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Sugihara, Kenta; Onodera, Naoyuki*; Idomura, Yasuhiro; Sitompul, Y.; Yamashita, Susumu
Journal of Computational Physics, 547, p.114534_1 - 114534_25, 2026/02
Times Cited Count：0 Percentile：0.00(Computer Science, Interdisciplinary Applications)Fukuda, Takanari; Yamashita, Susumu; Yoshida, Hiroyuki
Journal of Computational Physics, 545, p.114485_1 - 114485_32, 2026/01
Times Cited Count：0 Percentile：0.00(Computer Science, Interdisciplinary Applications)This paper puts forward a novel approach for the evaluation of the geometrical fidelity and the interface sharpness of the VOF advection schemes separately and quantitatively. This new evaluation has elucidated the trade-off relationship of the geometrical fidelity and the interface sharpness between the existing schemes of the original THINC and the THINC/WLIC. By investigating and resolving this trade-off relationship, we have developed a novel THINC-based scheme that exhibits high performance with regard to both geometrical fidelity and interface sharpness, despite employing an algorithm as concise as those of the original THINC and the THINC/WLIC. The novel scheme, designated "THINC/Advanced WLIC (THINC/AWLIC)," has been developed by redefining the weight function of the preceding THINC/WLIC so that the contribution of the first-order upwind flux can be variably blended with the usage of the control parameter. The results of the multiple benchmark tests in two and three dimensions demonstrate that both the geometrical fidelity and the interface sharpness are significantly enhanced if the control parameter is appropriately determined. Furthermore, the associated error of THINC/AWLIC is comparable to that of the geometrical scheme, although the implementation complexity is unchanged from that of the simple THICN/WLIC.
Fukuda, Takanari; Yamashita, Susumu; Yoshida, Hiroyuki
Journal of Nuclear Science and Technology, 62(12), p.1264 - 1278, 2025/12
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)This study compared three interface capturing schemes (ICSs) for multi-phase flow simulations based on the VOF method, focusing on bubble volume conservation. The THINC/WLIC scheme showed significant VOF diffusion and underestimated total bubble volume, while the original THINC and PLIC conserved bubble volumes. Moreover, an analysis of THINC/WLIC based on a new visualization approach revealed that VOF fragments were ripped off by shear forces around interface, making it unsuitable for accurate void fraction prediction in boiling water reactors. The original THINC may be a viable alternative to PLIC due to its simplicity.
Fukuda, Takanari; Uesawa, Shinichiro; Yamashita, Susumu
Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 12 Pages, 2025/09
A comparative study was conducted on three interface capturing schemes (ICSs) of the VOF method: THINC/WLIC, THINC/AWLIC, and PLIC for simulating gas-liquid two-phase flow in a BWR reactor core. The simulations in a rod bundle geometry were compared qualitatively and quantitatively with experimental data obtained with a high-speed camera and wire mesh sensors. The results showed that the all ICSs yielded reasonable agreements with experimental data, but THINC/WLIC had a significant issue in which the VOF value diffuses and dissipates over the simulation geometry. THINC/AWLIC, developed by the authors, improved the VOF diffusion issue of the THINC/WLIC and predicted the void fraction close to that of the highly accurate ICS of PLIC, despite its simpler algorithm. However, the numerical bubble coalescence was still an issue, particularly at low gas flow rates, which calls for further research to refine the simulation results to better reflect actual bubble behavior.
Yamashita, Susumu
JAEA-Data/Code 2025-003, 262 Pages, 2025/07
JAEA-Data-Code-2025-003.pdf:9.4MB
A multi-phase, multi-component, detailed thermal-hydraulic code JAEA Utility Program for Interdisciplinary Thermal-hydraulics Engineering and Research (JUPITER) has been developed to simulate the thermal-hydraulic behavior in nuclear reactors under steady-state and severe accident conditions in a mechanism-based manner. JUPITER can faithfully reproduce thermal hydraulics based on the governing equations. In addition, eutectic reactions at dissimilar metal contact surfaces and oxidation reactions between steam and zirconium alloys, which are important phenomena in severe accidents, can be analyzed. It is also applicable to porous media flow, which is often used for flow phenomena in fine particles. In this report, the governing equations and physical models of JUPITER and its numerical methods are outlined, and an input manual for JUPITER is presented as an appendix.
