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Uesawa, Shinichiro; Ono, Ayako; Nagatake, Taku; Yamashita, Susumu; Yoshida, Hiroyuki
Journal of Nuclear Science and Technology, 62(5), p.432 - 456, 2025/05
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)We performed electrostatic simulations of a wire-mesh sensor (WMS) for a single spherical bubble and bubbly flow to clarify the accuracy of the WMS. The electrostatic simulation for the single bubble showed the electric current density distribution and the electric current path from the excited transmitter to receivers for various bubble locations. It indicated systematic errors based on the nonuniform current density distribution around the WMS. The electrostatic simulation for the bubbly flow calculated by the computational fluid dynamics code, JAEA Utility Program for Interdisciplinary Thermal-hydraulics Engineering and Research (JUPITER), indicated that the WMS had difficulty in quantitatively measuring the intermediate values of the instantaneous void fraction between 0 and 1 because they cannot be estimated by previous transformation methods from the WMS signal to the void fraction, such as linear approximation or Maxwell's equation, and have a significant deviation of the void fraction of 
0.2 for the WMS signal. However, the electrostatic simulation indicated that the time-averaged void fractions around the center of the flow channel can be estimated using linear approximation, and the time-averaged void fraction near the wall of the flow channel can be estimated using Maxwell's equation.
Uesawa, Shinichiro; Ono, Ayako; Yoshida, Hiroyuki
Konsoryu, 39(1), p.61 - 71, 2025/03
Bubble visualization using a high-speed video-camera has been used as a measurement technique of bubble diameters and velocities. However, the bubble detection was difficult under the condition of the high void fraction because the overlapping bubbles for the sight direction of the camera increased with the void fraction. Additionally, the visualization for a system with objects, such as rod bundle flow channels, becomes more difficult. In this study, we applied a deep learning-based bubble detection technique with Shifted Window Transformer to bubble images shoot from two directions to identify the bubble size, three-dimensional (3D) positions of bubbles, 3D bubble trajectories in the rod bundle flow channel. Furthermore, we used perfluoroalkoxy alkane tubes with almost the same reflection as water in the channel to visualize the bubbly flow in the whole of the flow channel. We confirmed that the detection technique can segment individual bubbles in overlapping bubbles and bubbles behind the rod. By using the detection results, we estimated the diameter and velocity of each bubble and cross-sectional void fraction.
Hamase, Erina; Miyake, Yasuhiro*; Imai, Yasutomo*; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki
Nuclear Engineering and Design, 432, p.113738_1 - 113738_12, 2025/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)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, the reactor vessel model using a computational fluid dynamics code (RV-CFD) is required to be established. Previously, the CFD model based on the subchannel analysis was developed. In this study, to achieve lower computational cost maintaining the prediction accuracy, the coarse-mesh subchannel CFD (CMSC) model was 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 core-plenum interaction under NC conditions.
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.
Kamiya, Tomohiro; Nagatake, Taku; Ono, Ayako; Tada, Kenichi; Kondo, Ryoichi; Nagaya, Yasunobu; Yoshida, Hiroyuki
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 7 Pages, 2024/11
We have developed the JAEA Advances Multi-Physics Analysis platform for Nuclear systems (JAMPAN) to realize high-fidelity neutronics/thermal-hydraulics coupling simulations. We will perform MVP/JUPITER coupling simulation for a single BWR fuel assembly in order to confirm that the neutronics/thermal-hydraulics coupling through JAMPAN is feasible. This presentation explains how to send and receive data between MVP and JUPITER through JAMPAN and simulation results.
Uesawa, Shinichiro; Ono, Ayako; Yoshida, Hiroyuki
Konsoryu Shimpojiumu 2024 Koen Rombunshu (Internet), 2 Pages, 2024/09
Bubble visualization using a high-speed video-camera has been used as measurement techniques of bubble diameters, interfacial area concentrations, and void fractions in dispersed bubbly flow. However, the bubble detection was difficult under the condition of the high void fraction because the overlapping bubbles for the sight direction of the camera increased with the increase in the void fraction. In this study, we developed the deep learning-based bubble detector with Shifted window Transformer (Swin Transformer) to overcome the issue. To verify the performance, we used the synthetic bubble images obtained by Generative Adversarial Networks (GAN) and obtained average precisions (APs) for the number of the training dataset. The result showed that the AP was large enough for 50 datasets, and bubble detection was possible even with a small number of the training data. Additionally, we confirmed that the detector can detect and segment individual bubbles in overlapping bubbles obtained in the visualization experiments of pipe and bundle flows. By using the detection results, we estimated the interfacial area concentrations and void fractions. In comparison with commonly used relations, the results were in good agreement with the relations. Thus, the detector can measure not only bubble diameters but also interfacial area concentrations and void fractions.
