Study on SCC Propagation Behavior of a Dissimilar Metal Weld Joint of Nuclear Power Squib Valve in PWR Primary Water

This study aims to elucidate stress corrosion cracking (SCC) behavior in 316LN stainless steel (SS) and alloy 690 dissimilar metal weld joints of AP1000 pressurized water reactor (PWR) squib valves, critical for ensuring structural integrity in passive safety systems. It seeks to quantify crack growth rate (CGR) in the weld and heat affected zone (HAZ) under simulated PWR primary water environments, analyze the effects of temperature, corrosion potential, welding residual strain, and loading, and provide data to support aging management, in-service inspection, and life prediction of these safety-critical components. Experiments employed 0.5T compact tension specimens extracted from a DN450 single-shear cover squib valve mock-up, with weld and HAZ regions of nuclear-grade 316LN SS. Microstructure, hardness, and residual strain were characterized using metallographic analysis, microhardness testing, and electron backscatter diffraction (EBSD). SCC tests were conducted in a high-pressure autoclave replicating PWR primary water conditions (325 ℃, 15.5 MPa, 1 200 ppm B, 2.2 ppm Li), utilizing direct current potential drop (DCPD) for real-time CGR monitoring. Pre-cracking involved step-wise frequency reduction and waveform adjustment to ensure accurate crack propagation. Test parameters, including stress intensity factor (K), temperature (60-325 ℃), and water chemistry (dissolved oxygen or dissolved hydrogen), were systematically varied to assess their influence. Post-test fracture morphology and crack paths were analyzed via scanning electron microscopy (SEM) and EBSD. Results indicate that the HAZ exhibits heightened SCC susceptibility due to elevated welding residual strain, equivalent to 15%-25% cold work, with CGR of (5.6-7.0)×10⁻⁸ mm/s at K=20-40 MPa· \sqrt\mathrmm in oxygenated water, comparable to 20% cold work 316LN SS. The weld, containing ～6% δ-ferrite, shows reduced SCC sensitivity, with CGR around 4.5×10⁻⁸ mm/s, attributed to δ-ferrite’s crack-blocking effect. Temperature drives CGR following the Arrhenius law, with a 48 kJ/mol activation energy for the HAZ. Dissolved oxygen elevates CGR by an order of magnitude compared to dissolved hydrogen conditions, where weld crack growth nearly ceases. Loading influences HAZ CGR via a power-law relationship (CGR∝K^1.8 in dissolved oxygen, K^2.0 in dissolved hydrogen), while weld CGR exhibits minimal K dependence. Fracture morphology reveals intergranular cracking with secondary cracks in the HAZ under dissolved oxygen conditions and mixed intergranular-transgranular cracking in the weld. Crack path analysis confirms δ-ferrite’s role in impeding weld crack growth through bifurcation. The study concludes that the HAZ’s high residual strain renders it the most SCC-susceptible region, posing a significant risk to valve integrity, whereas the weld’s δ-ferrite enhances SCC resistance. Temperature and dissolved oxygen accelerate SCC, while dissolved hydrogen mitigates it, emphasizing the importance of water chemistry control. These findings provide critical data for optimizing aging management, refining in-service inspection protocols, and improving life prediction models to ensure the long-term reliability of nuclear squib valves in PWR environments.