Yada, Hiroki; Furuya, Masahiro*
no journal, ,
Yamashita, Susumu; Yoshida, Hiroyuki
no journal, ,
The JAEA has developed a detailed thermal-hydraulic analysis code JUPITER based on mechanistic fluid flow simulation methodology to obtain melt relocation behavior instead of fuel assembly melt experiments. JUPITER has not only the capability to mechanistically analyze the relocation behavior and heat transport of melts with free interfaces, but also a eutectic and an oxidation reaction model, which have a significant impact on the accident progression. Although simulations have been performed for each model, simulations under the composite condition of these phenomena have not yet been performed. In order to confirm the feasibility of JUPITER for the experimental analysis of the melt relocation behavior under the composite condition, we applied JUPITER to the experimental analyses: XR2-1 experiment in which the melt conditions inside the mock-up fuel assembly of the BWR were obtained, and the LEISAN experiment in which the melting behavior of control rods due to steam concentration was observed. In the XR2-1 experimental analysis, the stainless steel/boron carbide (B
C) and the Zircaloy melts were poured into the mock-up fuel assembly from the top boundary. Then, as in the experiment, several discharge paths of their melts were confirmed. In addition, the failure to the fuel assembly due to contact with the melts was also confirmed. In the LEISAN experiment analysis, the stainless steel control rod blade including the BC absorber was heated. As a result, the control rod melted at a temperature lower than the melting point of the stainless steel/BC due to the eutectic reaction between them.
Horiguchi, Naoki; Yoshida, Hiroyuki; Kitatsuji, Yoshihiro; Fukumori, Mai*; Hasegawa, Makoto*; Kishimoto, Tadafumi*
no journal, ,
A Li-7 enriched pH adjuster is essential for water quality control on PWRs. As the required Li-7 enrichment technology, we have developed the multi-channel counter-current electrophoresis (MCCCE) method. Here, Li-7 enrichment using the MCCCE method is to separate Li-7 and Li-6 ions moving in an aqueous solution by different velocities through electric and flow fields. As a part of the development, we have developed a numerical simulation method to understand the behavior of ions. In the numerical method, we calculate electric and flow fields using finite volume methods and the movement of an ion as a mass particle to reproduce the behavior of Li ions. Comparison of the numerical simulation with the experiment is required to validate the numerical simulation, but it is impossible to track ions individually and evaluate their behavior in the experiment. Then, we focused on separation coefficient estimated in the experiments. In this work, we applied the numerical method to the behavior of ions in a simulated channel and estimated the separation coefficient for validation. The separation coefficient is generally estimated using concentration ratios of Li-7 and Li-6, but we estimated the separation coefficient using the ratios of the numbers of ions, which equal to the concentration ratios. Then, we conducted a sensitivity analysis of the separation coefficient by changing the applied voltage and reproduced the change trend of the experiment.