Unsteady Elbow Pipe Flow to Develop a Flow-Induced Vibration Evaluation Methodology for Japan Sodium-Cooled Fast Reactor

スポンサーリンク

概要

• 論文の詳細を見る

• This paper describes the current status of flow-induced vibration evaluation methodology development for primary cooling pipes in the Japan sodium-cooled fast reactor (JSFR), with particular emphasis on recent research and development activities that investigate unsteady elbow pipe flow. Experimental efforts have been made using 1/3-scale and 1/10-scale single-elbow test sections for the hot-leg pipe. The 1/10-scale experiment simulating the hot-leg pipe indicated no effect of pipe scale in comparison with the 1/3-scale experiment under inlet-rectified-flow conditions. The next experiment using the 1/3-scale test section was performed to investigate the effect of swirl flow at the inlet. Although the flow separation region was deflected at the downstream from the elbow, the experiment clarified a less significant effect of swirl flow on pressure fluctuation onto the pipe wall. An additional experiment was intended to study the effect of elbow curvature. The experiments with water revealed no clear flow separation in a larger curvature elbow case than that of the JSFR. Since the interference of multiple elbows should be investigated to understand turbulent flow in the cold-leg pipe geometry, 1/15-scale experiments with double elbows were carried out to clarify that flow in the first elbow influenced a flow separation behavior in the second elbow. Simulation activities include Unsteady Reynolds Averaged Navier Stokes equation (U-RANS) approach with a Reynolds stress model using a commercial computational fluid dynamics (CFD) code and Large Eddy Simulation (LES) approach using an in-house code. A hybrid approach that combined with RANS and LES was also applied using a CFD code. Several numerical results appear in this paper, focusing on its applicability to the hot-leg pipe experiments. These simulations reasonably agreed with the experimental data using the 1/3-scale test section.

• 2011-04-01

著者

• TANAKA Masaaki

  Japan Atomic Energy Agency

• YAMANO Hidemasa

  Japan Atomic Energy Agency

• MURAKAMI Takahiro

  Central Research Institute of Electric Power Industry

• IWAMOTO Yukiharu

  Ehime University

• YUKI Kazuhisa

  Tohoku University

• SAGO Hiromi

  Mitsubishi Heavy Industries, Ltd.

• Hayakawa Satoshi

  Mitsubishi Fbr Systems Inc.

• Sago Hiromi

  Mitsubishi Heavy Industries Ltd.

関連論文

• Thermal Mixing in T-Junction Piping System Related to High-Cycle Thermal Fatigue in Structure
• Thermophysical Properties of Lead-Bismuth Eutectic Alloy in Reactor Safety Analyses
• ICONE11-36299 STUDY ON FLUID TEMPERATURE FLUCTUATION AND TRANSFER TO WALL IN A MIXING TEE
• Unsteady Elbow Pipe Flow to Develop a Flow-Induced Vibration Evaluation Methodology for Japan Sodium-Cooled Fast Reactor
• ICONE11-36362 STUDY ON THERMAL-HYDRAULICS IN A POROUS BLOCKAGE IN A FUEL SUBASSEMBLY OF LMFBR
• Unsteady Elbow Pipe Flow to Develop a Flow-Induced Vibration Evaluation Methodology for Japan Sodium-Cooled Fast Reactor
• Thermal-Hydraulic Analysis for Inversely Stratified Molten Corium in Lower Vessel
• Effect of Swirl Inflow on Flow Pattern and Pressure Fluctuation onto a Single-Elbow Pipe in Japan Sodium-Cooled Fast Reactor

スポンサーリンク