Hamase, Erina; Miyake, Yasuhiro*; Imai, Yasutomo*; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki
Nihon Kikai Gakkai 2024-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2024/09
In a design study of sodium-cooled fast reactors, we have developed the practical reactor vessel thermal-hydraulic analysis method (RV-CFD) that had a low computational cost about the thermal-hydraulics in the core to evaluate the core-plenum interactions occurred in the natural circulation decay heat removal during the dipped-type direct heat exchanger operation. In this study, the non-equilibrium thermal model which considered the thermal inertia of fuel pins was developed and incorporated into the core of RV-CFD. Through the transient analysis simulating the power reduction due to reactor scram, the applicability of RV-CFD to the transient analysis was confirmed.
Hamase, Erina; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Imai, Yasutomo*
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NTHOS-14) (Internet), 12 Pages, 2024/08
We have been developed a thermal-hydraulic analysis model in the reactor vessel using the computational fluid dynamics code with a low computational cost to evaluate core-plenum interactions during a natural circulation decay heat removal using a dipped-type direct heat exchanger in a design of sodium-cooled fast reactors. In this study, we investigate the coarse mesh modeling of interwrapper gap (IWG) using correlations for the purpose of the development of a practical model which can reduce the computational cost maintaining the prediction accuracy. An influence of combinations of the coarse mesh and the correlation for pressure loss in the IWG on the thermal-hydraulics and the core temperature distribution is revealed through the numerical analysis of a sodium experiment.
Uesawa, Shinichiro; Ono, Ayako; Yoshida, Hiroyuki
Dai-52-Kai Kashika Joho Shimpojiumu Koen Rombunshu (Internet), 2 Pages, 2024/07
In order to obtain 3D behavior of bubbles, visualization using high-speed video-cameras has been used to identify 3D positions of bubbles. However, it was difficult to apply the technique to bubbly flow with the high void fraction because overlapping bubbles for the sight direction of the camera increased with the increase in the void fraction. JAEA has developed the deep learning-based bubble detector with Shifted window Transformer (Swin Transformer) to overcome the issue for the overlapping bubbles. In this study, we applied the bubble detection technique to images of bubble swarms visualized from two directions other than the direction of main flow and visualized 3D behavior of dispersed bubbles. The result showed that individual bubbles in bubble swarms were detected, and bubble diameters and aspect ratios were measured. Additionally, we obtained 3D positions of bubbles and 3D bubble velocities by linking the bubble positions for the direction of main flow in both images.
Ono, Ayako; Sakashita, Hiroto*; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
Mechanical Engineering Journal (Internet), 11(4), p.24-00188_1 - 24-00188_12, 2024/07
Japan Atomic Energy Agency (JAEA) is developing the evaluation method for a two-phase flow in the reactor core using simulation codes based on the Volume Of Fluid (VOF) method. JAEA started developing a Simplified Boiling Model (SBM) for the large-scale two-phase flow in the fuel assemblies. In the SBM, the motion and growth equations of the bubble are solved to obtain their diameter and time length at the detachment, of which size scale is within/around the calculation grid size of the numerical simulation. JUPITER calculates the bubble behavior with a scale of more than several 
m. In this study, the convection boiling on a vertical heating surface is simulated using the developed SBM. The comparison between the simulation and experimental results showed good reproducibility of the heat flux and velocity dependency on the passage period of the bubble.
Ono, Ayako; Okamoto, Kaoru*; Makino, Yasushi*; Hosokawa, Shigeo*; Yoshida, Hiroyuki
Proceedings of Specialist Workshop on Advanced Instrumentation and Measurement Techniques for Nuclear Reactor Thermal-Hydraulics and Severe Accidents (SWINTH-2024) (USB Flash Drive), 13 Pages, 2024/06
JAEA has been developing an advanced neutronic/thermal-hydraulics coupling simulation system. In the coupling simulation system, the detailed thermal-hydraulics codes based on an interface-capturing method (JUPITER or TPFIT) will be adopted to simulate thermal-hydraulics behavior in a fuel bundle. The experimental data and findings relating to the gas-liquid two-phase flow in a fuel bundle are especially required to validate JUPITER/TPFIT. In this study, we therefore develop a measurement method by combining Laser-Doppler Velocimetry (LDV) and photodiodes, which can access to a small flow channel such as a subchannel of a fuel bundle. The developed measurement method is validated by comparison with the measument by a electrical conductance probe. Finally, we obtain experimental data on local flow structures and interactions between gas and liquid phases. The developed measurement method is actually applied to an air-water dispersed bubbly flow to confirm its capability.
