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Okajima, Satoshi; Mori, Takero; Kikuchi, Norihiro; Tanaka, Masaaki; Miyazaki, Masashi
Mechanical Engineering Journal (Internet), 10(4), p.23-00042_1 - 23-00042_12, 2023/08
In this paper, we propose the simplified procedure to estimate failure probability of components subjected to thermal transient for the design optimization. Failure probability can be commonly used as an indicator of component integrity for various failure mechanisms. In order to reduce number of analyses required for one estimation, we have adopted the First Order Second Moment (FOSM) method as the estimation method of failure probability on the process of the optimization, and an orthogonal table in experiment design method is utilized to define conditions of the analyses for the evaluation of the input parameters for the FOSM method. The superposition of ramp responses is also utilized to evaluate the time history of thermal transient stress instead of finite element analysis. Through the demonstration study to optimize thickness of cylindrical vessel subjected to thermal transient derived from shutdown, we confirmed that the procedure can evaluate the failure probability depending on the cylinder thickness with practical calculation cost.
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.
Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Gerschenfeld, A.*
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
Removal of core decay heat by utilizing natural circulation is expected as a significant measure to enhance the safety of sodium-cooled fast reactors (SFRs). Accurate evaluation of the temperature distribution in the fuel assembly (FA) at the low Re regime in natural circulation operation is demanded. A detailed thermal-hydraulics analysis code named SPIRAL has been developed to clarify thermal-hydraulic phenomena in the FA at various operation conditions. In this study, SPIRAL with the hybrid turbulence model was applied to analyze a large-scale fuel assembly experiment of a 91-pin bundle for two cases at the mixed and the natural convection conditions respectively in low Re regime with heat transfer from outside of the FA. The applicability of the SPIRAL to the thermal-hydraulics evaluation of FA at mixed and natural convection conditions was confirmed by the comparisons of temperatures predicted by SPIRAL with those measured in the experiment.
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.
Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki
Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 8 Pages, 2021/08
In the design study of an advanced sodium-cooled fast reactor (Advanced-SFR) in JAEA, the use of a specific fuel assembly (FA) with an inner duct structure called FAIDUS has been investigated to enhance safety of Advanced-SFR. Due to the asymmetric layout of fuel rods by the inner duct, it is necessary to estimate the temperature distribution to confirm feasibility of FAIDUS. For the FAIDUS, confirmation of validity of the numerical results by a subchannel analysis code named ASFRE was not enough because the reference data on the thermal hydraulics in FAIDUS have not been obtained by the mock-up experiment, yet. Therefore, the code-to-code comparisons with numerical results of ASFRE and those of a CFD code named SPIRAL was conducted. The applicability of ASFRE was indicated through the confirmation of the consistency of mechanism of the specific temperature and velocity distributions appearing around the inner duct between the numerical results by ASFRE and those by SPIRAL.
Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Gerschenfeld, A.*
Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo 2020 (SNA + MC 2020), p.73 - 80, 2020/10
A finite element thermal-hydraulics simulation code SPIRAL has been developed in Japan Atomic Energy Agency (JAEA) to analyze the detailed thermal-hydraulics phenomena in a fuel assembly (FA) of Sodium-cooled Fast Reactors (SFRs). The numerical simulation of a large-scale sodium test for 91-pin bundle (GR91) at low flow rate condition was performed for the validation of SPIRAL with the hybrid k-e turbulence model to take into account the low Re number effect near the wall in the flow and temperature fields. Through the numerical simulation, specific velocity distribution affected by the buoyancy force was shown on the top of the heated region and the temperature distribution predicted by SPIRAL agreed with that measured in the experiment and the applicability of the SPIRAL to thermal-hydraulic evaluation of large-scale fuel assembly at low flow rate condition was indicated.
Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki
Journal of Nuclear Engineering and Radiation Science, 5(2), p.021001_1 - 021001_12, 2019/04
In the design study of an advanced loop-type sodium-cooled fast reactor in Japan, a specific fuel assembly (FA) named FAIDUS (Fuel Assembly with Inner DUct Structure) has been considered as one of the measures to enhance safety of the reactor during the core disruptive accident. In this study, thermal-hydraulics in FAIDUS was investigated with the in-house subchannel analysis code named ASFRE. Before the application to FAIDUS, applicability of ASFRE to FAs was confirmed through the numerical simulations for the experiments of simulated FA. Through the comparisons between the numerical results of thermal-hydraulic analyses of FAIDUS and a typical FA without the inner duct, it was indicated that significant asymmetric temperature distribution would not occur in FAIDUS at both high and low flow rate conditions.
