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Hamada, Noriaki*; Shiina, Koji*; Fujimata, Kazuhiro*; Hayakawa, Satoshi*; Watanabe, Osamu*; Yamano, Hidemasa
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00
In a sodium-cooled fast reactor, a vortex cavitation evaluation methodology was developed to predict a possible cavitation generated by vortex at the center of accelerating flow. This methodology was applied to a scaled model experiment, leading to the prospect that the cavitation can be predicted.
Aizawa, Kosuke; Nakanishi, Shigeyuki; Yamano, Hidemasa; Kotake, Shoji; Hayakawa, Satoshi*; Watanabe, Osamu*; Fujimata, Kazuhiro*
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 7 Pages, 2008/11
To evaluate the flow-induced vibration in the actual-sized pipings of JSFR, computer simulation is necessary. In this study, as the first step, sensitivity analysis of turbulence flow models for unsteady short-elbow pipe flow has been carried out with the STAR-CD thermal-hydraulic simulation code. Through the sensibility analysis, the objective of this study is to propose the best analysis models which can reproduce the unsteady characteristics obtained in the 1/3-scale test results with 9.2 m/s of main flow. In this study, to take into account anisotropic characteristics of turbulence, two turbulent flow models were used: large eddy simulation (LES) and Reynolds stress model (RSM). The both validated simulations have reproduced flow separation region and periodic vortex shedding. The simulation results with both models were compared with power spectrum densities of pressure fluctuations which were used in the pipe vibration evaluation. Only the RSM simulation with the best combination has reproduced the pressure-fluctuation power spectrum densities, which were characterized by a peak frequency of 10 Hz in the 1/3 test with 9.2 m/s.
Aizawa, Kosuke; Yamano, Hidemasa; Kotake, Shoji; Fujimata, Kazuhiro*
Proceedings of 7th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-7) (CD-ROM), 14 Pages, 2008/10
The present study has numerically investigated using the STAR-CD code the fluid-flow characteristics within a large-diameter piping with a short-radius elbow, which is adopted in an advanced large-sized sodium-cooled fast reactor (named JSFR). This study reports the result of numerical steady-state calculations of the 1/3-scale experiment with 9.2m/s of velocity performed at the first step. Since the experiments have revealed that dominant fluctuating pressures were generated at the boundary of flow separation and reattachment point on the pipe wall, this study focused on the flow separation size as one of the flow characteristics. The calculation has reproduced the flow characteristics, such as the measured velocity profile in the flow separation region, by specifying appropriate analytical models and conditions. With the validated models, the effect of the coolant viscosity has also been investigated as well as the piping scale. In order to examine the disturbance at the piping inlet, the flow dynamics within the reactor vessel were also calculated by modeling an entire upper sodium plenum region including various components within the reactor vessel in the JSFR design. This upper plenum calculation had to reduce the spatial resolution within the hot-leg piping because of numerous computational meshes needed in this calculation. The plenum calculation has shown several vortexes and flow distortion at the hot-leg inlet. The hot-leg inlet flow condition obtained in the plenum calculation was interpolated for the calculation simulating the hot-leg piping, where the spatial resolution was better than in the plenum calculation. The numerical calculation under the reactor condition involving the inlet disturbance has indicated the flow separation size became smaller than that in no disturbance case. This calculation implies that the inlet disturbance may play an important role to mitigate the flow-induced vibration force in the flow separation region.
