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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; 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.
Aizawa, Kosuke; Tsuji, Mitsuyo; Kobayashi, Jun; Kurihara, Akikazu; Miyake, Yasuhiro*; Nakane, Shigeru*; Ishida, Katsuji*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 10 Pages, 2022/04
In sodium-cooled fast reactors (SFRs), optimizing the design and operate decay heat removal systems (DHRSs) is important for safety enhancement against severe accidents that could lead to core melting. The natural circulation phenomena in a reactor vessel during operating a DHRS were clarified by conducting water experiments using a 1:10 scale experimental facility (PHEASANT) simulating the reactor vessel of loop-type SFRs. In this study, we investigated the natural circulation phenomena under conditions of operating the dipped-type DHX and RVACS using the results of temperature and particle image velocimetry (PIV) measurements, respectively. Furthermore, the effects of temperature fluctuation on the PIV measurement were quantitatively evaluated.
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.
Tsuji, Mitsuyo; Aizawa, Kosuke; Kobayashi, Jun; Kurihara, Akikazu; Miyake, Yasuhiro*
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 5 Pages, 2020/08
The water experiments using a 1/10 scale experimental apparatus simulating the reactor vessel of SFR were conducted to investigate the natural circulation phenomena in a reactor vessel. In this paper, the natural circulation flow field in the reactor vessel was measured by the Particle Image Velocimetry (PIV) method. The PIV measurement was carried out under the operation of the dipped-type direct heat exchanger (DHX) installed in the upper plenum when 20% of the core fuel fell to the lower plenum and accumulated on the core catcher. From the results of PIV measurement, it was quantitatively confirmed that the upward flow occurred at the center region of the lower and upper plenums. In addition, the downward flows were confirmed near the reactor vessel wall in the upper plenum and through outermost layer of the simulated core in the lower plenum. Moreover, the relationship between the temperature field and the velocity field was investigated in order to understand the natural circulation phenomenon in the reactor vessel. From the above results, it was confirmed that the natural circulation cooling path was established under the dipped-type DHX operation.
Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki
Mechanical Engineering Journal (Internet), 7(3), p.19-00546_1 - 19-00546_11, 2020/06
Fully natural circulation decay heat removal systems (DHRSs) are to be adopted for sodium fast reactors, which is a passive safety feature without any electrical pumps. It is required to grasp the thermal-hydraulic phenomena in the reactor vessel and evaluate the coolability of the core under the natural circulation not only for the normal operating condition but also for severe accident conditions. In this paper, the numerical results of the preliminary analysis for the sodium experimental condition with the PLANDTL-2 are discussed to establish an appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX. From these preliminary analyses, the characteristics of the thermal-hydraulics behavior in the PLANDTL-2 to be focused are extracted.
Tsuji, Mitsuyo; Aizawa, Kosuke; Kobayashi, Jun; Kurihara, Akikazu; Miyake, Yasuhiro*
Proceedings of 14th International Symposium on Advanced Science and Technology in Experimental Mechanics (14th ISEM'19) (USB Flash Drive), 4 Pages, 2019/11
The particle image velocimetry (PIV) was measured in scaled-model water experiments simulating a natural circulation flow field in a sodium-cooled fast reactor vessel. The temperature fluctuation in the natural circulation flow field causes the distribution of the refractive index. Thus, the temperature fluctuation affects the uncertainty of the velocity in the PIV measurement. In this study, the authors evaluated the effects of the temperature fluctuation on the PIV measurement in the natural circulation flow field.
Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05
Decay heat removal system (DHRS) by using the natural circulation without depending on the pump as the mechanical equipment is recognized as one of the most effective methodologies for the sodium-cooled fast reactor from the viewpoint of the safety enhancement. In this paper, the numerical simulation results of the preliminary analysis for the sodium experiment with the apparatus of PLANDTL-2, in which the core and the upper plenum with a dipped-type direct heat exchanger (DHX) were modeled, were discussed, in order to establish appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX.
Tanaka, Masaaki; Ono, Ayako; Hamase, Erina; Ezure, Toshiki; Miyake, Yasuhiro*
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2018 Koen Rombunshu (CD-ROM), 4 Pages, 2018/08
Decay heat removal system (DHRS) by using the natural circulation without depending on the pump as the mechanical equipment is recognized as one of the most effective methodologies for the sodium-cooled fast reactor from the viewpoint of the safety enhancement. The numerical estimation method which can predict thermal hydraulic phenomena in the natural circulation under the plant cooling process by operating the various DHRSs including the severe accident is necessarily required. In this paper, the numerical results of the preliminary analysis for the sodium experiment condition with the apparatus of PLANDTL-2, in which the core and the upper plenum with a dipped-type direct heat exchanger (DHX) were modeled, were discussed, in order to establish an appropriate numerical models for the direct heat exchanger (DHX).