Kikuchi, Norihiro; Doda, Norihiro; Hashimoto, Akihiko*; Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki
Dai-23-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 5 Pages, 2018/06
For the thermal-hydraulic design regarding a fuel assembly of sodium-cooled fast reactors, a subchannel analysis code ASFRE has been developed by JAEA. ASFRE was applied to numerical simulations of several kinds of water and sodium experiments as its validation studies and it was confirmed that pressure drops and temperature distributions measured in the experiments can be well reproduced. To enhance safety of sodium-cooled fast reactor, it is required to evaluate thermal-hydraulics in a core during decay heat removal by natural circulation. It is necessary to estimate radial heat transfer phenomena between fuel assemblies. In this study, a numerical simulation of a 37-pin bundle sodium experiment with radial heat flux was carried out and it was confirmed that ASFRE can be qualitatively reproduced temperature distributions in a fuel assembly affected by radial heat transfer.
Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2017 Koen Rombunshu (CD-ROM), 4 Pages, 2017/08
A specific fuel assembly named FAIDUS (Fuel Assembly with Inner Duct Structure) has been developed as one of the measures to enhance safety of the reactor in the core disruptive accident (CDA) in JAEA. Thermal-hydraulics evaluations in FAIDUS under various operation conditions including the CDA are required to confirm its design feasibility. Therefore, numerical simulations by using thermal-hydraulics analysis program named SPIRAL developed in JAEA are conducted to analyze the thermal-hydraulics in the FAIDUS. Through the numerical simulation in the FAIDUS under tentative rated operation condition of an Advanced SFR, it was indicated that the flow and temperature distribution in the FAIDUS showed the same tendency as that in ordinary FA and specific characteristics was not observed.
Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 12 Pages, 2017/07
In the design study of an advanced loop-type SFR in JAEA, a specific fuel assembly (FA) named FAIDUS (Fuel Assembly with Inner DUct Structure) has been adopted as one of the measures to enhance safety of the reactor. Thermal-hydraulics evaluations of FAIDUS under various operation conditions are required to confirm its design feasibility. In this study, after the applicability of ASFRE to FAs was confirmed through the numerical analysis using simulated FA tests, thermal-hydraulic analyses of a FA without an inner duct and a FAIDUS were conducted. Through the numerical analyses, it was indicated that asymmetric temperature distribution in a FAIDUS would not be occurred and characteristics of the temperature distribution was almost the same as that in a FA without an inner duct. Under the low flow rate condition, it was expected that the local flow acceleration caused by the buoyancy force in a FAIDUS could bring the flow redistribution and make the temperature distribution flat.
Kikuchi, Norihiro; Ohshima, Hiroyuki; Tanaka, Masaaki; Hashimoto, Akihiko*
Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2016/06
For the thermal-hydraulic design and safety assessment regarding a fuel assembly of sodium-cooled fast reactors, a subchannel analysis code ASFRE has been and is continuously developed in JAEA. In the numerical simulation of ASFRE confirmed that the tendency to overestimate the maximum coolant temperature in a fuel assembly still remains. In this study, Distributed Resistance Model (DRM), which deals with wire-spacer wrap volumetric effect in subchannels on peripheral and axial directions, was modified and its calibration factor was optimized in order to improve the prediction accuracy of the maximum coolant temperature. A numerical simulation of a 37-pin bundle sodium experiment was also carried out and the result showed the validity of the modified DRM.
Kikuchi, Norihiro; Ohshima, Hiroyuki; Imai, Yasutomo*; Hiyama, Tomoyuki; Nishimura, Masahiro; Tanaka, Masaaki
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2015 Koen Rombunshu, p.179 - 180, 2015/08
In an economically improved sodium-cooled fast reactor, a narrower gap is considered among the fuel pins so as to achieve a high burn-up. Therefore, it is needed to evaluate thermal-hydraulic characteristics in case of a change of the gap geometry due to deformation of fuel pin caused by such as a swelling and a thermal bowing. For this purpose, a FEM analysis code, SPIRAL has been being developed in JAEA and the code validations using water or sodium experimental results have also being performed. In this study, a numerical analysis of a flow field around wire-wrapped fuel pins based on a 3-pin bundle water experiment was carried out as a validation study of SPIRAL. As a result, it was demonstrated that the hybrid-type turbulent model incorporated in SPIRAL has a high applicability to investigate the flow structure of the narrow gap in the fuel assembly.
