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Matsushita, Kentaro; Ezure, Toshiki; Tanaka, Masaaki; Imai, Yasutomo*; Fujisaki, Tatsuya*; Sakai, Takaaki*
Nuclear Engineering and Design, 432, p.113785_1 - 113785_16, 2025/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Establishing an evaluation method for the gas entrainment (GE) of argon cover gas due to surface vortices is required in terms of safety design of sodium-cooled fast reactors. To modify the evaluation model in an in-house evaluation tool for GE, StreamViewer, a modified evaluation model on the pressure distribution along the vortex center line (PVL model) was proposed to identify the vortex center lines by connecting continuous vortex center points from the suction port to the surface and evaluate gas core length based on the balance between the hydrostatic pressure and the pressure decrease distribution along the vortex center line. PVL model was applied the three-dimensional numerical analysis results for the experiments where a plate induced unsteady traveling vortices in the open channel flow. Consequently, the GE evaluation using StreamViewer with PVL model could reproduce the relation between the inlet flow velocity and the gas core length in the unsteady vortex flow experiments.
Alzahrani, H.*; Sakai, Takaaki*; Matsushita, Kentaro; Ezure, Toshiki; Tanaka, Masaaki
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.1262 - 1275, 2023/08
Development of evaluation method for cover gas entrainment by vortices generated at free surface in upper plenum of sodium-cooled fast reactor is required, and an evaluation method by predicting vortices from flow velocity distribution obtained by CFD analysis is developed. In this study, Adaptive Mesh Refinement (AMR) method is examined to improve efficiency of CFD analysis. Initial mesh was refined with two indexes: the first index (Index-1) is when the second invariant, Q, of velocity gradient tensor is negative and the second one (Index-2) is pressure gradient index added to Index-1. As a result of applying AMR method to unsteady vortices system with a flat plate and performing transient analyses with refined meshes, the result of pressure distribution and velocity around the flat plate in mesh using Index-2 was similar to the result of all refined mesh. It was also confirmed that vortices generation and growth was better simulated by refining meshes around separation area.
Song, K.*; Ito, Kei*; Ito, Daisuke*; Odaira, Naoya*; Saito, Yasushi*; Matsushita, Kentaro; Ezure, Toshiki; Tanaka, Masaaki
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
Gas entrainment (GE) phenomena caused by a free surface vortex may cause the disturbance in core power of sodium-cooled fast reactor (SFR). For this reason, the entrained gas flow rate by the GE should be evaluated accurately for the practical safety design of SFRs. In this study, for the purpose of examining the applicability of CFD for the accurate evaluation of GE phenomena, a CFD is applied to the simulation of the free surface vortex and accompanied GE phenomena in a cylindrical vessel with a suction pipe, and the CFD results and the experimental data of the GE are compared. As a result, the CFD and experiments show similar two-phase flow pattern inside the suction pipe, and the shape of the gas core at the free surface is also very similar. Therefore, it is confirmed that the CFD can predict the GE phenomena triggered by a free surface vortex properly and accurately within the acceptable error range.
Alzahrani, H.*; Matsushita, Kentaro; Sakai, Takaaki*; Ezure, Toshiki; Tanaka, Masaaki
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 6 Pages, 2022/10
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. An evaluation method by predicting vortices from flow velocity distribution obtained by 3D CFD analysis is developed, and Adaptive Mesh Refinement (AMR) method is examined to improve efficiency of CFD analysis is examined. In this study, mesh refinement with two conditions were examined. The first one is to use negative second invariant of velocity gradient tensor, Q, and the second one is to use pressure gradient condition with Q
0. As a result of applying AMR method to unsteady vortices system with a flat plate, the mesh near stagnation area around flat plate was refined in the latter condition compared with the former. Transient analyses were performed with refined mesh by AMR method, the result of mesh using the latter condition was closer to the result of all refined mesh with pressure distribution near flat plate.
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.
Torikawa, Tomoaki*; Odaira, Naoya*; Ito, Daisuke*; Ito, Kei*; Saito, Yasushi*; Matsushita, Kentaro; Ezure, Toshiki; Tanaka, Masaaki
Konsoryu, 36(1), p.63 - 69, 2022/03
On free surface of a sodium cooled fast reactor, gas entrainment can be caused by free surface vortices, which may result in disturbance in core power. It is important to develop an evaluation model to predict accurately entrained gas flow rate. In this study, entrained gas flow rate a simple gas entrainment experiment is conducted with focusing on effect of pressure difference between upper and lower tanks. Pressure difference between upper and lower tanks are controlled by changing gas pressure in lower tank. As a result, it is confirmed that the entrained gas flow rate increases with increasing pressure difference between upper and lower tanks. By visualization of swirling annular flow in suction pipe, it is also observed that pressure drop in suction pipe increases with increase in entrained gas flow rate, which implies that entrained gas flow rate can be predicted by evaluation model based on pressure drop in swirling annular flow region.
Uchida, Mao*; Alzahrani, H.*; Shiono, Mikihito*; Sakai, Takaaki*; Matsushita, Kentaro; Ezure, Toshiki; Tanaka, Masaaki
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
Gas entrainment from cover gas is one of key issues for sodium-cooled fast reactors design to prevent unexpected effects to core reactivity. A vortex model based evaluation method has been developed to evaluate the surface vortex gas core growth at the free surface in the reactor vessel. In this study, water experiments were performed to clarify the prediction accuracy for the vortex gas core growth during the vortex drift motion using a circulating water tunnel with an open flow channel test section. Gas core growth were predicted by applying the evaluation method to the numerical analyses performed in the same geometry of the experiments, and compared with the experimental results. It was observed the gas core growth became large at downstream region where downward velocity became large in experiment. However, the gas core length which were predicted from numerical result showed a discrepancy with the experimental result on the peak position and an overestimation of peak value.
