This paper presents a multi-dimensional numerical analysis of the transient thermal-hydraulic response of a steam generator secondary side to a double-ended guillotine break of the main steam line attached to the steam generator at a pressurized water reactor plant. A simplified analysis model is designed to include both the steam generator upper space where steam occupies and a part of the main steam line between the steam generator outlet nozzle and the pipe break location upstream of the main steam isolation valve. The transient steam flow through the analysis model is simulated using the shear stress transport turbulence model. The steam is treated as a real gas. To model the steam generation by heat transfer from the primary coolant to the secondary side coolant for a short period during the blow down process following the main steam line break accident, a constant amount of steam is assumed to be generated from the bottom of the steam generator upper space part. Using the numerical approach mentioned above, calculations have been performed for the analysis model having the same physical dimensions of the main steam line pipe and initial operational conditions as those for an actual operating plant. The calculation results have been discussed in detail to investigate their physical meanings and validity. The results demonstrate that the present CFD model is applicable for simulating the transient thermal-hydraulic responses in the event of the MSLB accident including the blowdown-induced dynamic pressure disturbance in the SG. In addition, it has been found that the dynamic hydraulic loads acting on the SG tubes can be increased by 2 to 8 times those loads during the normal reactor operation. This implies the need to re-assess the potential for single or multiple SG tube ruptures due to fluidelastic instability for ensuring the reactor safety.
Transient Thermal-Hydraulic Responses of the Nuclear Steam Generator Secondary Side to a Main Steam Line Break
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Jo, JC, & Moody, FJ. "Transient Thermal-Hydraulic Responses of the Nuclear Steam Generator Secondary Side to a Main Steam Line Break." Proceedings of the ASME 2014 Pressure Vessels and Piping Conference. Volume 4: Fluid-Structure Interaction. Anaheim, California, USA. July 20–24, 2014. V004T04A095. ASME. https://doi.org/10.1115/PVP2014-28134
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