Evidence of tribocorrosion wear in pressurized water reactors

The wear of the cobalt-based hardfacing alloy Stellite grade 6, which is used for part of the latch arms of control rod drive mechanisms (CRDM) in pressurized water nuclear reactors, has been studied. The investigations were based on field experience (nondestructive inspection) and on metallurgical examinations of specimens from laboratory studies. A wear mechanism based on tribocorrosion was found to account well for all available data relating to the wear characteristics of latch arms. In particular, such a mechanism is compatible with the specular polished aspect of the worn surface. A simple model based on two main parameters has been developed to assess the remaining life of CRDM latch arms. The first and principal parameter is the number of mechanical interactions between the latch arms and the drive rod (which defines the number of depassivation steps). The second parameter is the mean time between two consecutive steps (which defines the period for material removal during repassivation). The shape of the time distribution of steps does not have a great effect on life prediction within an error band of 25%. This model accounts well for the discrepancies between laboratory loop results and wear measurements in plants. The wear rate is at least four times higher in plants than in test loops. The values of the coefficient comparing the relative experience between field and laboratory are consistent with the knowledge of the electrochemical behavior of alloys that passivate in PWR primary water (C) 2001 Elsevier Science B.V. All rights reserved.