Matsushita, Kentaro; Ito, Kei*; Ezure, Toshiki; Tanaka, Masaaki
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 9 Pages, 2019/05
A numerical simulation code named SYRENA has been developed in JAEA to analyze the behavior of entrained bubbles and dissolved gas in the primary coolant of sodium-cooled fast reactor (SFR). In the present study, a flow network model of SYRENA to a hypothetical pool type reactor was developed and the non-condensable gas behavior was investigated through the comparison with that in the loop type reactor. The effect of the dipped-plate (D/P) tentatively introduced into the pool-type reactor on the gas behavior was investigated through the parametric analyses about the sodium exchange flow rate through the D/P and the gas entrainment rate at the free surface. It was suggested that the increase in the exchange flow rate through the D/P doesn't always work to decrease the bubble volume in the primary coolant system.
Sugimoto, Taro*; Saito, Shimpei*; Kaneko, Akiko*; Abe, Yutaka*; Uchibori, Akihiro; Ohshima, Hiroyuki
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 7 Pages, 2018/07
A computational fluid dynamics code for a sodium-water reaction phenomenon in a steam generator of sodium-cooled fast reactors has been developed. In order to provide the data for validation of this code, the visualization experiment on liquid droplet entrainment in the high-pressure air jet submerged in the water pool was carried out. The experiment successfully elucidated the behavior, such as atomization of the relatively large diameter liquid droplet generated from the gas-liquid interface.
Ito, Kei; Koizumi, Yasuo; Ohshima, Hiroyuki; Kawamura, Takumi*
Mechanical Engineering Journal (Internet), 3(3), p.15-00671_1 - 15-00671_9, 2016/06
Uchibori, Akihiro; Ohshima, Hiroyuki
Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.533 - 544, 2015/08
For assessment of the wastage environment under tube failure accident, a mechanistic computer code called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction has been developed. In this study, applicability of the SERAPHIM code including the numerical model for liquid droplet entrainment and transport was investigated through the analysis of the basic experiment and the experiment under actual condition of the steam generator. In the analysis of the basic experiment, the calculated pressure variation during liquid droplet entrainment was consistent with the experimental result. In the analysis of the actual condition, the calculated temperature distribution agreed with the measurement result well. The region with higher impingement velocity of the liquid droplet was close to the wastage region confirmed in the experiment. It was demonstrated that the SERAPHIM code could predict the wastage environment under the actual condition.
Ito, Kei; Koizumi, Yasuo*; Ohshima, Hiroyuki; Kawamura, Takumi*
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
The authors are developing a high-precision CFD code with an interface tracking method to simulate the gas entrainment (GE) phenomena in sodium-cooled fast reactors (SFRs), which might be caused by a highly-intensified free surface vortex. The GE in SFRs is characterized by an elongated interfacial dent along the vortex core and the bubble pinch-off at the tip of the dent. To simulate this complicated phenomenon, our simulation code has physics-basis algorithms which model accurately the interfacial dynamic behavior, the pressure jump condition at an interface and the surface tension. Several verification problems have been already solved and the accuracy of each individual algorithm is confirmed. In this paper, a basic experiment of the GE is simulated to validate the developed code. The simulation result of the entrained flow rate shows comparable value to the experimental data, that is, our simulation code is considered applicable to the evaluation of the GE in SFRs.
Ito, Kei; Ohno, Shuji; Koizumi, Yasuo*; Kawamura, Takumi*
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 10 Pages, 2014/12
Ito, Kei; Ezure, Toshiki; Ohshima, Hiroyuki
Nihon Kikai Gakkai Rombunshu (Internet), 80(818), p.FE0299_1 - FE0299_9, 2014/10
A vortex is considered as one of significant phenomena which may cause gas entrainment (GE) and/or vortex cavitation in sodium-cooled fast reactors. In this study, a new vortex model with realistic axial velocity distribution is proposed. As the verification, the new vortex model is applied to the evaluation of a simple vortex experiment, and shows good agreements with the experimental data in terms of the circumferential velocity distribution and the free surface shape. In addition, it is confirmed that the Burgers vortex model fails to calculate accurate velocity distribution with the assumption of uniform axial velocity. However, the calculation accuracy of the Burgers vortex model can be enhanced close to that of the new vortex model in consideration of the effective axial velocity which is calculated as the average value only in the vicinity of the vortex center.
Koizumi, Yasuo*; Ote, Naosuke*; Kamide, Hideki; Ohno, Shuji; Ito, Kei
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 7 Pages, 2014/07
; Kukita, Yutaka*; Tsuji, Yoshiyuki*
Eighth Int. Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), 1, p.40 - 47, 1997/00
no abstracts in English
Yonomoto, Taisuke
JAERI-Research 96-024, 154 Pages, 1996/05
JAERI-Research-96-024.pdf:5.87MB
no abstracts in English
JAERI-Data/Code 96-004, 109 Pages, 1996/02
JAERI-Data-Code-96-004.pdf:2.3MB
no abstracts in English
Nakamura, Hideo; Kukita, Yutaka;
Journal of Nuclear Science and Technology, 32(7), p.641 - 652, 1995/07
Times Cited Count：8 Percentile：58.70(Nuclear Science & Technology)no abstracts in English
Nakamura, Hideo; Kukita, Yutaka;
Journal of Nuclear Science and Technology, 31(2), p.113 - 121, 1994/02
Times Cited Count：1 Percentile：17.67(Nuclear Science & Technology)no abstracts in English
