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Yoshimura, Kazuo; Doda, Norihiro; Tanaka, Masaaki; Fujisaki, Tatsuya*; Murakami, Satoshi*
Annals of Nuclear Energy, 226, p.111896_1 - 111896_11, 2026/02
At the Japan Atomic Energy Agency, a multilevel simulation (MLS) methodology which enables consistent evaluation from whole plant behavior to local phenomena in the plant components is being developed to attempt plant design and enhance the safety of sodium-cooled fast reactors. To validate the coupling method in the MLS system, the 1D-CFD coupling method using Super-COPD for 1D plant dynamics analysis and Fluent for multi-dimensional CFD analysis was applied to the analyses of loss of flow tests in EBR-II. It was confirmed that it could predict multi-dimensional thermal-hydraulic phenomena such as thermal stratification in the upper plenum, Z-shaped pipe, and cold pool, holding the whole plant behavior simultaneously. Moreover, the applicability of the 1D-CFD coupling method to the evaluation of the phenomena in natural circulation conditions was confirmed by comparing the results of the 1D-CFD couple analyses and the measured data.
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.
Matsushita, Kentaro; Ezure, Toshiki; Fujisaki, Tatsuya*; Nakamine, Yoshiaki*; Imai, Yasutomo*; Tanaka, Masaaki
Nihon Kikai Gakkai 2025-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2025/09
In the design of sodium-cooled fast reactors (SFRs), it is important to evaluate the transition behavior of non-condensable gas entrained into the primary coolant system due to cover gas entrainment and dissolution. In this study, trajectories of non-condensable gas bubbles in the cold plenum of the pool-type SFR evaluated by three-dimensional CFD analyses applying Discrete Phase Model. As the result of sensitivity analyses regarding bubble radius flowing into the cold plenum, it was clarified that the release rate of bubbles showed an increase according to the increase of bubble radius and an asymptotic increasing behavior in the large bubble radius cases.
Hamase, Erina; Kawamura, Takumi*; Doda, Norihiro; Tanaka, Masaaki
Nihon Kikai Gakkai 2025-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2025/09
To investigate the applicability of uncertainty quantification (UQ) and sensitivity analysis (SA) methods for validating a fast reactor plant dynamics analysis code, forward UQ and SA employing Sobol' method were performed for FFTF LOFWOS test No.13. The result demonstrated that validity can be judged if the test results fall within the quantified uncertainty range, and that the dominant input parameters influencing uncertainty can be quantitatively evaluated, enabling prioritization of parameters for uncertainty reduction. This confirms the applicability of forward UQ and SA employing Sobol' method.
Kobayashi, Jun; Tanaka, Masaaki; Hamase, Erina; Ezure, Toshiki
JAEA-Data/Code 2025-009, 74 Pages, 2025/08
JAEA-Data-Code-2025-009.pdf:4.7MB
In a sodium-cooled fast reactor, a diversified auxiliary cooling system to remove decay heat from the core is required to enhance its safety. The decay heat removal systems (DHRSs) include a direct reactor auxiliary cooling system (DRACS) with a heat exchanger in the upper plenum (UP) of the reactor vessel (RV), a primary reactor auxiliary cooling system (PRACS) with a heat exchanger in the primary heat transport system (PHTS), an intermediate reactor auxiliary cooling system (IRACS) with a heat exchanger in the secondary heat transport system (SHTS), a heat removal system which employs a steam generator, and a reactor vessel auxiliary cooling system (RVACS) that effects cooling from outside the RV. In the operation of the DRACS with a dipped-type direct heat exchanger (D-DHX) installed in the UP of the RV (UP-RV), the thermal interaction, called core-plenum interaction (CPI), regarding the thermal-hydraulic phenomena in the UP-RV and the core is observed. The CPI includes the penetration flow of the sodium at a low temperature from the D-DHX into the core assemblies, the flow in the gap between assemblies, and the radial heat transfer through sodium in the gap. On the other hand, in the operation of the PRACS or IRACS, where the flowrate in the PHTS is maintained, the core coolability is affected by plant operating conditions. Two transient tests conducted at the PLANDTL-DHX sodium test facility in Japan Atomic Energy Agency were provided to develop an appropriate numerical analysis model for prediction of transient thermal-hydraulics in the DHRSs, the core, and the PHTS. In this document, the geometry information of the RV and the PHTS, including the heat exchangers for the DHRS, and the measured flowrate and temperature transients at each inlet of the intermediate heat exchanger (IHX) on the SHTS side and DHRS were specified as the boundary conditions for the benchmark analyses.
