New insights into the influence of permeated H on stress corrosion cracking of Alloy 690 in high-temperature water

Stress corrosion cracking (SCC) of Alloy 690 in high-temperature water is critical to the integrity of nuclear components, yet the influence of permeated H on SCC mechanism in high-temperature water remains unclear. Slow strain rate tests were conducted on solution-annealed Alloy 690 with and without internal H charging in high-temperature water. Permeated H promotes the initiation and propagation of SCC through chemical-mechanical synergistic effects. Chemically, H suppresses Cr diffusion in diffusion-induced grain boundary migration (DIGM) zones, weakens grain boundary (GB) cohesion, and raises CrOOH accumulation at crack tips. Mechanically, H lowers the geometrically necessary dislocations (GND) threshold for SCC initiation, intensifies dislocation pile-up and slip blocking, thus inducing SCC in low-angle grain boundaries (LAGBs). In nuclear power plants (NPPs), it is necessary to consider the impact of permeated H on the slip-dissolution mechanism to evaluate the behavior of SCC in high-temperature water.