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Eguchi, Yuzuru*; Murakami, Takahiro*; Ohshima, Hiroyuki; Yamano, Hidemasa; Kotake, Shoji
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 8 Pages, 2008/11
The unsteady turbulent flow in a short-elbow pipe to be employed in a Japanese sodium-cooled fast breeder reactor was computed to examine the fundamental features of the flow, especially, pressure fluctuation to cause unsteady fluid force on the pipe. An FEM-based large-eddy simulation code, named SMART-fem, was used for the computation. The results at Re=3.2
10
and 1.210
show that two separation regions exist on the inner urvature of the elbow around 45-degree (middle of elbow) and 90-degree (end of elbow) positions. The statistical quantities of pressure fluctuation such as deviation, skewness and flatness were computed and analyzed, showing that there exist two symmetric regions of significant pressure fluctuation on the wall of inner curvature of the elbow. It has turned out that the pressure loss coefficient of the elbow pipe agrees well among the computation, experiment and authoritative reference data.
Iwamoto, Yukiharu*; Yasuda, Kazunori*; Sogo, Motosuke*; Yamano, Hidemasa; Kotake, Shoji
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 6 Pages, 2008/11
Pressure measurement, laser Doppler velocimetry (LDV) and flow visualization were carried out using the 1/10-scale model of a hot leg piping installed in a Japanese sodium-cooled fast breeder reactor. LDV measurement with Reynolds number of 50000 showed the following results: (1) A flow separation was confirmed in the region between 45 degrees from the elbow inlet and 0.3 times of pipe diameter downstream of the elbow. (2) There appeared two kinds of fluctuations in the present study. In the case of Reynolds number of 320000, it was found that the height of the flow separation downstream of the elbow became smaller, since the inertia of the flow became superior to the inverse pressure gradient.
Nakatsuka, Toru; Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 4 Pages, 2008/11
In order to confirm thermal safety of the supercritical-water-cooled reactor (SCWR), it is important to assess the thermal hydraulics in rod bundles of the core. In the present study, the three-dimensional two-fluid model analysis code ACE-3D, which has been developed in JAEA for the two-phase flow thermal-hydraulics of light water reactors, was improved to handle the thermal hydraulic properties of water at supercritical region. Heat transfer experiments of supercritical water flowing in a vertical annular channel around a heater pin, which simulates core flow around a fuel rod, were analyzed with the ACE-3D to assess the prediction performance of the code. As a result, it was confirmed that the calculated wall surface temperature agreed with the measured results and the code is applicable to prediction of heat transfer of supercritical water in the system that simulates the SCWR core. To improve prediction accuracy for heat transfer deterioration is a subject for future study.
Hirota, Kazuo*; Ishitani, Yoshihide*; Nishida, Keigo*; Sago, Hiromi*; Xu, Y.*; Yamano, Hidemasa; Nakanishi, Shigeyuki; Kotake, Shoji
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 8 Pages, 2008/11
CFD simulation using the Reynolds stress model was performed to evaluate turbulence-induced forces on the piping. The turbulence energy with the CFD simulation was compared with pressure fluctuation distributions obtained by the test with a 1/3 scale elbow simulating the JSFR hot-leg piping. The profile of turbulence energy was good agreement with that of the pressure fluctuation. The magnitude of pressure fluctuation can also be estimated from calculated turbulence energy multiplied by a certain coefficient. In the vibration analysis, the power spectrum density (PSD) of the pressure fluctuation was derived from the measured normalized PSD multiplied by the coefficient. The vibration analysis method was proposed based on the PSDs derived by the above procedure and correlation lengths. The analysis results of vibration response showed good agreement with the flow-induced-vibration test results, thereby it can be said that the vibration analysis method developed in this study is valid.
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.
Yoshida, Hiroyuki; Nakatsuka, Toru; Suzuki, Takayuki*
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 6 Pages, 2008/11
Kamide, Hideki; Aizawa, Kosuke; Oshima, Jun*; Nakayama, Okatsu*; Kasahara, Naoto
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 8 Pages, 2008/11
Development of advanced loop type sodium cooled fast reactor is under going. An upper internal structure (UIS) has a radial slit to reduce the reactor vessel diameter. This UIS slit allows a high velocity from the core fuel subassemblies and influences the gas entrainment in the reactor vessel and also the delayed neutron precursor sampling for a failed fuel detection and location system. Then flow visualization and velocity measurements were carried out in an 1/10 scale water test model. The velocity measurement using particle image velocimetry showed that velocity in the slit region was accelerated at the heights of the UIS horizontal plates and kept higher value at the middle height of the upper plenum. Numerical simulation using a commercial CFD code was also carried out for this complex geometry of UIS to know adequate simulation method. The comparisons of velocity profiles in the UIS between the experiment and analysis showed good agreements.
Ohno, Shuji; Oki, Hiroshi*; Sugahara, Akihiro*; Ohshima, Hiroyuki
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 7 Pages, 2008/11
Validation study of numerical simulation method is in progress for thermal stratification phenomena in a reactor vessel upper plenum of advanced sodium-cooled fast reactors. This paper describes the current status of the study using two kinds of thermal stratification experiments and commercial CFD codes STAR-CD, FLUENT, and an in-house code AQUA.
