Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 136

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Development of a coarse-mesh subchannel CFD model for prediction of core thermal-hydraulics in natural circulation conditions

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

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.

Journal Articles

Development of VVUQ method for ensuring credibility of plant dynamics analysis results based on statistical approach

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.

Journal Articles

Validation study on SFR core bowing codes using Joyo ex-core experiment data; Single duct bowing benchmark

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.

Journal Articles

Validation study on SFR core bowing codes using Joyo ex-core experiment data; Multiple duct bowing benchmark

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.

Journal Articles

Development of reactor vessel thermal-hydraulic analysis method in natural circulation conditions; Applicability investigation for transient analysis

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.

Journal Articles

Study on safety analyses for metal-fueled sodium-cooled fast reactors; Project overview

Yamano, Hidemasa; Futagami, Satoshi; Doda, Norihiro; Tagami, Hirotaka; Uchibori, Akihiro; Ogata, Takanari*; Ota, Hirokazu*

Nihon Kikai Gakkai 2024-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2024/09

Journal Articles

Development of reactor vessel thermal-hydraulic analysis method in natural circulation conditions; Investigation of interwrapper Gap model

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.

Journal Articles

Development of ARKADIA for the innovation of advanced nuclear reactor design process (Development status of the design optimization support tool, ARKADIA-Design)

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.

Journal Articles

Development of a design optimization framework for sodium-cooled fast reactors, 3; Development of a prototype with user interface

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.

Journal Articles

Development of core design optimization process by integrated analysis with neutronics and plant dynamics

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.

Journal Articles

Analysis of fuel assemblies inclination due to upper core support plate deflection for reactivity evaluation

Yoshimura, Kazuo; Doda, Norihiro; Igawa, Kenichi*; Uwaba, Tomoyuki; Tanaka, Masaaki; Nemoto, Toshiyuki*

Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 8 Pages, 2024/03

To investigate possibility of the insertion of the reactivity by the deflection of the upper core support plate, structural mechanics analyses of the domain consisting of the fuel assemblies and core support plates and evaluation of the reactivity due to the inclination of the fuel assemblies in EBR-II were carried out. As a result, it was indicated that the upper core support plate deflected downward larger at the low flowrate condition than that at the high flowrate condition and positive reactivity was inserted due to the inclination of the fuel assemblies at the low flowrate condition.

Journal Articles

Validation of the fast reactor plant dynamics analysis code Super-COPD using FFTF loss of flow without scram test #13

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:34.39(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.

Journal Articles

Numerical simulation technologies for safety evaluation in plant lifecycle optimization method, ARKADIA for advanced reactors

Uchibori, Akihiro; Doda, Norihiro; Aoyagi, Mitsuhiro; Sonehara, Masateru; Sogabe, Joji; Okano, Yasushi; Takata, Takashi*; Tanaka, Masaaki; Enuma, Yasuhiro; Wakai, Takashi; et al.

Nuclear Engineering and Design, 413, p.112492_1 - 112492_10, 2023/11

 Times Cited Count:2 Percentile:57.39(Nuclear Science & Technology)

The ARKAIDA has been developed to realize automatic optimization of plant design from safety evaluation for the advanced reactors represented by a sodium-cooled fast reactor. ARKADIA-Design offers functions to support design optimization both in normal operating conditions and design basis events. The multi-level simulation approach by the coupled analysis such as neutronics, core deformation, core thermal hydraulics was developed as one of the main technologies. On the other hand, ARKAIDA-Safety aims for safety evaluation considering severe accidents. As a key technology, the numerical methods for in- and ex-vessel coupled phenomena during severe accidents in sodium-cooled fast reactors were tested through a hypothetical severe accident event. Improvement of the ex-vessel model and development of the AI technology to find best design solution have been started.

Journal Articles

Development of virtual plant model for design rationalization of fast reactors by multi-level simulation system; Confirmation of functionality in application to U.S. experimental fast reactor EBR-II

Yoshimura, Kazuo; Doda, Norihiro; Nakamine, Yoshiaki*; Fujisaki, Tatsuya*; Igawa, Kenichi*; Iida, Masaki*; Tanaka, Masaaki

Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09

In Japan Atomic Energy Agency, a virtual plant model of the sodium-cooled fast reactor plant composed in a computer is being developed to reduce the development cost, by replacing the experiments to the numerical simulations with coupled analyses of the physical phenomena accounting for the interaction between components under various plant conditions. Through the numerical analysis of the ULOHS test in the U.S. experimental fast reactor named EBR-II, applicability of the virtual plant model was confirmed in comparison with the measured data including the core inlet temperature and the reactor power.

Journal Articles

Development of reactor vessel thermal-hydraulic analysis method in natural circulation conditions; Investigation of thermal-hydraulic analysis model for interwrapper gap between assemblies

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.

Journal Articles

Development of a statistical evaluation method for core hot spot temperature in sodium-cooled fast reactor under natural circulation conditions

Doda, Norihiro; Igawa, Kenichi*; Iwasaki, Takashi*; Murakami, Satoshi*; Tanaka, Masaaki

Nuclear Engineering and Design, 410, p.112377_1 - 112377_15, 2023/08

 Times Cited Count:1 Percentile:34.39(Nuclear Science & Technology)

To enhance the safety of sodium-cooled fast reactors, the decay heat in the core must be removed by natural circulation even if the AC power supply to the forced circulation equipment is lost. Under natural circulation conditions, sodium flow is driven by buoyancy, and flow velocity and temperature distribution influence each other. Thus, it is difficult to evaluate the core hot spot temperature by deterministically considering the uncertainties affecting flow and heat. In this study, a statistical evaluation method is developed for the core hot spot temperature by using Monte Carlo sampling methods. The applicability of the core hotspot evaluation method was confirmed in three representative events during natural circulation decay heat removal operations in loop-type sodium-cooled fast reactors.

Journal Articles

Validation practices of multi-physics core performance analysis in an advanced reactor design study

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.

Journal Articles

Investigation on applicability of subchannel analysis code ASFRE to thermal hydraulics analysis in fuel assembly with inner duct structure of sodium cooled fast reactor

Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki

Journal of Nuclear Engineering and Radiation Science, 9(3), p.031401_1 - 031401_11, 2023/07

In the design study of an advanced sodium-cooled fast reactor (Advanced-SFR) investigated in JAEA, the use of a specific fuel assembly with an inner duct structure called FAIDUS has been investigated to enhance safety of Advanced-SFR. Since the fuel rods have an asymmetric layout by the inner duct, the validity confirmation of the numerical results of an in-house subchannel analysis code named ASFRE was required. In this paper, therefore, the code-to-code comparisons was applied with numerical results of ASFRE and those of an in-house CFD code named SPIRAL. The applicability of ASFRE was indicated through the confirmation of the consistency of specific temperature distributions.

Journal Articles

Development of a design optimization framework for sodium-cooled fast reactors, 2; Development of optimization analysis control function

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.

Journal Articles

Investigation of optimization process for core design with integrated analysis between neutronics and plant dynamics

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.

136 (Records 1-20 displayed on this page)