Yamashita, Susumu
JAEA-Data/Code 2025-002, 243 Pages, 2025/07
JAEA-Data-Code-2025-002.pdf:9.37MB
A multi-phase, multi-component, detailed thermal-hydraulic code JAEA Utility Program for Interdisciplinary Thermal-hydraulics Engineering and Research (JUPITER) has been developed to simulate the thermal-hydraulic behavior in nuclear reactors under steady-state and severe accident conditions in a mechanism-based manner. JUPITER can faithfully reproduce thermal hydraulics based on the governing equations. In addition, eutectic reactions at dissimilar metal contact surfaces and oxidation reactions between steam and zirconium alloys, which are important phenomena in severe accidents, can be analyzed. It is also applicable to porous media flow, which is often used for flow phenomena in fine particles. In this report, the governing equations and physical models of JUPITER and its numerical methods are outlined, and an input manual for JUPITER is presented as an appendix.
Uesawa, Shinichiro; Yamashita, Susumu; Sano, Yoshihiko*; Yoshida, Hiroyuki
Journal of Nuclear Science and Technology, 62(6), p.523 - 541, 2025/06
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Japan Atomic Energy Agency (JAEA) has developed a numerical method with the JUPITER code with a porous medium model to calculate the thermal behavior in PCVs of 1F. In this study, we performed an experiment and numerical simulation of the natural convective heat transfer with the porous medium to validate JUPITER with the porous medium model. In comparison of the temperature and velocity distributions between the experiment and simulation, the temperature distribution in the simulation was in good agreement with the distribution in the experiment except the temperature near the top surface of the porous medium. The velocity distribution also agreed qualitatively with the experimental result. In addition, we also performed the numerical simulations with various effective thermal conductivity models to discuss the effect of the conductivity based on the internal structure of porous media on the natural convective heat transfer. The result indicated that the temperature distribution in the porous medium and the velocity distribution of the natural convection were significantly different for each model, and thus the conductivity of the fuel debris was one of the key parameters of in the thermal behavior analysis in 1F.
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.
Zhou, W.*; Miwa, Shuichiro*; Yamashita, Susumu; Okamoto, Koji*
Progress in Nuclear Energy, 177, p.105441_1 - 105441_17, 2024/12
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Understanding air entrainment phenomena induced by plunging water jets is critical in the fields of nuclear and hydraulic engineering. Air entrainment is one of the key safety design parameters for nuclear systems. However, most existing studies rely on empirical correlations or curve-fitting models to estimate bubble penetration depth, and no agreed-upon calculation principle exists for varying jet conditions. To address these limitations, this research developed two advanced AI approaches: an improved YOLOv5 for segmenting air entrainment and the NSGA-III-BPNN method for predicting penetration depth. The improved YOLOv5 enables real-time segmentation and extraction of air entrainment motion and dynamics under diverse conditions, demonstrating high scalability and robustness. The penetration depth estimated using the improved YOLOv5 shows greater accuracy compared to conventional empirical correlations and is more efficient than traditional image post-processing techniques for classifying shape regimes based on dynamic air entrainment patterns. To overcome the limitations of object segmentation, which typically relies on video or image data, the NSGA-III-BPNN method predicts maximum penetration depths with greater accuracy than YOLOv5, offering a more effective prediction model for air entrainment penetration depth. By leveraging advanced AI techniques, the research not only provides valuable segmentation data for refining computational fluid dynamics (CFD) modeling but also paves the way for significant advancements in both nuclear and hydraulic engineering.
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.
Sugihara, Kenta; Onodera, Naoyuki; Sitompul, Y.; Idomura, Yasuhiro; Yamashita, Susumu
EPJ Web of Conferences, 302, p.03002_1 - 03002_10, 2024/10
Times Cited Count：3 Percentile：98.86(Computer Science, Interdisciplinary Applications)In simulations of gas-liquid two-phase flows using conventional interface capture methods, we observed that when bubbles come close to each other, they tend to merge numerically, despite experimental evidence indicating that they would repel each other. Given the significant impact of sequential numerical coalescence on flow patterns, it is necessary to regulate the merging behavior of close bubbles. To address this issue, we introduced the Multi- Phase Field (MPF) method, which mitigates bubble coalescence by applying an independent fluid fraction function to each bubble. In this study, we employed the MPF based on the N-phase model to minimize numerical errors associated with surface interactions at triple junction points. Additionally, we implemented the Ordered Active Parameter Tracking (OAPT) method to efficiently store several hundreds of fluid fraction functions. To validate the MPF method, we conducted analysis of turbulent bubbly pipe flows and compared the results against experimental data from Colin et al. The validation results showed reasonable agreements with respect to the bubble distribution and the flow velocity profiles.
