Scale model test results for an inverted u-tube steam generator with comparisons to heat transfer correlations

Conference · Thu Jan 01 04:00:00 EST 1987

OSTI ID:5157708

Boucher, T J

To provide data for assessment and development of thermal-hydraulic computer codes, bottom main feedwater-line-break transient simulations were performed in a scale model (Semiscale Mod-2C) of a pressurized water reactor (PWR) with conditions typical of a PWR (15.0 MPa primary pressure, 600 K steam generator inlet plenum fluid temperatures, 6.2 MPa secondary pressure). The state-of-the-art measurements in the scale model (Type III) steam generator allow for the determination of U-tube steam generator secondary component interactions, tube bundle local radial heat transfer, and tube bundle and riser vapor void fractions for steady state and transient operations. To enhance the understanding of the observed phenomena, the component interactions, local heat fluxes, local secondary convective heat transfer coefficients and local vapor void fractions are discussed for steady state, full-power and transient operations. Comparisons between the measurement-derived secondary convective heat transfer coefficients and those predicted by a number of correlations, including the Chen correlation currently used in thermal-hydraulic computer codes, show that none of the correlations adequately predict the data and points out the need for the formulation of a new correlation based on this experimental data. The unique information presented herein should be of interest to anyone involved in modeling inverted U-tube steam generator thermal-hydraulics or forced convection boiling/vaporization heat transfer.

• OSTI ID: 5157708
• Report Number(s): CONF-871234-
• Country of Publication: United States
• Language: English

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