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Hamase, Erina; Miyake, Yasuhiro*; Imai, Yasutomo*; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki
Nuclear Engineering and Design, 432, p.113738_1 - 113738_12, 2025/02
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.
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.
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.
Tanaka, Masaaki; Doda, Norihiro; Hamase, Erina; Kuwagaki, Kazuki; Mori, Takero; Okajima, Satoshi; Kikuchi, Norihiro; Yoshimura, Kazuo; Matsushita, Kentaro; Hashidate, Ryuta; et al.
Dai-28-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2024/06
To assist conceptual studies of various reactor systems conducted by private sectors in nuclear power innovation, development of an innovative design system named ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle) is undergoing. 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 to be used for the detailed analysis to confirm the plant performance after optimization are introduced at this point in time.
Doda, Norihiro; Nakamine, Yoshiaki*; Yoshimura, Kazuo; Kuwagaki, Kazuki; Hamase, Erina; Yokoyama, Kenji; Kikuchi, Norihiro; Mori, Takero; Hashidate, Ryuta; Tanaka, Masaaki
Keisan Kogaku Koenkai Rombunshu (CD-ROM), 29, 6 Pages, 2024/06
As a part of the development of the "Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle (ARKADIA)" to utilize the knowledge obtained through the sodium-cooled fast reactors (SFRs) and combine the latest numerical simulation technologies, ARKADIA-Design is being developed to support the optimization of SFRs in the conceptual design stage. ARKADIA-Design consists of three systems of Virtual Plant Life System (VLS), Enhanced and AI-aided optimization System (EAS), and Knowledge Management System (KMS). A design optimization framework controls the linkage among the three systems through the interfaces in each system. In this study, we have developed a prototype of the framework for core design optimization using the coupled analysis functions in VLS and optimization control function in the linkage of EAS and VLS to investigate the applicability of the framework to the SFR design optimization process.
Hamase, Erina; Kuwagaki, Kazuki; Doda, Norihiro; Yokoyama, Kenji; Tanaka, Masaaki
Mechanical Engineering Journal (Internet), 11(2), p.23-00440_1 - 23-00440_14, 2024/04
The core design optimization process is being developed as part of the design optimization support tool named ARKADIA-Design. The process performs the integrated analysis with neutronics, thermal-hydraulics, fuel integrity, and plant dynamics using the Bayesian optimization (BO) algorithm, and obtains the optimal design parameters efficiently. In this study, the representative problem was defined based on core design experiences, and the process was specified. To confirm the appropriateness of the definition of representative problem, as a minimum requirement, the single-objective optimization problem was solved by the integrated analysis with neutronics and plant dynamics using the BO. We found the existence of the optimal solution and the agreement between this solution and the reference one. There was the prospect that the process was applicable to the representative problem.
Hamase, Erina; Ohgama, Kazuya; Kawamura, Takumi*; Doda, Norihiro; Tanaka, Masaaki; Yamano, Hidemasa
Annals of Nuclear Energy, 195, p.110157_1 - 110157_14, 2024/01
Times Cited Count：1 Percentile：35.82(Nuclear Science & Technology)To validate the fast reactor plant dynamics analysis code Super-COPD for the loss of flow without scram (LOFWOS) event, we participated in the IAEA benchmark for the LOFWOS test No.13 performed at the FFTF as one of the passive safety demonstration test. In the blind phase, there were challenges to reproduce outlet temperatures of fuel assemblies and the total reactivity. To improve the evaluation accuracy of them, the whole core model considering the radial heat transfer and interwrapper flow and the simplified assembly bowing reactivity model were introduced. As a result of the final phase, the second peak of outlet temperatures was reproduced successfully, and the total reactivity could generally follow the measured data. Super-COPD was validated for the LOFWOS event.
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.
Doda, Norihiro; Kato, Shinya; Hamase, Erina; Kuwagaki, Kazuki; Kikuchi, Norihiro; Ohgama, Kazuya; Yoshimura, Kazuo; Yoshikawa, Ryuji; Yokoyama, Kenji; Uwaba, Tomoyuki; et al.
