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Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki
Nuclear Technology, 210(5), p.814 - 835, 2024/05
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)In the study of safety enhancement on advanced sodium-cooled fast reactor, it is essential to clarify the thermal-hydraulics under various operation conditions in a fuel assembly (FA) with the wire-wrapped fuel pins to assess the structural integrity of the fuel pin. A finite element thermal-hydraulics analysis code named SPIRAL has been developed to analyze the detailed thermal-hydraulics phenomena in a FA. In this study, the numerical simulations of the 37-pin bundle sodium experiments at different Re number conditions, including a transitional condition between laminar and turbulent flows and turbulent flow conditions, were performed to validate the hybrid turbulence model equipped in SPIRAL. The temperature distributions predicted by SPIRAL was consistent with those measured in the experiments. Through the validation study, the applicability of the hybrid turbulence model in SPIRAL to thermal-hydraulic evaluation of sodium-cooled FA in the wide range of Re number was confirmed.
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.
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.
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.
Matsushita, Kentaro; Ezure, Toshiki; Imai, Yasutomo*; Fujisaki, Tatsuya*; Tanaka, Masaaki
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 8 Pages, 2022/08
Development of evaluation method for cover gas entrainment (GE) by vortices generated at free surface in upper plenum of sodium-cooled fast reactor (SFR) is required. GE evaluation tool, named StreamViewer, based on method using numerical results of three-dimensional computational fluid dynamics analysis for loop-type SFRs has been developed. In this study, modification of evaluation method of StreamViewer to rationalize conservativeness in evaluation results was examined by identifying vortex center lines and calculating three-dimensional distribution of pressure decrease along vortex center lines. The applicability of modified method was checked using water experimental result in rectangular open channel where unsteady vortices are generated. As the result, it was indicated that evaluation results on gas core depth which were excessive in current method were improved in modified method, and it is confirmed that modified method may discriminate onset of GE with appropriate criteria.
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As studied by X-ray magnetic circular dichroism and resonant inelastic X-ray scatteringSuzuki, Hakuto*; Zhao, G.*; Okamoto, Jun*; Sakamoto, Shoya*; Chen, Z.-Y.*; Nonaka, Yosuke*; Shibata, Goro; Zhao, K.*; Chen, B.*; Wu, W.-B.*; et al.
Journal of the Physical Society of Japan, 91(6), p.064710_1 - 064710_5, 2022/06
Times Cited Count：0 Percentile：0(Physics, Multidisciplinary)Matsushita, Kentaro; Ezure, Toshiki; Imai, Yasutomo*; Fujisaki, Tatsuya*; Tanaka, Masaaki
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2021-Nendo Koen Rombunshu (Internet), 4 Pages, 2021/08
For evaluation of gas entrainment phenomenon at free surface in reactor vessel of sodium-cooled fast reactor, the gas entrainment evaluation tool named "Stream Viewer" has been developed. In Stream Viewer, depth of surface vortex dimple is predicted by calculating pressure decrease at the vortex center using velocity distribution around the vortex and Burgers vortex model. In this report, a method to identify continuous vortex center lines from a velocity distribution is newly developed. It becomes possible to evaluate three-dimensional distribution of pressure decrease along vortex center line. Then, the method is validated by applying Stream Viewer to an open channel experiment. As the result, it was confirmed that vortex center lines were successfully identified by the improved Stream Viewer. Moreover, it was also shown that the evaluation accuracy of gas entrainment was expected to be improved by considering distribution of pressure decrease along vortex center line.
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.
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.
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.
Ito, Kei*; Ito, Daisuke*; Saito, Yasushi*; Ezure, Toshiki; Matsushita, Kentaro; Tanaka, Masaaki; Imai, Yasutomo*
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.6632 - 6642, 2019/08
In this paper, a mechanistic model is proposed to calculate the entrained gas flow rate by a free surface vortex. The model contains the theoretical equation of transient gas core elongation and the empirical equation of critical gas core length for gas bubble detachment. Based on those two equations, the entrained gas flow rate is calculated as the portion of the gas core elongated beyond the critical gas core length per unit time. Then, the mechanistic model was applied to the calculation of the entrained gas flow rate in a simple water experiment. As a result, it is confirmed that the entrained gas flow rate grows rapidly when the liquid (water) flow rate, which determine the strength of a free surface vortex, exceeds a certain threshold value.
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; 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.
Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki; Imai, Yasutomo*
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 12 Pages, 2016/10
A sodium-water coupled thermal-hydraulics simulation code TSG has been developed for numerical estimation of three-dimensional thermal-hydraulic phenomena in the straight-tube steam generator. The water analysis module was developed by using the parallel channel model of heat transfer tubes, and the sodium analysis module was developed by using porous body approach. As the first step of validation, simulation results by TSG were compared with the measured data of 1MWt SG experiments under steady state conditions. Through the numerical simulation, the coupled simulation method used in TSG was validated and applicability of TSG to simulate thermal-hydraulics of the straight tube SG in the steady state was confirmed.
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.
Yoshikawa, Ryuji; Ohshima, Hiroyuki; Tanaka, Masaaki; Imai, Yasutomo*
JAEA-Data/Code 2014-034, 84 Pages, 2015/03
JAEA-Data-Code-2014-034.pdf:2.35MB
TSG (Three-dimensional Thermal-hydraulics Analysis Code for Steam Generators) is being developed for analyses of thermal hydraulics in double wall straight tube steam generator of Fast Breeder Reactor. TSG code is a thermal hydraulics simulation system which couples sodium side three dimensional simulation with water side multi-channel simulation. The three dimensional flow field in the sodium side is simulated by a commercial code FLUENT with porous media model. The multi-channel two-phase flow is simulated by an in-house module with drift-flux model. The sodium side simulation is coupled with the water side simulation by the transmission of heat transfer rate through the heat transfer tube. This report presents a description of the computational models, input and output as the user's manual of TSG water side module.
Okayasu, Satoru; Nishio, Taichiro; Ono, Masao; Mashimo, Tsutomu; Tanaka, Yasutomo*; Iyo, Akira*
Physica C, 445-448, p.245 - 248, 2006/10
Times Cited Count：1 Percentile：6.63(Physics, Applied)Vortex imagings of Tl-2223 thin film are achieved below Tc with a scanning SQUID microscope. Vortex arrangements are almost the same just below Tc, indicating the existence of strong pinning centers in the sample. The origin of the strong pinning centers comes from the morphological inhomogeneity on the surface.
Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki; Imai, Yasutomo*
no journal, ,
In this study, numerical simulation code for the straight-tube type steam generator of the sodium-cooled fast reactor has been developed. Since accurate evaluation of three dimensional distributions of temperature and velocity is required to estimate the structural integrity of the water-steam tubes in the steam generator, a simulation code TSG coupled with the multi-channels water-steam system analysis module and multi-dimensional sodium thermal-hydraulics analysis module has been constructed. At this moment, the numerical models for the multiple plugged tubes conditions were developed. Through the numerical simulations with plugged tubes condition at tentative boundary conditions, characteristics of the temperature distributions were evaluated and the capability of the numerical models was confirmed.
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.