Machida, Masahiko; Yamada, Susumu; Kim, M.; Tanaka, Satoshi*; Tobita, Yasuhiro*; Iwata, Ayako*; Aoki, Yuto; Aoki, Kazuhisa; Yanagisawa, Kenichi*; Yamaguchi, Takashi; et al.
RIST News, (70), p.3 - 22, 2024/09
Inside the Fukushima Daiichi Nuclear Power Plant (1F), there are many locations with high radiation levels due to contamination by radioactive materials that leaked from the reactor. These pose a significant obstacle to the smooth progress of decommissioning work. To help solve this issue, the Japan Atomic Energy Agency (JAEA), under a subsidy from the Ministry of Economy, Trade, and Industry's decommissioning and contaminated water management project, is conducting research and development on digital technologies to improve the radiation environment inside the decommissioning site. This project, titled "Development of Technology to Improve the Environment Inside Reactor Buildings (Enhancing Digital Technology for Environment and Source Distribution to Reduce Radiation Exposure)," began in April of FY 2023. In this project, the aim is to develop three interconnected systems: FrontEnd, Pro, and BackEnd. The FrontEnd system, based on the previously developed 3D-ADRES-Indoor (prototype) from FY 2021-2022, will be upgraded to a high-speed digital twin technology usable on-site. The Pro system will carry out detailed analysis in rooms such as the new office building at 1F, while the BackEnd system will serve as a database to centrally manage the collected and analyzed data. This report focuses on the FrontEnd system, which will be used on-site. After point cloud measurement, the system will quickly create a 3D mesh model, estimate the radiation source from dose rate measurements, and refine the position and intensity of the estimated source using recalculation techniques (re-observation instructions and re-estimation). The results of verification tests conducted on Unit 5 are also presented. Furthermore, the report briefly discusses the future research and development plans for this project.
Fukuda, Takanari; Yamashita, Susumu; Yoshida, Hiroyuki
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NUTHOS-14) (Internet), 12 Pages, 2024/08
The VOF method is a type of CFDs that is most widely applied to multiphase flow analysis involving advective interfaces, and several interface-capturing schemes have been developed for an accurate advection of VOF values. However, the performance of these schemes has typically been evaluated only for limited numerical problems where velocity fields are spatially orderly and fixed in time. Few studies have been conducted to evaluate the performance of these schemes for more realistic and complex conditions, such as gas-liquid two-phase flows in nuclear reactors. Therefore, in this study, three-dimensional analysis of bubble flows has been conducted using the interface-capturing schemes of THINC and THINC/WLIC, which have been developed relatively recently. Evaluation is performed using more engineering indicators such as the number, volume, and trajectory of bubbles, which can influence the void fraction distribution in reactor cores. The results of these comparisons showed that the VOF value could be significantly diffused, leading to numerical brake-up and dissipation of the bubbles, with the influence of interface-capturing scheme.
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.
Yamashita, Susumu; Kondo, Nao; Sugawara, Takanori; Monji, Hideaki*; Yoshida, Hiroyuki
Journal of Nuclear Science and Technology, 61(6), p.740 - 761, 2024/06
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)To confirm the validity of the thermal-hydraulics design tool based on the Ansys Fluent, we used a detailed computational fluid dynamics code named JAEA Utility Program for Interdisciplinary Thermal-hydraulics Engineering and Research (JUPITER) for the thermal-hydraulics around the beam window (BW) of the Accelerator-Driven System (ADS). The Fluent uses the Reynolds-Averaged Navier-Stokes (RANS) model and can quickly calculate the turbulent flow around the BW as a BW design tool. At first, we compared the results of JUPITER with the experimental results using a mock-up BW system in water to confirm the validity of JUPITER. As a result, we confirmed that numerical results are in good agreement with the experimental results. Thus, we showed that JUPITER could be used as a benchmark code. We also performed a benchmark simulation for the Fluent calculation using validated JUPITER to show the applicability of JUPITER as an alternative of experiments. As a result, the mean values around the BW agreed with each other, e.g., the mean velocity profile for stream and horizontal directions. Therefore, we confirmed that JUPITER showed a good performance in validating the thermal-hydraulics design tool as a fluid dynamics solver. Moreover, Fluent has enough accuracy as a thermal-hydraulics design tool for the ADS.
