Survey of thermal-hydraulic correlations for gas, lead and lead-bismuth coolants

not registered; Ohshima, Hiroyuki

Feasibility study is being carried out at JNC to generate new concepts of practical fast breeder reactors. ln this study, appropriate thermal-hydraulic correlations for several kinds of coolants are required to assess thermal-hydraulics of proposed core/fuel-assembly designs, which have different characteristics from traditional liquid-sodium cooled fast reactors, e.g., ribbed fuel pins and fuel pin square arrangement with spacer. ln the present report thermal-hydraulic correlations for carbon di-oxide, helium, lead, and lead-bismuth cooled reactors were surveyed. Several candidates for pressure drop coefficient and heat transfer coefficient of each coolant were picked from available papers and literatures and were examined by using the design specifications of ETGBR (carbon di-oxide cooled reactor), GBR4(helium cooled reactor) and BREST300 (lead, lead-bismuth cooled reactor) as well as existing experimental data. Finally thermal-hydraulic correlations of each coolant, which are applicable to the regions from laminar to turbulent flow, were proposed.