4 simulated fuel bundle for validation of thermal-hydraulics simulation codesOno, Ayako; Nagatake, Taku; Uesawa, Shinichiro; Shibata, Mitsuhiko; Yoshida, Hiroyuki
Proceedings of Specialist Workshop on Advanced Instrumentation and Measurement Techniques for Nuclear Reactor Thermal-Hydraulics and Severe Accidents (SWINTH-2024) (USB Flash Drive), 7 Pages, 2024/06
Japan Atomic Energy Agency (JAEA) is developing a neutronics/thermal-hydraulics coupling simulation code for light-water reactors. Thermal-hydraulic simulation codes applied to the coupling code are expected to calculate the void fraction distribution in a rod bundle under operational conditions, which are necessary for neutron transport simulation, and need to be validated using void fraction distribution data in a rod bundle under high-temperature and high-pressure conditions. Therefore, we have conducted the measurement of the instantaneous void distribution in the 4 4 simulated fuel bundle using a developed wire mesh sensor, which is installed in the pressurized two-phase flow experimental loop of JAEA to obtain the data for code validation.
Tada, Kenichi; Kondo, Ryoichi; Kamiya, Tomohiro; Nagatake, Taku; Ono, Ayako; Nagaya, Yasunobu; Yoshida, Hiroyuki
Proceedings of International Conference on Physics of Reactors (PHYSOR 2024) (Internet), p.1488 - 1497, 2024/04
JAEA has developed a new high-fidelity multi-physics platform JAMPAN for connecting single-physics codes such as a neutronics code and a thermal-hydraulics code. It consists of the HDF5 formatted data container and input and output data handler modules to generate the input file and read the output file of the single-physics code. Users can easily add or exchange the code by implementing input and output data handler modules for this code. The first target of JAMPAN is the coupling of neutronics and thermal-hydraulics calculations to provide reference results of core analysis codes. The current version of JAMPAN couples the neutronics code MVP and the thermal-hydraulics codes JUPITER, ACE-3D, and NASCA. Users can select the thermal-hydraulics code depending on the scale of problems to be solved, computational performance, and so on. This presentation explains the overview of JAMPAN and shows the results of the neutronics and thermal-hydraulics coupling calculation.
Katsuta, Nagayoshi*; Umemura, Ayako*; Naito, Sayuri*; Masuki, Yuma*; Itayama, Yui*; Niwa, Masakazu; Shirono, Shinichi*; Yoshida, Hidekazu*; Kawakami, Shinichi*
Spectrochimica Acta, Part B, 210, p.106817_1 - 106817_11, 2023/12
Times Cited Count：1 Percentile：21.21(Spectroscopy)X-ray fluorescence (XRF) scanning of lacustrine sediments has been used to extend the approach to a wider range of elemental records in both ages and timescales of variations in past environments and climates. However, one of severe problems with effects of the XRF intensity by grain size and mineralogical composition known as "heterogeneity effects" have been pointed out. This study investigated the heterogeneity effect of Fe intensities on X-ray beam using several binary powder mixtures and lacustrine sediment cores.
Koyama, Shinichi; Ikeuchi, Hirotomo; Mitsugi, Takeshi; Maeda, Koji; Sasaki, Shinji; Onishi, Takashi; Tsai, T.-H.; Takano, Masahide; Fukaya, Hiroyuki; Nakamura, Satoshi; et al.
Hairo, Osensui, Shorisui Taisaku Jigyo Jimukyoku Homu Peji (Internet), 216 Pages, 2023/11
In FY 2021 and 2022, JAEA perfomed the subsidy program for "the Project of Decommissioning and Contaminated Water Management (Development of Analysis and Estimation Technology for Characterization of Fuel Debris (Development of Technologies for Enhanced Analysis Accuracy, Thermal Bahavior Estimation, and Simplified Analysis of Fuel Debris)" started in FY 2021. This presentation material summarized the results of the project, which will be available shortly on the website of Management Office for the Project of Decommissioning, Contaminated Water and Treated Water Management.
