ROSA/LSTF Test on Pressurized Water Reactor Steam Generator Tube Rupture Accident Induced by Main Steam Line Break with Recovery Actions
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ROSA/LSTF Test on Pressurized Water Reactor Steam Generator Tube Rupture Accident Induced by Main Steam Line Break with Recovery Actions

Authors: Takeshi Takeda


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.

Keywords: LSTF, SGTR, thermal stratification, RELAP5.

Digital Object Identifier (DOI):

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