The steam generator (SG) in a pressurized water reactor (PWR) is an important component as boundary between the primary loop and the secondary loop. In this study, we performed two-phase flow analysis of SG reliability tests conducted by the Nuclear Power Engineering Corporation (NUPEC) using the two-fluid models of a thermal-hydraulic computer code PHOENICS. It was difficult to calculate the location of the boiling initiation accurately because the location was greatly affected by the friction coefficients (i.e. velocity distributions) and the heat transfer distributions. However, the friction coefficients and the heat transfer distributions did not greatly affect the void fractions in the upper region of the U-bent tubes and the calculated average void fractions agreed with the measured within 4%.

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