4 simulated fuel bundle for validation of thermal-hydraulics simulation codesNagatake, Taku; Shibata, Mitsuhiko; Uesawa, Shinichiro; Ono, Ayako; Yoshida, Hiroyuki
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
JAEA is developing a neutronics/thermal-hydraulics coupling simulation code for light-water reactors. Thermal-hydraulic simulation codes applied to the platform are expected to evaluate void fraction distributions in fuel assemblies under operational conditions, which is necessary for neutron transport simulation, and need to be validated using void fraction distribution data in a rod bundle under high-temperature and high-pressure conditions. To obtain the data for code validation, we have been measuring the instantaneous void fraction distribution in a 44 simulated fuel assembly by a wire mesh sensor. In this paper, we report the results of the experiments with pressure and flow rate as parameters at a maximum pressure of 2.6 MPa.
Hamase, Erina; Miyake, Yasuhiro*; Imai, Yasutomo*; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki
Nihon Kikai Gakkai 2023-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2023/09
In sodium-cooled fast reactors, decay heat removal systems under natural circulation with a dipped-type direct heat exchanger (D-DHX) have been investigated. During the D-DHX operation, the cold sodium from the D-DHX flows into the assemblies and the interwrapper gap (IWG) between them. To evaluate such phenomena in design studies, the reactor vessel thermal-hydraulic analysis method (RV-CFD) which has the accuracy required for design studies while reducing the computational cost is required. In this study, with the aim of developing the practical RV-CFD with a low computational cost, the influence of the combination of the mesh number in the IWG and the pressure loss coefficient on the core temperature distribution was investigated through the numerical analysis of a sodium experimental apparatus named PLANDTL-1.
Yamashita, Susumu; Uesawa, Shinichiro; Ono, Ayako; Yoshida, Hiroyuki
Mechanical Engineering Journal (Internet), 10(4), p.22-00485_1 - 22-00485_25, 2023/08
A detailed evaluation for air cooling of fuel debris in actual reactors will be essential in fuel debris retrieval under dry conditions. To understand the heat transfer in and around fuel debris, which is assumed as a porous medium in the primary containment vessel (PCV) mechanistically, we newly applied the porous medium model to the multiphase and multicomponent computational fluid dynamics code named JUPITER (JAEA Utility Program for Interdisciplinary Thermal-hydraulics Engineering and Research). We applied the Darcy-Brinkman model as for the porous medium model. This model has high compatibility with JUPITER because it can treat both a pure fluid and a porous medium phase simultaneously in the same manner as the one-fluid model in multiphase flow simulation. We addressed the case of natural convection with a high-velocity flow standing out nonlinear effects by implementing the Forchheimer model, including the term of the square of the velocity as a nonlinear effect to the momentum transport equation of JUPITER. We performed some simple verification and validation simulations, such as the natural convection simulation in a square cavity and the natural convective heat transfer experiment with the porous medium, to confirm the validity of the implemented model. We confirmed that the result of JUPITER agreed well with these simulations and experiments. In addition, as an application of the updated JUPITER, we performed the preliminary simulation of air cooling of fuel debris in the condition of the Fukushima Daiichi Nuclear Power Station unit 2 including the actual core materials. As a result, JUPITER calculated the temperature and velocity field stably in and around the fuel debris inside the PCV. Therefore, JUPITER has the potential to estimate the detailed and accurate thermal-hydraulics behaviors of fuel debris.
Kamiya, Tomohiro; Ono, Ayako; Tada, Kenichi; Akie, Hiroshi; Nagaya, Yasunobu; Yoshida, Hiroyuki; Kawanishi, Tomohiro
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 8 Pages, 2022/11
JAEA started to develop the advanced reactor analysis code JAMPAN (JAEA advanced multi-physics analysis platform for nuclear systems). The current version of JAMPAN handles the continuous energy Monte Carlo code MVP and the detailed thermal-hydraulics analysis code for multiphase and multicomponent JUPITER. JAMPAN is designed to consider the extensibility and it does not depend on the analysis codes. All calculations in JAMAPAN are not directly connected. JAMPAN has data containers, and all input and output data of each analysis code are set in these data containers. JAMPAN will easily exchange the calculation codes and add the other calculations, e.g., structure calculation and irradiation calculation since the input and the output format of each code has no impact on the other calculation codes. The 4 by 4 pin-cell geometry was used as the demonstration calculation of JAMPAN and the physically reasonable calculation results were obtained.
Yamashita, Susumu; Uesawa, Shinichiro; Ono, Ayako; Yoshida, Hiroyuki
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 8 Pages, 2022/10
no abstracts in English
