Numerical Analysis of Flow-Induced Vibration of Large Diameter Pipe With Short Elbow

2017 ◽  
Author(s):  
Shigeru Takaya ◽  
Tatsuya Fujisaki ◽  
Masaaki Tanaka
Keyword(s):  
Numerical Simulation ◽  
Numerical Analysis ◽  
Fast Reactor ◽  
Pressure Fluctuation ◽  
Cooling System ◽  
Large Diameter ◽  
Recirculation Region ◽  
Flow Induced Vibration ◽  
Scaled Model ◽  
Energy Agency

Japan Atomic Energy Agency is now conducting design study and R&D of an advanced loop-type sodium cooled fast reactor. The cooling system is planned to be simplified by employing a two-loop configuration and shortened piping with less elbows than a prototype fast reactor in Japan, Monju, in order to reduce construction costs and enhance economic performance. The design, however, increases flow velocity in the hot-leg piping and induces large flow turbulence around elbows. Therefore, flow-induced vibration (FIV) of a hot-leg piping is one of main concerns in the design. Numerical simulation is a useful method to deal with such a complex phenomenon. We have been developing numerical analysis models of the hot-leg piping using Unsteady Reynolds Averaged Navier-Stokes simulation with Reynolds stress model. In this study, numerical simulation of a 1/3 scaled-model of the hot-leg piping was conducted. The results such as velocity profiles and power spectral densities (PSD) of pressure fluctuations were compared with experiment ones. The simulated PSD of pressure fluctuation at the recirculation region agreed well with the experiment, but it was found some underestimation at other parts, especially in relatively high frequency range. Eigenvalue vibration analysis was also conducted using a finite element method. Then, stress induced by FIV was evaluated using pressure fluctuation data calculated by URANS simulation. The calculated stress generally agrees well the measurement values, which indicates the importance of precise evaluation of the PSD of pressure fluctuation at the recirculation region for evaluation of FIV of the hot-leg piping with a short elbow.

Numerical Analysis of Unsteady Phenomena in Upper Plenum and Hot-Leg Piping System of Advanced Loop-Type Sodium Cooled Fast Reactor

2016 ◽  
Author(s):  
Shigeru Takaya ◽  
Tatsuya Fujisaki ◽  
Masaaki Tanaka
Keyword(s):  
Flow Field ◽  
Fast Reactor ◽  
Cooling System ◽  
Coupling Effect ◽  
Integrated Model ◽  
Piping System ◽  
Scaled Model ◽  
Energy Agency ◽  
Simulation Results ◽  
Inlet Conditions

Japan Atomic Energy Agency is now conducting design study and R&D of an advanced loop-type sodium cooled fast reactor. The cooling system is planned to be simplified by employing a two-loop configuration and shortened piping with less elbows than a prototype fast reactor in Japan, Monju, in order to reduce construction costs and enhance economic performance. The design, however, increases flow velocity in the hot-leg piping and induces large flow turbulence around elbows. Therefore, flow-induced vibration of a hot-leg piping is one of main concerns in the design. The flow field in the hot-leg piping is affected by flow disturbance at the inlet, so it is important to evaluate flow field including the upper plenum. In this study, we analyzed unsteady fluid flow by using an integrated model of the upper plenum and the hot-leg piping system. Unsteady Reynolds averaged Navier-Stokes simulation with Reynolds stress model was used for the numerical simulation. The results were compared with experiment results of 1/3 scaled-model of hot-leg piping with the inlet conditions of rectified, swirling and deflected flows as well as simulation results of 1/3 scaled-model of hot-leg piping with rectified flow. In general, the simulation results obtained by using the integrated model show a similar tendency with the experiment results of deflect flows in the downstream region from the elbow exit. The coupling effect of swirling and deflected flows seems to be not significant although further investigation is needed.

Development of System Based Code: Part 1—Reliability Target Derivation of Structures and Components

2010 ◽  
Author(s):  
Kenichi Kurisaka ◽  
Ryodai Nakai ◽  
Tai Asayama ◽  
Shigeru Takaya
Keyword(s):  
Fast Reactor ◽  
Present Method ◽  
Structural Reliability ◽  
Cooling System ◽  
New Method ◽  
Analysis Model ◽  
Energy Agency ◽  
Target Value ◽  
Static Failure

The present paper describes a new method for determining the target value of structural reliability in the framework of the System Based Code (SBC) by considering the safety point of view. In the new method, the reliability target is derived from the proposal to a quantitative safety goal that was published by the nuclear safety commission (NSC) of Japan and the quantitative safety design requirements on the core damage frequency (CDF) and the containment failure frequency (CFF) that were determined in the Fast Reactor Cycle Technology Development (FaCT) project by Japan Atomic Energy Agency (JAEA), by utilizing analysis models of a probabilistic safety assessment (PSA). The present method was applied to determination of the reliability target of the structures and components which constitute the reactor cooling system in the Japanese sodium-cooled fast reactor (JSFR). The risk from the reactor is expressed with sum of combination of various elements in the PSA analysis model. Those elements include not only static failure of the structures and components. However, the present study focuses on the sequences including the static failure, and the probability of dynamic failures and human errors in those sequences is conservatively assumed as a unity. It was confirmed that the present method combined with the PSA analysis model for internal initiating events is applicable to determination of the reliability target associated with a random failure of the structures and components, and that the method related to seismic initiating events can derive the target value of the occurrence frequency at which any of the important structures and components fails due to an earthquake.

