Abstract: An experiment was performed for the OECD/NEA ROSA-2 Project employing the ROSA/LSTF (rig of safety assessment/large-scale test facility), which simulated a steam generator tube rupture (SGTR) accident induced by main steam line break (MSLB) with operator recovery actions in a pressurized water reactor (PWR). The primary pressure decreased to the pressure level nearly-equal to the intact steam generator (SG) secondary-side pressure even with coolant injection from the high-pressure injection (HPI) system of emergency core cooling system (ECCS) into cold legs. Multi-dimensional coolant behavior appeared such as thermal stratification in both hot and cold legs in intact loop. The RELAP5/MOD3.3 code indicated the insufficient predictions of the primary pressure, the SGTR break flow rate, and the HPI flow rate, and failed to predict the fluid temperatures in the intact loop hot and cold legs. Results obtained from the comparison among three LSTF SGTR-related tests clarified that the thermal stratification occurs in the horizontal legs by different mechanisms.
Abstract: The LSTF experiment simulating the SGTR accident at
the Mihama Unit-2 reactor is analyzed using the RELAP5/MOD3.3
code. In the accident, and thus in the experiment, the ECC water was
injected not only into the cold legs but into the upper plenum. Overall
transients during the experiment such as pressures and fluid
temperatures are simulated well by the code. The cold-leg fluid
temperatures are shown to decrease if the upper plenum injection
system is connected to the cold leg. It is found that the cold-leg fluid
temperatures also decrease if the upper-plenum injection is not used
and the cold-leg injection alone is actuated.