Sugihara, Kenta; Onodera, Naoyuki; Sitompul, Y.; Idomura, Yasuhiro; Yamashita, Susumu
Keisan Kogaku Koenkai Rombunshu (CD-ROM), 29, 6 Pages, 2024/06
When calculating a flow containing many bubbles, it is difficult to reproduce the flow pattern observed in experiments because bubbles that are close to each other merge numerically in gas-liquid two-phase flow analysis using the conventional interface capturing method. In this study, the Multi-Phase Field method is used to prevent the numerical bubble merging by representing each bubble using an independent fluid fraction function. The applicability of the proposed method is verified by comparing bubbly flow analysis with experimental results.
Sano, Naruto; Yamashita, Naoki; Watanabe, Masaya; Tsukada, Manabu*; Hoshino, Kazutoyo*; Hirai, Koki; Ikegami, Yuta*; Tashiro, Shinsuke; Yoshida, Ryoichiro; Hatakeyama, Yuichi; et al.
JAEA-Technology 2023-029, 36 Pages, 2024/03
JAEA-Technology-2023-029.pdf:2.47MB
At the Waste Safety Testing Facility (WASTEF), the gamma ray irradiation device "Gamma Cell 220" was relocated from the 4th Research Building of the Nuclear Science Research Institute in FY2019, and the use of gamma ray irradiation has begun. Initially, Fuel Cycle Safety Research Group, Fuel Cycle Safety Research Division, Nuclear Safety Research Center, Sector of Nuclear Safety Research and Emergency Preparedness, the owner of this device, conducted the tests as the main user, but since 2022, other users, including those outside the organization, have started using it. The gamma ray irradiation device "Gamma Cell 220" is manufactured by Nordion International Inc. in Canada. Since it was purchased in 1989, the built-in 
Co radiation source has been updated once, and safety research related to nuclear fuel cycles, etc. It is still used for this purpose to this day. This report summarizes the equipment overview of the gamma ray irradiation device "Gamma Cell 220", its permits and licenses at WASTEF, usage status, maintenance and inspection, and future prospects.
Yamashita, Naoki; Aoyama, Takahito; Kato, Chiaki; Sano, Naruto; Tagami, Susumu
JAEA-Technology 2023-028, 22 Pages, 2024/03
JAEA-Technology-2023-028.pdf:1.9MB
At the Fukushima Daiichi Nuclear Power Station (1F), which is currently undergoing decommissioning, there is growing interest in the effects of radiation-emitting radionuclides such as 
Sr and 
Cs on the structural integrity. In particular, the corrosion behavior of carbon steel, which is used in many parts of 1F, is known to change depending on metal cations in solution, but the effects of Sr and Cs on corrosion are not yet understood. In addition, it is important to investigate the distribution of Sr and Cs in the rust layer in order to understand the corrosion behavior, but the method has not yet been established. In this study, a glove box was prepared to conduct corrosion tests of carbon steel in NaCl containing Sr and Cs in the glove box. In addition, in order to clarify the influence of Sr and Cs, which exist as metal cations in the solution, on the corrosion behavior of carbon steel, we attempted to establish a detection method for radioactive materials in the rust layer using an imaging plate.
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 Behavior Estimation, and Abbreviated Analysis))" 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.
Sugihara, Kenta; Onodera, Naoyuki; Idomura, Yasuhiro; Yamashita, Susumu
JAEA-Research 2023-006, 47 Pages, 2023/10
JAEA-Research-2023-006.pdf:3.28MB
This report presents a new surface capturing method based on the phase field model for gas-liquid two-phase flows simulation. In the conventional phase field model, the interface correction strength parameter was determined from the maximum flow velocity in the computational domain, but because the interface correction was applied uniformly to the entire space, it was also applied to locations that did not require correction. In the new method, the phase field parameter or the intensity of the phase field model is extended to have a spatial distribution, allowing us to set the optimal parameters depending on the local flow velocity fields. We also propose a method to derive the optimal phase field parameter based on systematic parameter scans using error analysis of the interface advection test and bubble rising calculations. Through benchmark tests of gas-liquid two-phase flows, the proposed model is verified, and it is shown that the proposed model has higher accuracy than the conventional phase field model.
