In support of developing SuperCritical Water-cooled Reactors (SCWRs), studies are currently being conducted for heat-transfer at supercritical conditions. This paper presents an analysis of heat-transfer to SuperCritical Water (SCW) flowing in bare vertical tubes as a first step towards thermohydraulic calculations in a fuel-channel. A large set of experimental data, obtained in Russia, was analyzed. Two updated heat-transfer correlations for forced convective heat transfer in the normal heat transfer regime to SCW flowing in a bare vertical tube were developed. It is expected that the next generation of water-cooled nuclear reactors will operate at supercritical pressures (∼25 MPa) with high coolant temperatures (350–625°C). Currently, there are no experimental datasets for heat transfer from power reactor fuel bundles to the fuel coolant (water) available in open literature. Therefore, for preliminary calculations, heat-transfer correlations obtained with bare tube data can be used as a conservative approach. The analyzed experimental dataset was obtained for SCW flowing upward in a 4-m-long vertical bare tube. The data was collected at pressures of about 24 MPa for several combinations of wall and bulk-fluid temperatures that were below, at, or above the pseudocritical temperature. The values for mass flux ranged from 200–1500 kg/m2s, for heat flux up to 1250 kW/m2 and inlet temperatures from 320–350°C. The Mokry et al. correlation was developed as a Dittus-Boelter-type correlation, with thermophysical properties taken at bulk-fluid temperatures. Alternatively, the Gupta et al. correlation was developed based on the Swenson et al. approach, where the majority of thermophysical properties are taken at the wall temperature. An analysis of the two updated heat-transfer correlations is presented in this paper. Both correlations demonstrated a good fit (±25% for Heat Transfer Coefficient (HTC) values and ±15% for calculated wall temperatures) for the analyzed dataset. Thus, these correlations can be used for preliminary HTC calculations in SCWR fuel bundles as a conservative approach, for SCW heat exchangers, for future comparisons with other independent datasets and for the verification of computer codes for SCWR core thermohydraulics.

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