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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.
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.
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.
Tamai, Hidesada; Yoshida, Hiroyuki; Masuko, Kenji*; Akimoto, Hajime
Proceedings of 4th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-4), p.230 - 236, 2004/12
no abstracts in English
Kureta, Masatoshi; Tamai, Hidesada
Proceedings of 5th International Conference on Multiphase Flow (ICMF 2004) (CD-ROM), 10 Pages, 2004/06
3D void fraction distribution of boiling flow in a tight-lattice 7-rod bundle was measured by neutron radiography 3D computed tomography (neutron tomography) to investigate the flow characteristics in tight-lattice rod bundles and to verify the numerical analysis codes. The test section simulates the fuel rod bundle of the RMWR and consists of 7 heater rods with gap of 1.0mm and with diameter of 12.0mm. In this paper, the neutron tomography system, experiments and comparison of the measured data with a subchannel analysis code, COBRA-TF, are reported. It was found from this experiment that water layer which surrounds the heater rod becomes thick between rods, narrow region, and steam accumulates at the center region among three rods. COBRA-TF code overestimates the void fraction in a tight-lattice bundle compared with the present data.
Okubo, Tsutomu; X.Jiang*; Araya, Fumimasa; Ochiai, Masaaki
JAERI-Research 98-042, 49 Pages, 1998/08
JAERI-Research-98-042.pdf:1.58MB
no abstracts in English
Okubo, Tsutomu; Araya, Fumimasa; Iwamura, Takamichi; Kusunoki, Tsuyoshi
Fourth Int. Seminar on Subchannel Analysis (ISSCA-4), p.267 - 286, 1997/00
no abstracts in English
Kikuchi, Norihiro; Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki
no journal, ,
no abstracts in English
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; Yokoyama, Kenji; Tanaka, Masaaki; Takata, Takashi; Ohshima, Hiroyuki
no journal, ,
JAEA has developed a multi-level plant simulation system to enhance a safety assessment technology for sodium cooled fast reactor (SFR). The system, in which high-efficient plant dynamics analyses with a one-dimensional system code and high-accurate local phenomena analyses with multi-dimensional codes are coupled, may apply to the various design options of SFR. We have developed two coupling methods with plant dynamics analysis code for fuel assembly thermal-hydraulics analysis code and for neutronics analysis code, respectively. Using the coupling methods, we performed analyses on the thermal-hydraulics coupling problem between whole core and a fuel assembly and on the nuclear-thermal coupling problem under unprotected conditions. The coupled analysis results were compared with the 1D code results of the same problems. The results showed that each coupling method was validated.
Doda, Norihiro; Yokoyama, Kenji; Tanaka, Masaaki; Takata, Takashi; Ohshima, Hiroyuki
no journal, ,
A multi-level plant simulation system has developed to enhance a safety assessment technology for sodium cooled fast reactor (SFR). The system, in which high-efficient analyses with a one-dimensional plant dynamics analysis code and high-accurate local phenomena analyses with multi-dimensional codes are coupled, may apply to the various design options of SFR. For a validation of the system, the analysis on the experimental fast reactor EBR-II with the SHRT-45R test condition and a virtual condition of control rod withdrawn condition were performed. From the comparison of the analysis results and the test results, it was confirmed that the system could perform the evaluation by changing the level of detail of the analysis model according to the intended use.
Kikuchi, Norihiro
no journal, ,
In the design study of a sodium-cooled fast reactor (SFR) in JAEA, the adoption of a fuel assembly (FA) with an inner duct structure called FAIDUS has been investigated to enhance safety of SFR. Due to the asymmetric layout of fuel rods by addition of the inner duct, the velocity and temperature distributions in FAIDUS may differ from that in typical FA. It is need to confirmed that applicability of the subchannel analysis code named ASFRE which has been developed as a FA design tool, to thermal hydraulics analysis in FAIDUS. Because the reference data on the thermal hydraulics in FAIDUS have not been obtained by the mock-up experiment, 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 occurring around the inner duct between the numerical results by ASFRE and SPIRAL.
