ANALYSIS OF STEAM GENERATOR TUBE PLUGGING IN A PWR. INFLUENCE IN THE EMERGENCY OPERATING PROCEDURES

A number of Nuclear Power Plants (NPPs) with Pressurized Water Reactors (PWR) in the world have replaced their steam generators (SG) due to degradation of the SG tubes caused by different problems (stress corrosion cracking, fretting, wear, pitting, denting, material problems, etc). Several methods were attempted to correct the defects of the tubes, but eventually the most permanent solution was found to be their plugging. The consequences of plugging the tubes is the subsequent loss of heat transfer surface, reduction of the primary system mass flow, and, for plugging percentage higher than a given value, the consequential reduction of reactor nominal power and economic losses. The objective of this paper is to present an analysis with Relap5/mod3.3 patch03 for the Spanish reactor ASCÓ-2, a 3-loops 2940.6 MWth Westinghouse PWR, in which steam generator tubes are simulated to be plugged in order to find the limit for the adequate operation of the plant. Several steady state calculations were performed with different fractions of plugged SG tubes, by modelling the reduction of the primary to secondary heat transfer surface and the reduction of the primary coolant mass flow area in the tubes as well. The results of the analysis yield that plugging 12% of the SG tubes is around the limit for optimal reactor operation. This limit considers also the proper operation of the turbine valves avoiding their complete opening. To complete the study two events in which the steam generators are used to cool-down the plant were simulated to find out if the plugging of SGs tubes could influence the efficiency of the operator actions described in the Emergency Operating Procedures (EOP) to handle this kind of events. The selected events have been the rupture of one tube of a SG (SGTR) and a small break LOCA (SBLOCA). Two cases with no plugging at all (0%) and 12% SG tube plugging were performed. The actions of the corresponding EOPs for SG tube rupture and SBLOCA were coded in the calculations. The results show no significant impact on operator actions.JRC.F.5-Nuclear Reactor Safety Assessmen