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Fukuda, Takanari; Yamashita, Susumu; Yoshida, Hiroyuki
Journal of Computational Physics, 545, p.114485_1 - 114485_32, 2026/01
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
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.
Kamiya, Tomohiro; Yoshida, Hiroyuki
Physics of Fluids, 37(10), p.103359_1 - 103359_23, 2025/10
In this study, we developed a conservative scheme based on a volume of fluid (VOF) and a ghost fluid method for liquid-gas two-phase compressible fluid simulations. We treated several one- and two-dimensional numerical problems to investigate the capability and applicability of the proposed method for compressible two-phase fluid simulations. The results agree well with the exact solutions or the numerical results of previous studies. Furthermore, the results also show that the proposed method can almost completely ensure the conservation property. Consequently, we concluded that the proposed method could simulate compressible two-phase flows and conserve mass, momentum, and total energy.
Ito, Kei*; Matsushita, Kentaro; Ezure, Toshiki; Tanaka, Masaaki; Odaira, Naoya*; Ito, Daisuke*; Saito, Yasushi*
Nihon Kikai Gakkai 2025-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2025/09
The estimation of entrained gas flow rate by a bathtub vortex is important in terms of a possibility to causes the performance degradation when the entrained bubbles are mixed into fluid machineries, e.g. pumps. In this study, to confirm the applicability of a model based on circulating annular flow model proposed by the authors, entrained gas flow rate is evaluated using the liquid velocity distribution around free surface dent of vortex (gas core), obtained by CFD data. As a result, it was indicated that it would be possible to evaluate the gas entrainment flow rate by setting an appropriate evaluation region.
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.
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.
Ono, Ayako; Sakashita, Hiroto*; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 7 Pages, 2022/10
The new prediction method of critical heat flux (CHF) of the fuel assemblies based on the mechanism is proposed in this study. The prediction method of CHF based on the mechanism has been needed for a long time to enhance the safety analysis and reduce the design cost. From several experimental findings of the liquid-vapor behavior near the heating surface from the nucleate boiling to the CHF, the authors consider that the macrolayer dryout model will be appropriate to predict the CHF under the reactor condition. The prediction method of the macrolayer thickness and the passage period of vapor mass in the fuel assemblies are needed to predict CHF from the macrolayer dryout model. In this study, the CHF under the forced convection is evaluated by combining the prediction methods for the macrolayer thickness and passage period of vapor mass, which are proposed by authors. The prediction of the CHF under the forced convection is examined and compared with the experimental data.
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.
Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
JAEA is developing the methodology to predict the critical heat flux based on a mechanism in order to reduce the cost for full mock-up test. The evaluation method based on a mechanism is expected to be able to predict in the wide range of parameter under the unexpected conditions including the severe accident. In this study, the JUPITER code developed by JAEA is examined to apply for the two-phase flow simulation of LWR fuel assembly with the spacer grid. The benchmark data of single-phase flow in the bundle with the spacers by KAERI were used to validate the simulation result by JUPITER. Moreover, the single-phase flow simulation was conducted by another simulation method, STAR-CCM+, as a supplemental analysis to consider the effect of the different simulation methods. Finally, the two-phase flow simulation for the bundle with the spacer was conducted by JUPITER. The effect of the spacer with a vane on the bubble behavior is discussed.
Okawa, Tomio*; Mori, Shoji*; Liu, W.*; Ose, Yasuo*; Yoshida, Hiroyuki; Ono, Ayako
Nihon Genshiryoku Gakkai-Shi ATOMO
, 63(12), p.820 - 824, 2021/12
The evaluation method of the critical heat flux based on the mechanism is needed for the efficient design and development of fuel in reactors and the appropriate safety evaluation. In this paper, the current researches relating to the mechanism of the critical heat flux are reviewed, and the issue to be considered in the future are discussed.