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.946 - 959, 2023/08
An innovative design system named ARKADIA is being developed to realize the design of advanced nuclear reactors as safe, economical, and sustainable carbon-free energy sources. This paper focuses on ARKADIA-Design for design studies as a part of ARKADIA and introduces representative verification methods for numerical analysis methods of the core design. ARKADIA-Design performs core performance analysis of sodium-cooled fast reactors using a multiphysics approach that combines neutronics, thermal-hydraulics, core mechanics, and fuel pin behavior analysis codes. To confirm the validity of these analysis codes, validation matrices are identified with reference to experimental data and reliable numerical analysis results. The analysis models in these codes and the representative practices for the validation matrices are described.
Doda, Norihiro; Nakamine, Yoshiaki*; Kuwagaki, Kazuki; Hamase, Erina; Kikuchi, Norihiro; Yoshimura, Kazuo; Matsushita, Kentaro; Tanaka, Masaaki
Keisan Kogaku Koenkai Rombunshu (CD-ROM), 28, 5 Pages, 2023/05
As a part of the development of the "Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle (ARKADIA)" to automatically optimize the life cycle of innovative nuclear reactors including fast reactors, ARKADIA-design is being developed to support the optimization of fast reactor in the conceptual design stage. ARKADIA-Design consists of three systems (Virtual plant Life System (VLS), Evaluation assistance and Application System (EAS), and Knowledge Management System (KMS)). A design optimization framework controls the connection between the three systems through the interfaces in each system. This paper reports on the development of an optimization analysis control function that performs design optimization analysis combining plant behavior analysis by VLS and optimization study by EAS.
Hamase, Erina; Kuwagaki, Kazuki; Doda, Norihiro; Yokoyama, Kenji; Tanaka, Masaaki
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
To innovate a core design process, an optimization process for the core design has been developed as a part of the design optimization support tool named ARKADIA-Design. The core design optimization process is integrated by the core design analysis of neutronics, thermal-hydraulics, and fuel integrity and plant dynamics analysis with the Bayesian optimization (BO) algorithm. The optimization problem for design parameters with high core performance and inherent safety in ULOF event was solved by the integrated analysis between the neutronics and plant dynamics with the BO in a primary loop system including a core consisting of two-dimensional RZ cylindrical geometry. It was indicated that the optimization process could obtain an optimal solution.
Tanaka, Masaaki; Miyake, Yasuhiro*; Ezure, Toshiki; Hamase, Erina
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05
The numerical analysis model for the computational fluid dynamics (CFD) code for the design study is developed to evaluate the thermal-hydraulics in the core under the core-plenum interaction (CPI) during the decay heat removal using the dipped type direct heat exchanger (D-DHX). To judge the adequacy of the numerical results for a validation study with the sodium experiment results conducted at PLANDTL-2 facility, the degree of difference (DoD) between the numerical and experimental results must be measured by using the area validation metrics (AVM). Through the examinations, the applicability of the AVM and MAVM based on the p-box method was confirmed.
Hamase, Erina; Ohgama, Kazuya; Kawamura, Takumi*; Doda, Norihiro; Yamano, Hidemasa; Tanaka, Masaaki
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 9 Pages, 2022/10
To improve the prediction accuracy of the plant dynamics analysis code named Super-COPD, JAEA has joined the IAEA benchmark for the FFTF Loss of Flow Without Scram Test No.13. In the first blind phase, there was the challenge to perform outlet temperatures of fuel assemblies more accurately. Hence, the renewed analysis was performed with the whole core multi-channel model in which each assembly was modelled to simulate the radial heat transfer among assemblies and the flow redistribution induced by the buoyancy in the NC conditions. Then, to validate the coupled transient analysis between the whole core multi-channel model and the one-point kinetics model, the analysis considering major reactivity feedbacks such as GEM, assembly bowing was performed. As a result, the second peak of outlet temperatures was reproduced successfully, and it was observed that the plant dynamics analysis could follow the measured data.
Hamase, Erina; Miyake, Yasuhiro*; Imai, Yasutomo*; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 12 Pages, 2022/09
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, development of the reactor vessel model including the more model by using a computational fluid dynamics code (RV-CFD) is required. Previously, the CFD model based on the subchannel analysis was developed. In this study, to achieve much lower computational cost maintaining the prediction accuracy, the coarse-mesh subchannel CFD (CMSC) model has been 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 radial heat transfer under NC conditions.