Ono, Ayako; Kurihara, Akikazu; Tanaka, Masaaki; Ohshima, Hiroyuki; Kamide, Hideki; Miyake, Yasuhiro*; Ito, Masami*; Nakane, Shigeru*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04
The water experiment apparatus simulating the thermal hydraulics in a reactor vessel under operating the decay heat removal systems (DHRSs) was fabricated. The theoretical evaluation for similarity and results of basic experiments show applicability for a scale model experiment of a sodium-cooled fast reactor. This paper, moreover, describes the results of flow visualization experiment under operating a dipped-type passive DHX, which is planned to be installed in both a loop type reactor and pool type reactor, and the calculation results using FLUENT comparing with the result of water experiment.
Ono, Ayako; Onojima, Takamitsu; Doda, Norihiro; Miyake, Yasuhiro*; Kamide, Hideki
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.2183 - 2192, 2016/04
Some auxiliary cooling systems to remove the decay heat of the core are under consideration for a sodium-cooled fast reactor, and two of the typical systems are primary reactor auxiliary cooling system (PRACS) and direct reactor auxiliary cooling system (DRACS). In this study, sodium experiments were conducted in order to confirm the applicability of the PRACS and DRACS under a situation assuming the severe accidents with core melting. The plant dynamics test loop was used for these experiments, which contains a simulated core, the PRACS and DRACS. The core melt situation is simulated by shutting off the inlet of subassemblies (S/A). The experimental results revealed the cooling process of the partially/completely inactive S/A and confirmed the long-term heat removal by the PRACS/DRACS.
Tanaka, Masaaki; Miyake, Yasuhiro*
Nihon Kikai Gakkai M&M 2015 Zairyo Rikigaku Kanfarensu Koen Rombunshu (Internet), 3 Pages, 2015/11
A prototype coupling method consisting of the fluid-structure thermal interaction simulation code MUGTHES and the structural thermal stress analysis code FINAS with interface program MUFIN has been developed in order to estimate the thermal fatigue in the SFRs. As a fundamental validation of the coupled method, it was applied to the water experiment for thermal mixing phenomena in a T-junction piping system. In the experiment, thermal interaction between the fluid and the structure made of aluminum installed to the branch pipe side wall was considered. Through the numerical simulations, applicability of the coupled method was confirmed.
Tanaka, Masaaki; Miyake, Yasuhiro*
Mechanical Engineering Journal (Internet), 2(5), p.15-00134_1 - 15-00134_20, 2015/10
In this study, numerical simulation for the water experiment of a T-junction piping system (T-pipe) was carried out to validate the MUGTHES and to investigate the relation between the mechanism of temperature fluctuation generation and the unsteady motion of large eddy structures. In the numerical simulation, the large eddy simulation (LES) approach with standard Smagorinsky model was employed as eddy viscosity model to simulate large scale eddy motion in the T-pipe. As for uncertainty quantification in the validation process, the modified method of the Grid Convergence Index (GCI) estimation based on the least squire version could successfully quantify uncertainty. Through the numerical simulations, it could be found that the thermal mixing phenomena in the T-pipe were caused by the mutual interaction of the necklace-shaped vortex around the wake from the front of the branch jet, the horseshoe-shaped vortex and the Karman's vortex motions in the wake.
Tanaka, Masaaki; Miyake, Yasuhiro*; Karakida, Yasuhisa*
Proceedings of 2nd International Conference on Maintenance Science and Technology (ICMST-Kobe 2014), p.79 - 80, 2014/11
A coupled method of fluid-structure thermal interaction simulation and thermal stress analysis has been developed through the interface program to carry out direct numerical estimation of the thermal fatigue. The prototype method was applied to the thermal mixing phenomena in T-junction Piping System.
Tanaka, Masaaki; Miyake, Yasuhiro*
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 13 Pages, 2014/07
In this study, numerical simulation for the WATLON experiment which was the water experiment of a T-junction piping system (T-pipe) was carried out to validate the MUGTHES and to investigate the relation between the mechanism of temperature fluctuation generation and the unsteady motion of large eddy structures. In the numerical simulation, the large eddy simulation (LES) approach with standard Smagorinsky model was employed as eddy viscosity model to simulate large scale eddy motion in the T-pipe. As for uncertainty quantification in the validation process, the modified method of the Grid Convergence Index (GCI) estimation based on the least squire version could successfully quantify uncertainty. Through the numerical simulations, it was indicated that the fine mesh arrangement could improve the temperature distribution in the wake. It could be found that the thermal mixing phenomena in the T-pipe were caused by the mutual interaction of the necklace-shaped vortex around the wake from the front of the branch jet, the horseshoe-shaped vortex and the Karman's vortex motions in the wake.
