SUMMARY OF THE LOW-ALLOY STEELS SESSION – Chaired by Hans-Peter Seifert

This year the low-alloy steel session on Monday afternoon after the coffee break consisted of only two presentations.

Martin Widera (retired from RWE, Germany) and Armin Roth (Honorary Member, Germany) gave an overview on the background, content and goals of a new vgbe technical-scientific report on strain-induced cracking in fossil-fired power-plants (VGB TW 532). Since decades, strain-induced corrosion cracking (SICC) has repeatedly led to costly damages in fossil fired and nuclear power plants. With the increasingly flexible operation of fossil power plants in Germany, damages caused by SICC have again increased since 2014. In 2020 a vgbe expert group started work on a technical-scientific report describing the state of knowledge on SICC in pressurized systems and components of fossil power plants. The report was published end of 2024 in German. An English translation will be published shortly. A major base for this report is the broad knowledge available on SICC from German nuclear power plants and associated R+D work. So this report also contains beyond fossil fired power plants a valuable source book for those working on SICC assessment for nuclear power plants.

Thereafter, the session chair, Hans-Peter Seifert (PSI, Switzerland), presented work on irradiation hardening and thermal ageing during LWR long-term operation (LTO) of low-alloy reactor pressure vessel (RPV) steels, which could increase their stress corrosion cracking (SCC) susceptibility in high-temperature water (HTW). While only moderate accelerating effects are anticipated/claimed, direct experimental evidence remains missing. Therefore, the SCC and corrosion fatigue (CF) behavior of a RPV steel in the as-received, thermally-aged and irradiated conditions was evaluated in HTW.
The investigations were performed with the well characterized JRQ RPV steel (SA533 B Cl. 1, 0.017 % P, 0.14 % Cu, 0.004 % S). The material was thermally aged by a special step-cooling heat treatment (593 to 468 °C) that maximizes P grain boundary enrichment with respect to thermal ageing in LTO. The irradiation was done 35 years ago in the STILO facility at the SAPHIR test reactor at PSI (simulated PWR spectrum) with n-fluences ranging from 4E18 to 5E19 n/cm2 (E > 1 MeV) at 290 °C. The highest fluences resulted in transition temperature and yield stress (YS) shifts of 96 °C and 111 MPa. The test matrix involved SCC initiation and crack growth tests with as-received, thermally-aged and irradiated (4×1019 n/cm2) flat tapered tensile and pre-cracked fracture mechanics compact tension (CT) and Charpy specimens in simulated BWR/normal water chemistry (NWC) environment at 274 °C with 2 ppm dissolved oxygen (DO). The high DO results in a realistic free corrosion potential ECP of +150 mVSHE as on the RPV SS cladding with a cladding penetrating crack. Few tests in hydrogenated HTW (BWR/HWC) complemented the test matrix. The thermally-aged RPV steel revealed a similar SCC initiation susceptibility as the as-received material with SCC threshold stresses σ(SCC) in the range of the YS at the test temperature of 274 °C, which were very similar for both conditions. Due to the 100 MPa higher YS of the irradiated steel, its SCC threshold σ(SCC) is higher, but also likely in the range of the irradiated YS. The number of irradiated test samples was too small for solid conclusions in the latter case. The corrosion fatigue and SCC crack growth behavior of the as-received, thermally-aged and irradiated conditions were very similar. Thermal ageing and irradiation thus had no adverse effect on CF and SCC crack growth. In the SCC tests with pre-cracked charpy specimens, the SCC growth rates were affected by dK/da effects, which shifted the transition to fast SCC to lower KI-values with respect to standard CT samples.