Pressure Fluctuation Characteristics of Complex Turbulent Flow in a Single Elbow With Small Curvature Radius for a Sodium-Cooled Fast Reactor

2010 ◽  
Vol 132 (11) ◽  
Author(s):  
Shinji Ebara ◽  
Yuta Aoya ◽  
Tsukasa Sato ◽  
Hidetoshi Hashizume ◽  
Yuki Kazuhisa ◽  
...  
Keyword(s):  
Reynolds Number ◽  
Turbulent Flow ◽  
Fast Reactor ◽  
Pressure Fluctuation ◽  
Scale Model ◽  
Curvature Radius ◽  
Piping System ◽  
Flow Induced Vibration ◽  
Good Correspondence ◽  
High Reynolds

A multi-elbow piping system is adopted for the Japan sodium-cooled fast reactor (JSFR) cold-legs. Flow-induced vibration (FIV) is considered to appear due to complex turbulent flow with very high Reynolds number in the piping. In this study, pressure measurement for a single elbow flow is conducted to elucidate pressure fluctuation characteristics originated from turbulent motion in the elbow, which lead potentially to the FIV. Two different scale models, 1/7- and 1/14-scale simulating the JSFR cold-leg piping, are tested experimentally to confirm whether a scale effect in pressure fluctuation characteristics exists. A distinguishing peak can be seen in each power spectrum density (PSD) profile of pressure fluctuation obtained in and downstream of the flow separation region for both scaled models. When nondimensionalized, the PSD profiles show good correspondence regardless of scale model and even of Reynolds number simulated in this study.

Flow-induced Vibration of Large-diameter Elbow : (1) Flow Pattern and Pressure Fluctuation

2004 ◽  
Vol 2004.7 (0) ◽  
pp. 103-104
Author(s):  
Tadashi SHIRAISHI ◽  
Hisato WATAKABE ◽  
Hiromi SAGO ◽  
Tomomichi NAKAMURA ◽  
Yoshihide ISHITANI ◽  
...  
Keyword(s):  
Flow Pattern ◽  
Pressure Fluctuation ◽  
Large Diameter ◽  
Flow Induced Vibration

Pressure Measurement Test of Single Elbow Simulating Na Cooled Fast Reactor Cold-Leg Piping

2010 ◽  
Author(s):  
Shinji Ebara ◽  
Yuta Aoya ◽  
Tsukasa Sato ◽  
Hidetoshi Hashizume ◽  
Kazuhisa Yuki ◽  
...  
Keyword(s):  
Power Spectrum ◽  
Fast Reactor ◽  
Pressure Fluctuation ◽  
Pressure Measurement ◽  
Power Spectrum Density ◽  
Piping System ◽  
Complex Flow ◽  
Flow Induced Vibration ◽  
Spectrum Density ◽  
Very High

Regarding the Japan Sodium-cooled Fast Reactor, a multi-elbow piping system is adopted for its cold-legs. Flow Induced Vibration (FIV) is considered to be caused by complex flow with very high velocity in the elbows. In this study, pressure measurement test of a single elbow flow is conducted to find out pressure fluctuation characteristic which is related to the elbow turbulent flow and lead potentially to the FIV. Two types of experimental loops, that is, 1/7 and 1/15-scale setup simulating the JSFR cold-leg pipings, are used for pressure measurement, and a distinguishing peak can be seen in the power spectrum density profile of pressure fluctuation obtained where flow separation occurs and at the downstream from it. This characteristics of pressure fluctuation is obtained from the two different scale experiments, and the scale effect is not found in terms of the pressure fluctuation.

Development of flow-induced vibration evaluation methodology for large-diameter piping with elbow in Japan sodium-cooled fast reactor

2011 ◽  
Vol 241 (11) ◽  
pp. 4464-4475 ◽  
Author(s):  
Hidemasa Yamano ◽  
Masa-aki Tanaka ◽  
Nobuyuki Kimura ◽  
Hiroyuki Ohshima ◽  
Hideki Kamide ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Large Diameter ◽  
Evaluation Methodology ◽  
Flow Induced Vibration
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