Fujii, Tadashi; Chikazawa, Yoshitaka; Konomura, Mamoru; Kamide, Hideki; Kimura, Nobuyuki; Nakayama, Okatsu; Ohshima, Hiroyuki; Narita, Hitoshi*; Fujimata, Kazuhiro*; Itooka, Satoshi*
JAEA-Research 2006-017, 113 Pages, 2006/03
JAEA-Research-2006-017.pdf:14.98MB
A conceptual design study of the sodium-cooled fast reactor is in progress in the Feasibility Study on Commercialized Fast Reactor Cycle Systems. Reduced scale water experiments are being performed in order to clarify the flow pattern in the upper plenum of the reactor which has higher velocity condition than the past design. In this report, the hydraulic analyses of the water experiments using the general-purpose thermal hydraulic analysis program were executed; and the applicability to evaluation of flow pattern and vortex cavitations for the designed reactor was examined. (1) Steady-state analyses under the Froude number similar condition were carried out for the 1/10th reduced scale plenum experiments. Analyses results reproduced the characteristic flow patterns in the upper plenum, such as gushed flow from the inside of the upper internal structure to reactor vessel wall and the jet flow from the slit of the upper internal structure. Further, it was confirmed that the calculated flow pattern of a designed reactor system agreed with that of the water experiment qualitatively. Moreover, the influence which setting of numerical solution and boundary condition etc. in analyzing causes to flow pattern in the plenum became clear. (2) The distribution of the vortices under the dipped plate region in the 1/10th plenum model was evaluated using the prediction method of a submerged vortex which is based on the stretching vortex theory. In case of the same velocity condition as the reactor, it identified the two vortices which were sucked into the hot leg piping from the cold leg piping wall as the submerged vortex cavitations. From this analysis result, it confirmed that the submerged vortex cavitations, which may occur in the reactor upper plenum steadily, could be identified using this prediction method.
Itooka, Satoshi*; Fujimata, Kazuhiro*
JNC TJ9440 2003-001, 286 Pages, 2003/03
JNC-TJ9440-2003-001.pdf:9.23MB
In Japan Nuclear Cycle Development Institute, the improvement of analysis method for overheating tube rapture was studied for the accident of sodium-water reactions in the steam generator of a fast breeder reactor and the evaluation of heat transfer condition in the tube were carried out based on study of critical heat flux (CHF) and post-CHF heat transfer equation in Light Water Reactors. In this study, the improvement of blow down model for the LEAP code was carried out taking into consideration the above-mentioned evaluation of heat transfer condition. Improvements of the LEAP code were following items. Calculations and verification were performed with the improved LEAP code in order to confirm the code functions. (1)The addition of critical heat flux (CHF) by the formula of Katto and the formula of Tong. (2)The addition of post-CHF heat transfer equation by the formula of Condie-Bengston Ⅳ and the formula of Groeneveld 5.9. (3)The physical properties of the water and steam are expanded to the critical conditions of the water. (4)The expansion of the total number of section and the improvement of the input form. (5)The addition of the function to control the valve setting by the PID control model.
Fujimata, Kazuhiro*; Nagashima, Hideaki*; Sukekawa, Masayuki*; Shibamoto, Hiroshi; Inoue, Kazuhiko*; Kasahara, Naoto
no journal, ,
no abstracts in English
Aizawa, Kosuke; Yamano, Hidemasa; Uto, Nariaki; Kotake, Shoji; Watanabe, Osamu*; Fujimata, Kazuhiro*
no journal, ,
A conceptual design study of a large-scale sodium-cooled fast reactor adopts a two-loop primary cooling system with large-diameter piping in order to reduce plant construction cost. In this design, one of issues is a flow-induced vibration behavior of the piping under a high Reynolds number of 10
order. To evaluate the piping integrity, it is necessary to obtain power spectrum densities of pressure fluctuations on the piping wall by a numerical unsteady flow simulation. In this study, the numerical simulation capability of Reynolds stress model and large-eddy simulation in the STAR-CD code has been investigated using the 1/3-scale hot-leg test data. Through the sensitivity analysis, the Reynolds stress model with appropriate analytical models has shown the best applicability to flow dynamics simulation in the short-elbow pipe.
Hamada, Noriaki*; Fujimata, Kazuhiro*; Shiina, Koji*; Hayakawa, Satoshi*; Watanabe, Osamu*; Yamano, Hidemasa
no journal, ,
A computational method for predicting vortex cavitation based on the theory of vortex streching was developed to predict possible cavitation generated by vortex at the center of accelerating swirl flow in the reactor vessel in a sodium-cooled fast reactor. This method was applied to a scale model test of a commercial fast reactor, leading to feasibility of this method that can predict the cavitation.
Abe, Takashi*; Nakahara, Hirotaka*; Aoyagi, Mitsuhiro; Fujimata, Kazuhiro*
no journal, ,
no abstracts in English
