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

• Published in: Nuclear engineering and design 2011-11, Vol.241 (11), p.4464-4475
• Main Authors: Yamano, Hidemasa, Tanaka, Masa-aki, Kimura, Nobuyuki, Ohshima, Hiroyuki, Kamide, Hideki, Watanabe, Osamu
• Format: Article
• Language: English
• Subjects: Fluid flow; Methodology; Navier-Stokes equations; Nuclear engineering; Nuclear power generation; Nuclear reactor components; Nuclear reactors; Piping
• ISSN: 0029-5493; 1872-759X
• DOI: 10.1016/j.nucengdes.2011.06.005

This paper describes the current status of flow-induced vibration evaluation methodology development for the primary piping in Japan sodium-cooled fast reactor, with particularly emphasis on the development approach and research activities that investigate unsteady hydraulic characteristics in a short-elbow piping. The approach to the methodology development was defined: experiment-based methodology and simulation-based one as well as extrapolation logic to the reactor condition based on no dependency on Reynolds number in the high Reynolds number range from the experimental results. Experimental efforts have been made using 1/3-scale single-elbow test sections for the hot-leg piping as the main activity. Recent experiments using the 1/3-scale test section revealed that a swirl flow at the inlet of the hot-leg piping hardly influenced pressure fluctuations onto the pipe though a slight deformation of flow separation was observed. Numerical results under different Reynolds number conditions appear in this paper using the unsteady Reynolds Averaged Navier Stokes equation approach, indicating its applicability to the hot-leg piping experiments.