Modeling and Loss-of-Coolant Accident Analysis of a Nuclear Power Plant Using RELAP5/MOD2

A best-estimate small-break loss-of-coolant accident (SBLOCA) analysis of a four-loop pressurized water reactor nuclear power plant is conducted using the computer code RELAP5/MOD2. A plant-specific RELAP5 model is developed, and steady-state operating conditions are calculated. The steady-state model is then employed to obtain SBLOCA scenarios for different break sizes. Transients resulting from the different breaks are studied to determine the limiting break size and obtain comprehensive transient scenarios. The effect of the hydraulic diameter on the transient behavior, related to the steam generator U-tubes, is also observed. The relationship between the break size and the peak cladding temperature is obtained. The study indicates that as the break size increases, a smaller core inventory instigates heatup during core boil-off.