Okagaki, Yuria; Yonomoto, Taisuke; Ishigaki, Masahiro; Hirose, Yoshiyasu
Fluids (Internet), 6(2), p.80_1 - 80_17, 2021/02
4 simulated bundleOno, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
Mechanical Engineering Journal (Internet), 7(3), p.19-00583_1 - 19-00583_12, 2020/06
JAEA is implementing the 3D detailed nuclear-thermal-coupled analysis code to analyze the transition state of the core and to reduce the likelihood of the design. In the development plan, the computational fluid dynamics code based on the VOF method, JUPITER, is applied for TH part of the 3D detailed nuclear-thermal-coupled analysis code.
Watanabe, Taro*; Takata, Takashi; Yamaguchi, Akira*
Nuclear Engineering and Design, 313, p.447 - 457, 2017/03
Times Cited Count：2 Percentile：16.60(Nuclear Science & Technology)Countercurrent flow limitation (CCFL) in a heat transfer tube at a steam generator (SG) of pressurized water reactor (PWR) is one of the important issues on the core cooling under a loss of coolant accident(LOCA). In order to improve the prediction accuracy of the CCFL characteristics in numerical simulations using the volume of fluid (VOF) method with less computational cost, a thin liquid film flow in a countercurrent flow is modeled independently and is coupled with the VOF method. Then, we have carried out numerical simulations of a countercurrent flow in a vertical tube so as to investigate the CCFL characteristics and compare them with the previous experimental results. As a result, it has been concluded that the effect of liquid film entrainment by upward gas flux will cause the difference in the experiments.
Gao Ming Quing*
PNC TN9410 97-016, 42 Pages, 1997/02
There is a free surface at the upper plenum in a reactor vessel of LMFBR.The free surface has spatial gradient caused by the internal coolant flow.This is a disadvantageous factor to engineering from the view point of gas entrainment into coolant. To eliminate the free surface gradients,ring plates about 20cm wide are fitted at about 1 meter under the free surface. They interfere fluid flow,and decrease the component velocity in vertical direction.To investigate the efficiency ofthe ringplates, analyses with the AQUA-VOF code were carried out.For contrast, three conditions were given:Case-1:Without ring plates.Case-2:Ring plates,fitted at 1.125m under the free surface.Case-3:Ring plates,fitted at 1.5m under the free surface. The results shown that the ring plateshave a sufficiently high potential to elminate the free surface gradients due to disperse the momentum along reactor vessel axis to radial direction.In the calculations with ring plate (Case-2 and -3),the maximum free surface heig
Yokoi, Shinobu*; Yamamoto, Tomohiko; Miyazaki, Masashi; Tanaka, Masaaki; Sago, Hiromi*; Morita, Hideyuki*; Ikesue, Shunichi*; Inoue, Takatoshi*
no journal, ,
The design basis ground motions have been revised to improve the seismic resistance of nuclear power plants. The reduction of seismic forces not only horizontally but also vertically has required more critical than in the past to ensure the seismic resistance of components. Notably, the design of a Sodium-Cooled Fast Reactor will require reducing the seismic forces applied to the components because of the components with thin wall thickness. To overcome this problem, the authors plan to introduce a seismic isolation system. When the sloshing wave height is small, it can be approximated with a linear vibration model. However, when the sloshing wave height increases and the sloshing becomes nonlinear, it is necessary to evaluate the wave height using other methods such as numerical analysis. Although the evaluation of nonlinear sloshing wave height and sloshing load is important, there are few examples which quantitatively evaluate the sloshing load acting on roof. This paper reports the results of the reproduction analysis carried out using the VOF method.