Hamase, Erina; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Imai, Yasutomo*
Proceedings of 21st International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-21) (Internet), 14 Pages, 2025/08
We have developed the reactor vessel thermal-hydraulic analysis model (RV-CFD) with the subchannel CFD (SC) model for assembly with a low computational cost to evaluate the core-plenum interactions occurred in the natural circulation decay heat removal during the dipped-type direct heat exchanger operation in sodium-cooled fast reactor. In this study, the non-equilibrium thermal model which can consider the heat capacity and thermal load of fuel pins was developed in the SC model. Through the transient analysis simulating the power reduction due to reactor scram using the RV-CFD, the applicability of RV-CFD to the transient analysis was confirmed.
Ezure, Toshiki; Matsushita, Kentaro; Sasaki, Keisuke; Tanaka, Masaaki
Dai-29-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Yokoshu (Internet), 5 Pages, 2025/06
In the design of the pool-type demonstration sodium-cooled fast reactor (demonstration reactor), the prevention of gas entrainment in the hot plenum of the reactor vessel is one of important issues to be addressed in the conceptual design of demonstration reactor. Related to this problem, the authors have been developing an evaluation approach combining the analysis method of entrained gas-transport in the primary circuit, SYRENA, and the gas entrainment evaluation method, StreamViewer, at the free surface in the hot plenum of the demonstration reactor. In this study, a development plan of StreamViewer is presented toward application to the evaluation of the demonstration reactor design. Furthermore, an overview of scaled model water experiment of the pool-type demonstration reactor to obtain the validation date for StreamViewer is also presented.
Alzahrani, H.*; Matsushita, Kentaro; Sakai, Takaaki*; Ezure, Toshiki; Tanaka, Masaaki
Nuclear Technology, 13 Pages, 2025/06
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Development of evaluation method for cover gas entrainment by vortices generated at free surface in upper plenum of sodium-cooled fast reactor is required, and an evaluation method by predicting vortices from flow velocity distribution obtained by CFD analysis is developed. In this study, Adaptive Mesh Refinement (AMR) method is examined to improve efficiency of CFD analysis. Initial mesh was refined with two indexes: the first index (Index-1) is when the second invariant of velocity gradient tensor, Q, is negative and the second one (Index-2) is pressure gradient index added to Index-1. As a result of applying AMR method to unsteady vortices system with a flat plate and performing transient analyses with refined meshes, the result of pressure distribution and velocity around the flat plate in mesh using Index-2 was similar to the result of all refined mesh. It was also confirmed that vortices generation and growth was better simulated by refining meshes around separation area.
Kawasaki, Takuro; Fukuda, Tatsuo; Yamanaka, Satoru*; Murayama, Ichiro*; Kato, Takanori*; Baba, Masaaki*; Hashimoto, Hideki*; Harjo, S.; Aizawa, Kazuya; Tanaka, Hirohisa*; et al.
Journal of Applied Physics, 137(9), p.094101_1 - 094101_7, 2025/03
Times Cited Count：0 Percentile：0.00(Physics, Applied)Tanaka, Masaaki; Doda, Norihiro; Hamase, Erina; Kuwagaki, Kazuki; Mori, Takero; Okajima, Satoshi; Kikuchi, Norihiro; Yoshimura, Kazuo; Matsushita, Kentaro; Hashidate, Ryuta; et al.
Nihon Kikai Gakkai Rombunshu (Internet), 91(943), p.24-00229_1 - 24-00229_12, 2025/03
To assist conceptual studies of various reactor systems conducted by private sectors in nuclear power innovation, an innovative design system named ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle) has been developed. In this paper, focusing on the ARKADIA-Design, achievements in the development of optimization processes in the fields of the core design, the plant structure design, and the maintenance schedule planning, as major function of ARKADIA-Design, and numerical analysis methods including coupled analysis to be used for the detailed analysis to confirm the plant performance after optimization are introduced at this point in time.
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.