Doda, Norihiro; Yoshimura, Kazuo; Hamase, Erina; Yokoyama, Kenji; Uwaba, Tomoyuki; Tanaka, Masaaki
Proceedings of Technical Meeting on State-of-the-art Thermal Hydraulics of Fast Reactors (Internet), 3 Pages, 2022/09
ARKADIA-Design is being developed to support the optimization of sodium-cooled fast reactors in the conceptual design stage. Design optimization requires various types of numerical analysis: 1-D plant dynamics analysis for efficient evaluation of various design options and multi-dimensional analysis for a detailed evaluation of local phenomena, including multi-physics. For those analyses, ARKADIA-Design performs whole plant analyses based on the multi-level simulation (MLS) technique in which the analysis codes are coupled to simulate the phenomena in an intended degree of resolution. This paper describes an outline of the coupling analysis methods in the MLS of the ARKADIA-Design and the numerical simulations of the experimental fast breeder reactor EBR-II tests by the coupled analysis.
Hamase, Erina; Miyake, Yasuhiro*; Imai, Yasutomo*; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki
Mechanical Engineering Journal (Internet), 9(4), p.21-00438_1 - 21-00438_15, 2022/08
To enhance the safety of sodium-cooled fast reactors, a dipped-type direct heat exchanger (D-DHX) has been investigated in a natural circulation decay heat removal system. During the D-DHX operation, the core-plenum interactions occurs and the thermal-hydraulics in the reactor vessel (RV) is complicated, the establishment of thermal-hydraulic analysis model in the RV for computational fluid dynamics code (RV-CFD) is required to simulate the thermal stratification in the upper plenum and thermal-hydraulics in the core. In this study, in terms of using RV-CFD for design study, the subchannel CFD model with low computational cost was adopted to the core of RV-CFD and the numerical simulation was carried out in comparison with the measured data in the sodium test facility named PLANDTL-1. As the result, the calculated sodium temperature in the core had good agreement with the experimental result and the applicability of the RV-CFD for the core-plenum interactions was confirmed.
Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.
Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07
This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.
Doda, Norihiro; Hamase, Erina; Kikuchi, Norihiro; Tanaka, Masaaki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 10 Pages, 2022/04
In conventional design studies of sodium-cooled fast reactors, plant dynamics and local phenomena were evaluated separately by using simple models and detailed models, respectively, and their interaction was considered through the boundary conditions settings with conservativeness for each individual analysis. Thus, the final result through the analyses may contain excessive conservativeness. Therefore, JAEA began to develop a multi-level simulation system in which detailed analysis codes are coupled with a plant dynamics analysis code. Focusing on core thermal-hydraulics, a coupled analysis method using a plant dynamics analysis code Super-COPD and a subchannel analysis code ASFRE has been developed. The analysis on a test in the experimental fast reactor EBR-II was performed to validate the coupled analysis. Through the comparison of the analysis results and the measurement, it was confirmed that the coupled analysis could predict the transient temperature distribution in the subassembly, and the multi-level simulation by changing the level of detail in analysis model could be performed for core thermal-hydraulics.
Hamase, Erina; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Imai, Yasutomo*
Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 10 Pages, 2021/08
To enhance the safety of sodium-cooled fast reactors, a dipped-type direct heat exchanger (D-DHX) has been investigated in a natural circulation decay heat removal system. During the D-DHX operation, the core-plenum interactions occurs, therefore, a thermal-hydraulic analysis model in the reactor vessel for computational fluid dynamics code (RV-CFD model) is necessarily required. In this study, the application of the subchannel analysis method for subassemblies to the RV-CFD model was attempted to reduce the calculation costs. Analysis results were compared to the experimental data obtained in the sodium experimental apparatus PLANDTL-1. As the result, the behavior of cold sodium into the simulated core was well grasped and the calculated sodium temperature in the core had good agreement with the experimental result. The applicability of the RV-CFD model was confirmed.
