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Kimura, Nobuyuki; Kamide, Hideki; Emonot, P.*; Nagasawa, Kazuyoshi*
Proceedings of 7th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-7) (CD-ROM), 17 Pages, 2008/10
A quantitative evaluation on thermal striping, in which temperature fluctuation due to convective mixing between hot and cold fluids causes thermal fatigue in structural components, is of importance for reactor safety. It is necessary for the quantitative evaluation to investigate occurrence and propagate processes of temperature fluctuation, e.g., decay of temperature fluctuation near structures and transfer of temperature fluctuation from fluid to structures. In Japan Atomic Energy Agency, an innovative sodium cooled fast reactor has been designed. The transfer characteristics of temperature fluctuation from fluid to structure in sodium are quite different from that in water. In order to realize the sodium cooled fast reactor, the clarification of the transfer characteristics of temperature fluctuation is of importance for the rational design against the thermal striping phenomena. In this study, sodium and water experiments of parallel triple jets configuration were performed. For these experiments, numerical simulations were carried out to evaluate the transfer characteristics of temperature fluctuation from fluid to structure. The analysis code, called Trio_U, used in the study has been developed at the CEA in France. The large eddy simulation model is incorporated in the code. Furthermore, the code can calculate fluid and structural domains simultaneously. In the simulations, the calculated time-averaged temperature distributions in fluid and structure were close to the experimental results in sodium and water. The power spectrum densities of temperature fluctuation in fluid and structure were also in good agreements between the experiments and calculations. Furthermore the calculated decay characteristics of temperature fluctuation from fluid to structure were in good agreements with the experimental results.
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.
Nishimura, Masahiro; Kamide, Hideki; Sugiyama, Kenichiro*; Otake, Shiro*
Proceedings of 7th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-7) (CD-ROM), 12 Pages, 2008/10
Liquid sodium is used as the coolant of the fast reactor (FR), because of its high thermal conductivity and wide temperature range of liquid phase. It is superior as thermal medium, however the chemical reactivity with water and oxygen is very high. Hence FR plants have been taking safety measures for these reactions. After the "Monju" sodium leak accident, it is desired that more physical and chemical understanding of reaction phenomena and more mechanistic analysis for the sodium fire from the view point of public acceptance. The purpose of this study is to understand oxidation behavior of a liquid sodium droplet precisely, which is a fundamental reaction of spray fire and is easy to observe the reaction interface. This study is also useful for the establishment of safety criterion to handle the remained non-burning sodium after the accident. The oxidation of a liquid sodium droplet was visualized by using a simple experimental setup and a high speed video camera. A sodium single droplet of ca. 50 mg was made at the tip of a nozzle in a combustion chamber. The oxidation was started by supply of oxygen and nitrogen mixture gas. The initial temperature of sodium droplet and the oxygen fraction in the atmosphere were selected as experimental parameters. It was shown that columnar oxides grew longer as initial temperature of sodium droplet was lower and oxygen fraction was lower. In addition, it was observed that sodium combustion with an orange light emission started from the tip of columnar oxides grown out from the droplet surface. These observations suggest the existence of mechanism that liquid sodium is drawn up from droplet to reaction interface by the capillary force caused in the porous oxides which are formed on the droplet surface.
Yamaguchi, Akira*; Takata, Takashi*; Tatsumi, Eisaku*; Ito, Kei; Ohshima, Hiroyuki; Kamide, Hideki; Sakakibara, Jun*
Proceedings of 7th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-7) (CD-ROM), 16 Pages, 2008/10
In a sodium-cooled fast reactor, inert gases exist in the primary coolant system either in a state of dissolved gas or free gas bubbles. The gas in the primary system may cause disturbance in reactivity, nucleation site for boiling, etc. Therefore, the allowance level is necessary regarding the gas entrainment and the gas bubble concentration. In the present study, a gas entrainment allowance level is discussed and rationalized. For the purpose, a plant dynamics code VIBUL has been developed to evaluate the concentration distribution of the dissolved gas and the free gas bubble. Assuming that the total void fraction should be kept below 105% of the background level, a preliminary design allowance level of gas entrainment is proposed as the map in terms of the entrainment ratio and the entrained bubble radius. Furthermore, the possibility of bubble removal is investigated to satisfy the allowance level.
Takeda, Tetsuaki*; Ichimiya, Koichi*; Nishio, Hitoshi*; Nakagawa, Shigeaki; Takamatsu, Kuniyoshi
Proceedings of 7th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-7) (CD-ROM), 11 Pages, 2008/10
Safety demonstration tests using the HTTR (High Temperature Engineering Test Reactor) are being performed to verify the inherent safety features and to validate the numerical code for the safety assessment of the VHTR (Very High Temperature Reactor). The partial loss of coolant flow test was performed under the condition of the ATWS (Anticipated Transient without Scram). We are planning to perform the test of loss of coolant flow and stopping the vessel cooling system (VCS). The test of the loss of coolant flow as one of safety demonstration tests is carried out by tripping all gas circulators, and the stopping VCS test is performed continuously after the loss of coolant flow. The objective of this study is to evaluate the temperature distribution of the reactor pressure vessel (RPV) and the VCS during the tests. It is necessary to consider the effect of thermal radiation from the RPV for evaluation of temperature of the VCS and concrete vessel.
Yuki, Kazuhisa*; Ohara, Hiroshi*; Hashizume, Hidetoshi*; Tanaka, Masaaki; Muramatsu, Toshiharu; Toda, Saburo*
Proceedings of 7th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-7) (CD-ROM), 21 Pages, 2008/10
Velocity measurement by PIV (Particle Image Velocimetry) and measurement of fluid-temperature are conducted in a mixing tee with an elbow on the upstream side of the tee. Parameters in the experiment are the distance between the elbow pipe exit and the tee and the curvature radius of the elbow pipe. By the experiment, certain suppression techniques for high cycle thermal fatigue in the mixing tee area are proposed to control the distance between the elbow and the tee according to the curvature radius of the elbow.