Matsushita, Kentaro; Ezure, Toshiki; Tanaka, Masaaki; Imai, Yasutomo*; Fujisaki, Tatsuya*; Sakai, Takaaki*
Nuclear Engineering and Design, 432, p.113785_1 - 113785_16, 2025/02
Times Cited Count：1 Percentile：37.73(Nuclear Science & Technology)Establishing an evaluation method for the gas entrainment (GE) of argon cover gas due to surface vortices is required in terms of safety design of sodium-cooled fast reactors. To modify the evaluation model in an in-house evaluation tool for GE, StreamViewer, a modified evaluation model on the pressure distribution along the vortex center line (PVL model) was proposed to identify the vortex center lines by connecting continuous vortex center points from the suction port to the surface and evaluate gas core length based on the balance between the hydrostatic pressure and the pressure decrease distribution along the vortex center line. PVL model was applied the three-dimensional numerical analysis results for the experiments where a plate induced unsteady traveling vortices in the open channel flow. Consequently, the GE evaluation using StreamViewer with PVL model could reproduce the relation between the inlet flow velocity and the gas core length in the unsteady vortex flow experiments.
Kuwagaki, Kazuki; Hamase, Erina; Yokoyama, Kenji; Doda, Norihiro; Tanaka, Masaaki
Annals of Nuclear Energy, 225, p.111754_1 - 111754_10, 2025/01
Times Cited Count：0Hayakawa, 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.
Ohgama, Kazuya; Doda, Norihiro; Uwaba, Tomoyuki; Futagami, Satoshi; Tanaka, Masaaki; Yamano, Hidemasa; Ota, Hirokazu*; Ogata, Takanari*; Wozniak, N.*; Shemon, E.*; et al.
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
To enhance the accuracy of the safety evaluations in sodium-cooled fast reactors, it is necessary to develop a method to realistically evaluate the reactivity caused by core deformation. In this regard, Japan and the United States jointly conducted a benchmark analysis of thermal bowing experiments of a single duct of Joyo-type fuel assembly. The aim was to confirm the validity of the core bowing analysis codes. Comparisons of analysis and test results revealed that the core bowing analysis codes used by both countries were able to reasonably predict the axial distribution of horizontal duct displacement of a single duct due to thermal bowing and the contact load on the duct pad.
Wozniak, N.*; Shemon, E.*; Feng, B.*; Ohgama, Kazuya; Doda, Norihiro; Uwaba, Tomoyuki; Futagami, Satoshi; Tanaka, Masaaki; Yamano, Hidemasa; Ota, Hirokazu*; et al.
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
To enhance the accuracy of the safety evaluations in sodium-cooled fast reactors, it is necessary to develop a method to realistically evaluate the reactivity caused by core deformation. In this regard, Japan and the United States jointly conducted a benchmark analysis of thermal bowing experiments using multiple ducts of Joyo-type fuel assembly. The aim was to confirm the validity of the core bowing analysis codes. Comparisons of analysis and test results revealed that the core bowing analysis codes used by both countries were able to reasonably predict the thermal bowing of a row of assemblies.
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.
Matsushita, Kentaro; Ezure, Toshiki; Fujisaki, Tatsuya*; Imai, Yasutomo*; Tanaka, Masaaki
Nihon Kikai Gakkai 2024-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2024/09
An evaluation method of gas entrainment phenomena due to free surface vortices has been developed for the design of a reactor vessel of sodium-cooled fast reactor. The method predicts vortex dimple using the vortex model to the flow field obtained from three dimensional hydraulic analyses of an evaluation area. In this study, the application of adaptive mesh refinement (AMR) method to a water flow experiment in a rectangular channel with advection vortices was examined to create analysis meshes automatically. Transient analyses were conducted using refined meshes obtained by AMR under different initial grid size conditions. Then, the quantities related to vortex formation and the computation cost were compared with the result in a reference mesh with uniformly fine grids. As the result, it was confirmed that the variation of the grid number is possible to use as a criterion to judge the refinement termination in AMR, and the calculated cost of transient analysis can be reduced by AMR.
Yoshikawa, Ryuji; Kikuchi, Norihiro; Tanaka, Masaaki
Nihon Kikai Gakkai 2024-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2024/09
In the study of safety enhancements on advanced sodium-cooled fast reactor, it has been essential to evaluate the influence of buoyancy on pressure drop in a fuel assembly at mixed convection condition during natural circulation under the decay heat removal operation. In this study, the numerical simulations of the 19-rod and 91-rod bundle water experiments at low flow rate conditions were performed for the validation of a thermal-hydraulic analysis code named SPIRAL with the hybrid turbulence model. The influence of buoyancy on the velocity and temperature distributions was analyzed, and the applicability of the hybrid turbulence model to the pressure drop evaluation was investigated by comparison with the experimental friction factors.