Takeda, Takeshi; Asaka, Hideaki*; Watanabe, Tadashi; Nakamura, Hideo
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 8 Pages, 2008/11
Two LSTF experiments were conducted for OECD/NEA ROSA Project simulating PWR 0.5% cold leg small break LOCA. Steam generator (SG) secondary-side depressurization was performed by fully opening the relief valves at 10 minutes after safety injection signal with or without non-condensable gas (air) inflow from accumulator tanks with total failure of high pressure injection system. Further assumptions were made to conduct enhanced SG depressurization by fully opening the safety valves when the primary pressure decreased to 2 MPa and no actuation of low pressure injection system, both to well observe natural circulation (NC) phenomena at low pressures. The primary depressurization rate decreased when non-condensable gas started to enter primary loops because of degradation in the condensation heat transfer in SG U-tubes, while two-phase flow NC has continued even after non-condensable gas inflow. Asymmetric NC behaviors appeared between two loops due probably to different number of forward flow SG U-tubes which would have been under influences of non-condensable gas. Post-test analyses by using JAEA-modified RELAP5/MOD3.2.1.2 code indicated that the code has remaining problems in proper prediction of primary loop flow rate and SG U-tube liquid level behaviors especially after non-condensable gas inflow. The improvement of the condensation heat transfer model under non-condensable gas mixture condition and the SG U-tube model may be necessary for correct analysis of the LSTF SG depressurization transients.
Ezure, Toshiki; Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Kamide, Hideki
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 6 Pages, 2008/11
Gas entrainment (GE) at free surface is one of the significant issues for the design of Japanese Sodium Cooled Fast Reactor. In the previous study, authors confirmed that GE did not occur at the rated operating mode in the reactor. In the present study, a water experiment to simulate the startup operation of reactor was performed in a large-scaled partial model of the reactor upper plenum in the reactor. The onset condition of GE was observed by the visualization of free surface. Vertical velocity distribution was also measured in order to qualify the mechanism of GE. As a result, it was confirmed that GE did not occur in the startup condition of reactor.
Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Kamide, Hideki
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 7 Pages, 2008/11
An innovative sodium cooled fast reactor has been investigated as part of the fast reactor cycle technology development (FaCT) project. Thermal stratification after a scram is one of main thermal loads of the reactor vessel (R/V). An upper inner structure (UIS) has a slit in radial direction for fuel handling. A water experiment using an 1/10 scale model was carried out. Steep temperature gradient across the stratification interface was observed at the neighborhood of the UIS slit in the experiment. In order to mitigate the temperature gradient across the stratification interface, the height of a plug, which was installed in the upper plenum to infill a hole at a dipped plate for a fuel handling machine, was changed as the parameter in the experiment. The experimental result shows that the temperature gradient near the R/V wall in the case of the lower plug position was 13% smaller than that in the case of the higher plug position.
Ito, Kei; Eguchi, Yuzuru*; Monji, Hideaki*; Ohshima, Hiroyuki; Uchibori, Akihiro; Xu, Y.*
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 8 Pages, 2008/11
A gas entrainment (GE) evaluation method presented at the previous symposium can predict a gas core length by applying local instant values (obtained from CFD results) to the extension vortex theory. However, in the GE evaluation method, a surface tension effect was not introduced. Therefore, it is valid to consider that gas core lengths were overestimated. In this study, the prediction accuracy of gas core lengths is improved by introducing the surface tension effects into the GE evaluation method. For that purpose, the mechanical balance between gravitational, centrifugal and surface tension forces are considered. The improved method was validated by predicting the gas core lengths in basic experiments. As the results, the predicted gas core length values by the improved evaluation method gave better agreements with the experimental results than the original evaluation method.
Tanaka, Masaaki; Ohshima, Hiroyuki; Monji, Hideaki*
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 8 Pages, 2008/11
Numerical simulation of thermal striping phenomena in a T-junction piping system is conducted using a numerical simulation code "MUGTHES". Standard Smagorinsky model is employed as eddy viscosity model with the model coefficient of 0.14 (=Cs). Numerical results are verified by the comparisons with experimental results of velocity and temperature. Applicability of MUGTHES to the thermal striping phenomena is confirmed and the characteristic large-scale eddy structure which dominates thermal mixing and may cause high-cycle thermal fatigue is revealed.
Uchibori, Akihiro; Watanabe, Akira*; Ohshima, Hiroyuki
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 8 Pages, 2008/11
When the pressurized water leaks from a failed heat transfer tube in a steam generator of sodium cooled fast reactors, a highly underexpanded jet with sodium-water chemical reaction may cause wastage of the adjacent tubes. A computer program SERAPHIM for compressible multi-phase flows with sodium-water chemical reaction has been developed for safety assessment of the steam generator. In this study, numerical analysis of highly underexpanded jets was performed by using the SERAPHIM program to investigate its applicability. Through the analysis of the air jet into the air, it was demonstrated that the flow pattern and the location of a shock wave (Mach disk) were reproduced by the use of the high-resolution TVD scheme. Accuracy of the numerical results by the TVD scheme was higher than that by the first-order upwind scheme. Behaviors of the air jet into the water were also reproduced well in the present analysis.
