Stress Corrosion Cracking of Low-Alloy Reactor Pressure Vessel Steels and of a Weld Filler Material under BWR Conditions

摘要

The stress corrosion cracking (SCC) crack growth behaviour of three different nuclear grade reactor pressure vessel (RPV) steels (SA 533 B Cl.1, SA 508 Cl.2, 20 MnMoNi 5 5) and of a RPV weld filler material under simulated boiling water reactor (BWR)/normal water chemistry (NWC) conditions was characterised by constant and ripple load tests with pre-cracked fracture mechanics specimens in oxygenated high-temperature water at temperatures of either 288, 250, 200 or 150 °C. Modern high-temperature water loops, on-line crack growth monitoring (DCPD) and fractographical analysis by SEM were used to quantify the cracking response. In agreement with former investigations and literature data the conservative character of the "BWR VIP 60 SCC disposition lines" for SCC crack growth in low-alloy steels (LAS) has been further confirmed by this study for 288 °C and RPV base and weld filler material. Preliminary results indicated, that these disposition lines may be significantly or slightly exceeded in the case of small load fluctuations at high load ratios (ripple loading) or at intermediate temperatures (200 - 250 °C) in RPV materials, which show a distinct susceptibility to dynamic strain ageing (DSA). It was concluded, that the concentration of "free" interstitial nitrogen/carbon might therefore be just as relevant for SCC susceptibility as the steel sulphur content or the morphology, size and distribution of the MnS-inclusions, at least under conditions where DSA is typically observed.