The dual-cooled nuclear reactor is currently considered for improving the designs of current/future nuclear reactors. Investigation of the thermal-hydraulic characteristics of the nuclear reactor via experiments is essential for commercializing the dual-cooled nuclear reactor. In this paper, the turbulent flow in square arrayed six-rod bundles in the form of magnified copies of the dual-cooled and current OPR-1000 nuclear reactor is experimentally investigated by means of hot-wire anemometry and smoke-wire generation methods. Vortex trains which do not exist in an ordinary reactor subchannel are presented in the subchannel of the dual-cooled reactor. The vortices are induced by a span-wise velocity gradient. This flow pulsation phenomenon increases the inter-channel mixing of the subchannel. To understand the periodic feature of the pulsation, axial/cross velocities are measured and the periodic characteristic frequencies are obtained by a Fast Fourier Transform (FFT) analysis. The peak frequency that represents the quasi-periodic pulsation of the flow is increased with an increase in the axial velocity while the wavelength of the pulsation remains constant within a tested range of the Reynolds number (9000-51000). The vortex trains are highly synchronized with each other, as confirmed by means of visualization.
Bibliographical noteFunding Information:
This work has been carried out under the Nuclear R&D Program supported by the Ministry of Education, Science and Technology of the Republic of Korea (Grant No. NRF-2012M2A8A5025824) and the National Research Foundation of Korea (NRF) grant funded by the Korea government (MEST) (Grant No. 2012-0005727)
All Science Journal Classification (ASJC) codes
- Materials Science(all)