TY - JOUR
T1 - Two-phase pressure drop prediction in helically coiled steam generators for nuclear power applications
AU - Cioncolini, Andrea
AU - Santini, Lorenzo
N1 - Publisher Copyright:
© 2016 Elsevier Ltd. All rights reserved.
PY - 2016
Y1 - 2016
N2 - This study considers the prediction of the pressure gradient with water-steam two-phase flows through helically coiled steam generator tubes, focusing in particular on the operating conditions of low-medium pressure, low mass flux and low heat flux typical of once-through steam generators with in-tube boiling adopted in small modular nuclear reactor systems. Twenty-five widely used empirical correlations have been tested against an experimental pressure drop databank drawn together in this study containing 980 data points. Since no existing correlation is capable of collapsing and satisfactorily fitting the collected databank, a new pressure drop prediction method for helically coiled tubes is proposed. This new prediction method is very simple to implement, as it is based on the homogeneous flow model, is asymptotically consistent with straight tube two-phase flows and is largely superior in accuracy to existing prediction methods (mean absolute error of 7.3%, and 9 points out of 10 captured to within ±15%). The new prediction method is applicable for operating pressures in the range of 0.75-9.0 MPa, mass fluxes from 400 kg/m2s to 1191 kg/m2s, heat fluxes up to 750 kW/m2, tube diameters within 5-20 mm and coil to tube diameter ratio above 32.4. Curvature effects on the pressure gradient in helical coil two-phase flows can be significant, particularly with high velocity flows in tight curvature coils where the centrifugal force is intense.
AB - This study considers the prediction of the pressure gradient with water-steam two-phase flows through helically coiled steam generator tubes, focusing in particular on the operating conditions of low-medium pressure, low mass flux and low heat flux typical of once-through steam generators with in-tube boiling adopted in small modular nuclear reactor systems. Twenty-five widely used empirical correlations have been tested against an experimental pressure drop databank drawn together in this study containing 980 data points. Since no existing correlation is capable of collapsing and satisfactorily fitting the collected databank, a new pressure drop prediction method for helically coiled tubes is proposed. This new prediction method is very simple to implement, as it is based on the homogeneous flow model, is asymptotically consistent with straight tube two-phase flows and is largely superior in accuracy to existing prediction methods (mean absolute error of 7.3%, and 9 points out of 10 captured to within ±15%). The new prediction method is applicable for operating pressures in the range of 0.75-9.0 MPa, mass fluxes from 400 kg/m2s to 1191 kg/m2s, heat fluxes up to 750 kW/m2, tube diameters within 5-20 mm and coil to tube diameter ratio above 32.4. Curvature effects on the pressure gradient in helical coil two-phase flows can be significant, particularly with high velocity flows in tight curvature coils where the centrifugal force is intense.
KW - Curvature effect
KW - Helical coil pressure drop
KW - Small modular nuclear reactor
KW - Steam generator
KW - Two-phase flow
UR - http://www.scopus.com/inward/record.url?scp=84973316066&partnerID=8YFLogxK
U2 - 10.1016/j.ijheatmasstransfer.2016.05.027
DO - 10.1016/j.ijheatmasstransfer.2016.05.027
M3 - 文章
AN - SCOPUS:84973316066
SN - 0017-9310
VL - 100
SP - 825
EP - 834
JO - International Journal of Heat and Mass Transfer
JF - International Journal of Heat and Mass Transfer
ER -