Ezure, Toshiki; Miyake, Yasuhiro*; Tobita, Akira; Kimura, Nobuyuki; Kamide, Hideki
JAEA-Research 2012-005, 56 Pages, 2012/05
JAEA-Research-2012-005.pdf:10.06MB
In the design of JSFR, a two-loop cooling system and a compact reactor vessel are employed to achieve the economical improvement. However, these innovative designs lead to the increase of coolant velocity. As the results, strong vortices at the H/L intake may cause the cavitation (vortex cavitation). Therefore, the evaluation of occurrence behavior of vortex cavitation is the important issue for the structural integrity of reactor. In the present study, fundamental water experiments were performed in the cylindrical tank geometry. The water temperature was varied from 10
C to 80C to clarify the influence of kinematic viscosity, 
. The occurrences behaviors of vortex cavitation were evaluated quantitatively by visualization measurement and image analysis. As the results, it was clarified that there was little dependence on under the small conditions, while it was relatively obvious under the large conditions.
Tanaka, Masaaki; Murakami, Satoshi*; Miyake, Yasuhiro*; Ohshima, Hiroyuki
Proceedings of ASME-JSME-KSME Joint Fluids Engineering Conference 2011 (AJK 2011-FED) (CD-ROM), 12 Pages, 2011/07
Authors have been developed a numerical simulation code MUGTHES by using the LES to investigate thermal striping phenomena in FBRs. In this paper, numerical simulations by the MUGTHES in T-junction piping system appear. The boundary conditions are chosen from an existing water experiment in JAEA. Numerical results are validated by comparisons with measured velocity and temperature profiles. Through the numerical simulations, generation mechanism of temperature fluctuation in thermal mixing process is revealed in the relation with the large-scale eddy motion.
Kimura, Nobuyuki; Hayashi, Kenji; Tobita, Akira; Kamide, Hideki; Miyake, Yasuhiro*
JAEA-Research 2009-026, 160 Pages, 2009/10
JAEA-Research-2009-026.pdf:15.97MB
Thermal stratification after a scram is one of main thermal loads of a reactor vessel in sodium cooled fast reactor. Water experiments using an 1/10th scaled model were carried out for an advanced loop type sodium cooled reactor. The reactor vessel is highly compact and has an upper inner structure (UIS), which has a slit in radial direction for fuel handling. The jet from core outlet goes through the UIS slit and makes asymmetric and locally high velocity field in the reactor upper plenum. Steep temperature distribution across the stratification interface and temperature fluctuations were found near the UIS slit. It was revealed that they were resulted from the impingement of the jet through the slit at the interface. Parameter experiments showed the characteristics of thermal stratification interface.
Kimura, Nobuyuki; Nagasawa, Kazuyoshi*; Miyake, Yasuhiro*; Ogawa, Hiroshi; Kamide, Hideki
JNC TN9400 2005-037, 82 Pages, 2005/08
JNC-TN9400-2005-037.pdf:6.54MB
A quantitative evaluation on thermal striping, in which temperature fluctuation due to convective mixing causes high cycle thermal fatigue in structural components, is of importance for reactor safety. The consideration of decay of temperature fluctuation based on the phenomenological mechanism enables a legitimate design with reactor integrity. In this study, we performed a parallel triple-Jet water experiment along a wall. In this experiment, the material of the wall was changed to acrylic resin, type 316 stainless steel and copper in order to evaluate an effect of thermal properties in wall materials on the transfor characteristics of temperature fluctuation from fluid to structure. The fluid temperature was measured by a movable thermocouple-tree and the structural temperature was measured by thermocouples embedded on the wall surface. The effect of the wall material on the temperature fluctuation in fluid was not observed at the further position from the wall. In the vicinity of the wall, on the other hand, the temperature fluctuation intensity in fluid decreased as the thermal diffusivity of the wall material was large. The temperature fluctuation intensity in structure was small as the thermal diffusivity of the wall material was large. For each wall material, the heat transfer coefficient was obtained from a transfer function between fluid and structure temperature fluctuations. It was seen that the heat transfer coefficient was large as the thermal diffusivity of material was large. In this experiment, furthermore, the dependence of Nusselt number on Reynolds number was dose to the existing correlation in every materials.