Fukuda, Takanari; Uesawa, Shinichiro; Yamashita, Susumu; Yoshida, Hiroyuki
no journal, ,
The volume of fluid (VOF) method is a prominent multiphase flow simulation method. Among different interface capturing schemes (ICSs) in the VOF method, the algorithm of the piecewise linear interface calculation (PLIC) is geometrically rigrous but complex to implement. In contrast, the tangent of hyperbola interface capturing/weighted line interface calculation (THINC/WLIC) offers a simpler algorithm but suffers from numerical diffusion, degrading interface quality. To address this tradeoff, we developed THINC/Advanced WLIC (THINC/AWLIC), which balances implementation cost and interface sharpness. Although these ICSs have undergone numerical benchmarking, their performance in practival engineering conditions has not been sufficiently investigated. To evaluate their applicability to boiling water reactor (BWR) core flows, a liquid-gas two-phase flow in a 
pin bundle geometry was simulated using PLIC, THINC/WLIC, and THINC/AWLIC, and compared with the void fraction data obtained by an experiment. The notable result is that the void fraction values for simply coded THINC/AWLIC are nearly identical with those of PLIC, which maximizes the geometrical fidelity with the expense of the algorithmic complexity. The results indicate the generally high applicability of THINC/AWLIC in predicting void fraction in pin bundle geometry and its advantages over conventional ICSs. However, regardless of ICSs, the simulation based on the VOF method still fails to reproduce experimental void fraction at low gas flow rates, where bubble coalescence is minimal in experiment.
Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
no journal, ,
JAEA is developing the methodology to predict the critical heat flux based on a mechanism in order to reduce the cost for full mock-up test. The evaluation method based on a mechanism is expected to be able to predict in the wide range of parameter under the unexpected conditions including the severe accident. In this study, the JUPITER code developed by JAEA is examined to apply for the two-phase flow simulation of LWR fuel assembly. The simulation results of single-phase flow were compared with the data of previous experimental study. The effect of the vane on the flow field and bubble behavior are discussed and some issues to be considered are extracted.
4 fuel bundle with the spacer by the mechanistically based methodOno, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
no journal, ,
JAEA is implementing the 3D detailed nuclear-thermal-coupled analysis code in order to analyze the transition state of the core and to reduce the likelihood of the design. In the development plan, the computational fluid dynamics code based on the VOF method, JUPITER, is applied for TH part of the 3D detailed nuclear-thermal-coupled analysis code. In this study, the JUPITER code is examined to apply for the two-phase flow simulation of 44 fuel assembly with the spacer grid. The simulation results and previous study relating to observation of the two-phase flow in the fuel bundle are compared to validate the simulation results and considered reproducibility of the effect of the spacer grid.
Fukuda, Takanari; Uesawa, Shinichiro; Yamashita, Susumu; Suzuki, Takayuki*
no journal, ,
Advancements in high-performance computing are enabling the application of computational fluid dynamics (CFD) to engineering-scale multi-phase flow problems. The volume of fluid (VOF) method, which tracks interface movement via scalar VOF value transport, is widely used in CFD. Among interface capturing schemes (ICSs), the piecewise linear interface calculation (PLIC) is the most geometrically accurate but computationally complex. In contrast, the tangent of hyperbola interface capturing/weighted line interface calculation (THINC/WLIC) offers a simpler algorithm but suffers from numerical diffusion, degrading interface quality. To address this tradeoff, we developed THINC/Advanced WLIC (THINC/AWLIC), which balances implementation cost and interface sharpness. Although these ICSs have undergone numerical benchmarking, their performance in complex engineering scenarios remains underexplored. To evaluate their applicability to boiling water reactor (BWR) core flows, a liquid-gas two-phase flow in a 33 bundle system experiment was simulated using PLIC, THINC/WLIC, and THINC/AWLIC. The results showed that PLIC and THINC/AWLIC maintained sharp interfaces and provided realistic results but required nearly three times the computational time of THINC/WLIC. While THINC/WLIC was computationally efficient, it exhibited qualitative discrepancies, including unphysical bubble coalescence and volume dissipation. This led to reduced void fractions near the pin-gap region due to fewer bubble coalescences, attributed to bubble size reduction from volume dissipation. A comparison between numerical simulations and experimental void fraction data will be presented to facilitate further discussion.
