抄録
The present study was intended to grasp flow-induced vibration characteristics in the hot-leg piping of Japan sodium-cooled fast reactor by newly taken experimental data as well as to verify a vibration analysis tool with its data. Considering reactor upper sodium plenum flow in the reactor condition, flow-induced-vibration tests were carried out to investigate the effect of swirl flow at the inlet of the hot-leg piping in this study. The parameters were a swirl flow velocity ratio, which was defined as the swirl flow velocity on the inner surface of the pipe divided by the mean velocity. The parameter range of the swirl flow velocity ratio was set 5%〜15% for the conservative evaluation. In these tests, the random force distributions along the pipe and their correlation lengths were measured with pressure sensors to evaluate the flow-induced vibrations. It was found that the influence of the pressure fluctuation due to the swirl flow was negligibly small. The power spectrum densities of pressure fluctuations and correlation lengths were classified into some sections in order to reasonably evaluate flow-induced-random vibration response for reactor power plant piping. The vibration analysis method was proposed based on the measured power spectrum densities and correlation lengths of turbulent-flow induced forces. 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 is valid.