Kikuchi, Norihiro; Imai, Yasutomo*; Ohshima, Hiroyuki; Tanaka, Masaaki
no journal, ,
A computer program SPIRAL, which has been developed for detailed thermal-hydraulic analyses in a fuel assembly of sodium-cooled fast reactors, was applied to a preliminary analysis of an FAIDUS (Fuel Assembly with an Inner Duct Structure)-type fuel assembly for investigating the applicability of the program and clarifying thermal-hydraulic characteristics of FAIDUS. Numerical simulations were conducted for a FAIDUS-type 33-pins assembly in which 4-pins located at a corner of hexagonal wrapper tube were removed to simulate inner duct part and a 37-pins assembly without inner duct part (reference case). The comparison of two cases showed that the inner duct structure had limited effect on the temperature distribution in the fuel assembly and the pressure loss coefficient was mostly unchanged. Through the numerical simulations, the applicability of SPIRAL to the FAIDUS-type assembly was confirmed.
Kikuchi, Norihiro; Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki
no journal, ,
no abstracts in English
Ohshima, Hiroyuki; Takata, Takashi; Doda, Norihiro; Kikuchi, Shin; Koga, Nobuyoshi*; Deguchi, Yoshihiro*
no journal, ,
Development of multi-level, multi-scenario plant simulation systems has started as a fundamental technology to support mechanism elucidation of phenomena, design optimization, and innovative technology development toward the commercialization of sodium cooled fast reactors. In this presentation, the overall plan of this development project is introduced.
Hamase, Erina; Kikuchi, Norihiro; Doda, Norihiro; Tanaka, Masaaki; Imai, Yasutomo*
no journal, ,
In order to develop numerical analysis that can perform a core-plenum interaction in a commercialized sodium-cooled fast reactor, a subchannel analysis based on a porous body model in CFD code was initially applied to the 37-pin fuel assembly sodium test. As a result of CFD analysis for typical high and low flow conditions, the sodium temperature and the axial velocity at upper heated section in the subassembly were consistent with the ASFRE code that has been validated with experiments data within relative error 1.5%. It was found that a subchannel analysis based on a porous body model in CFD code could apply to the thermal hydraulics analysis in the fuel assembly.
Hamase, Erina; Imai, Yasutomo*; Kikuchi, Norihiro; Doda, Norihiro; Tanaka, Masaaki
no journal, ,
The core-plenum interaction needs to be accurately simulated to evaluate the core coolability by a direct reactor auxiliary cooling system (DRACS) in a sodium-cooled fast reactor. Objective of this research is to develop a multidimensional analysis method using commercial CFD code which can analyze the complex thermal-hydraulic behavior including core-plenum interaction in case of operating various decay heat removal systems in natural circulation conditions. In this report, as the first step, the multi-dimensional analysis method for the fuel assembly applying the concept of the subchannel analysis was developed with the modification in the porous body model of the commercial CFD code that could take into account of the radial permeability in the thermal diffusion term whereas the porosity could only be considered. Through the numerical simulations of the 37 pin-bundle sodium experiment at Re = 52,000, it was indicated the numerical result well agreed with the experiment and applicability of this method was confirmed at the high flow rate condition.
Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Tanaka, Masaaki
no journal, ,
In design study of an advanced sodium-cooled fast reactor, adoption of Fuel Assembly (FA) with Inner DUct Structure (FAIDUS) has been investigated as one of the measures to enhance safety of the reactor. Numerical simulation of FAIDUS using subchannel analysis code named ASFRE for a design tool of FA was previously conducted to confirm its applicability to the thermal-hydraulics analysis in FAIDUS. In this study, numerical simulations of FAIDUS and a typical FA using detailed analysis code named SPIRAL under a high flow rate condition were performed. Numerical results using SPIRAL and ASFRE were well comparable. Applicability of ASFER for a thermal-hydraulic design tool of FAIDUS was potentially indicated in comparison with the results of SPIRAL.
Doda, Norihiro; Hamase, Erina; Kikuchi, Norihiro; Tanaka, Masaaki
no journal, ,
In the conventional analysis in the sodium-cooled fast reactor (SFR) plant design, the physical phenomena occurring in each part of the plant are evaluated individually and the interaction between the phenomena is considered under conservative boundary conditions. Therefore, it can be a conservative evaluation in viewpoint of the total balance in the plant design. In this investigation, the integrated platform wchich can couple the detailed analysis codes with the plant dynamics analysis code in order to consider the interaction between the phenomena has been developed as a support tool for achieving optimized design of the SFR plant. In this report, outlines of the development purpose of the integrated platform, the coupling method of analysis codes, and the future development is described.
Umeda, Ryota; Kurihara, Akikazu; Kikuchi, Shin; Kikuchi, Norihiro; Takata, Takashi; Ohshima, Hiroyuki
no journal, ,
Multi-level, multi-scenario simulation systems have been developed as safety fundamental technology for sodium-cooled fast reactors. In this report, as part of the verification and validation of the sodium fire analysis codes, in order to grasp the transport behavior of combustion generated aerosol during of sodium leak, the results of the transport behavior experiment using simulated particles is presented.