Liu, W.; Tamai, Hidesada; Yoshida, Hiroyuki; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 5 Pages, 2008/11
To discuss the feasibility of Steam Generator (SG) with a straight double-walled heat transfer tube that used in the Fast Breeder Reactor (FBR) system, we need to construct thermal hydraulic design method that can predict the flow instability accurately. To verify and to improve the correlations that used in the thermal-hydraulic design of the SG, Japan Atomic Energy Agency has started experiments under high pressure conditions. Detailed thermal hydraulic data including pressure drop data have been derived. This research does the analysis to the performed experiments with using TRAC-BF1 code. The pressure drop under high pressure condition is verified. It is found that with using the drift flux model in Track code for the void fraction calculation, Pffan's correlation for the friction pressure drop calculation in single phase flow and Martinelli-Nelson two-phase multiplier, the pressure drop can be predicted conservatively.
Ochi, Daisuke*; Someya, Satoshi*; Ohshima, Hiroyuki; Okamoto, Koji*
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 6 Pages, 2008/11
It is important to advance safety and efficiency of the FBR. A numerical prediction method of high accuracy and its validation with a simulated thermal flow field are indispensable. A wire-wrapped rod bundle system, which briefly simulated the fuel rods system in FBR, was built up in the experiments. The wire-wrapped rods were made from Mexflon-material, of which refractive index was exactly same with that of water. The particle image velocimetry was applied to measure the velocity field in the narrow gap between wire-wrapped rods, under variable flow rates with or without heating rods. The aim of this study was to contribute to the certification of results of numerical simulation for the safety design of the FBR.
Someya, Satoshi*; Ochi, Daisuke*; Ohshima, Hiroyuki; Okamoto, Koji*
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 6 Pages, 2008/11
Temperature sensitive particles incorporating phosphor molecules were synthesized. These particles, suitable for particle image velocimetry, were used to measure the velocity and temperature distributions in water flowing through a wire-wrapped rod bundle system, which simulated the fuel rod system in a fast breeder reactor. The particles were illuminated by a pulse laser at 20 Hz. A high speed camera was used to record 30 particle images at intervals of 25
50 micro seconds (2040 kHz) for each excitation laser pulse. From each series of images the velocity and temperature fields were calculated. This measurement technique should contribute to the experimental validation of numerical simulations for the safe design of fast breeder reactors.
Matsumoto, Toshinori*; Takata, Takashi*; Yamaguchi, Akira*; Kurihara, Akikazu; Ohshima, Hiroyuki
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 6 Pages, 2008/11
In a steam generator of sodium-cooled fast reactor, high temperature reacting jet is generated when a heat transfer tube fails and it might cause a secondary fauilure of neighboring tubes due to tube deterioration. Quantification of heat transfer from fluid to the tube is important perspective of safety evaluation. In this study, the heat transfer coefficient on the heat transfer tube under sodium-water reaction phenomena was numerically estimated based on the temperature measured in a sodium experiment using SWAT-1R test facility of JAEA. Furthermore, the floa characteristics on the heat transfer tube was investigated taking into account the variation of the heat transfer coefficient.
Uchida, Mitsunori*; Someya, Satoshi*; Okamoto, Koji*; Ohshima, Hiroyuki
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 6 Pages, 2008/11
When a heat exchanger in a fast breeder reactor cracks, a sodium-water reaction occurs. When a tube cracks, highly pressurized water or steam escapes into the surrounding liquid sodium. The release of steam into the liquid sodium media is a two-phase flow with an underexpansion. There have been few reports on the underexpansion of the gas-liquid phase. In this paper, qualitative measurement of the two-phase flow was carried out for the purpose of revealing the flow with the underexpanded gas jet injected into water. The gas jet range and the gas jet width were then obtained from averaged images of a high-speed camera. PIV was also carried out by observing scattering light from the gas bubbles. The gas jet range and the gas jet width increased approximately linearly with increasing pressure. The results of PIV showed that the bubble velocity increased increasing pressure.
Zhang, B.*; Harada, Tetsushi*; Hirahara, Daisuke*; Matsumoto, Tatsuya*; Morita, Koji*; Fukuda, Kenji*; Yamano, Hidemasa; Suzuki, Toru; Tobita, Yoshiharu
Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 8 Pages, 2008/11
In the present study, characteristics of self-leveling behaviors were investigated experimentally with simulant materials. The present experiments employed depressurization boiling of water to simulate axially increasing void distribution in a debris bed, which consists of solid particles of alumina or lead with different density. The particle size (from 0.5 mm to 6 mm in diameter) and shape (spherical or non-spherical particles) were also taken as experimental parameters. A rough criteria for self-leveling occurrence is proposed and compared with the experimental results.