Nuclear Technology / Volume 152 / Number 3 / December 2005 / Pages 302-313
Technical Paper / Thermal Hydraulics / dx.doi.org/10.13182/NT05-5
A pressurized water reactor steam generator tube rupture (SGTR) is of concern because it represents a bypass of the containment for radioactive materials to the environment. In a station blackout accident, tube integrity could be threatened by creep rupture, particularly if cracks are present in the tube walls. Methods are developed herein to improve assessment capabilities for SGTR by using the severe-accident code MELCOR. Best-estimate assumptions based on recent research and computational fluid dynamics calculations are applied in the MELCOR analysis to simulate two-dimensional natural circulation and to determine the relative creep-rupture timing in the reactor coolant pressure boundary components. A new method is developed to estimate the steam generator (SG) hottest tube wall temperature and the tube critical crack size for the SG tubes to fail first. The critical crack size for SG tubes to fail first is estimated to be 20% of the wall thickness larger than by a previous analysis. Sensitivity studies show that the failure sequence would change if some assumptions are modified. In particular, the uncertainty in the countercurrent flow limit model could reverse the failure sequence of the SG tubes and surge line.