High-Temperature Water Effects on the Fracture Behaviour of Low-Alloy Reactor Pressure Vessel Steels

The structural integrity of the reactor pressure vessel (RPV) of light water reactors (LWR) is of utmost importance regarding operation safety and lifetime. High-temperature water (HTW) and hydrogen absorbed from environment, in synergy with irradiation embrittlement, dynamic strain aging (DSA), environmentally-assisted cracking (EAC) or temper embrittlement (TE) may reduce the fracture resistance of RPV steels. The fracture behaviour in the upper shelf region of low-alloy RPV steels with different microstructures and DSA, EAC and TE susceptibilities in various simulated LWR environments was evaluated by elastic-plastic fracture mechanics tests. In the reference RPV steels with low sulphur and phosphorus contents and low DSA, EAC and TE susceptibilities, environmental effects on fracture resistance are absent or marginal in both oxygenated and hydrogenated HTW. However, moderate but clear reduction of fracture initiation resistance occurred in: a) Simulated coarse grain heat-affected zone material with high yield stress, since a higher yield stress facilitates the hydrogen enrichment in the fracture process zone; b) Low-sulphur RPV steels with high DSA susceptibility, where the reduction of fracture initiation resistance increased with decreasing loading rate at 288 °C and was most pronounced in hydrogenated HTW due to the localization of plastic deformation by the interaction between DSA and hydrogen; c) High-sulphur RPV steels with high EAC susceptibility in aggressive occluded crevice environment (oxygenated HTW with addition of impurities) with preceding EAC crack growth, resulting from the enhancement of hydrogen availability and uptake; and d) High-phosphorus RPV steel with high TE susceptibility, where the reduction of fracture initiation resistance was most pronounced in hydrogenated HTW, indicating that TE effect dominates over effects of occluded crevice chemistry. In hydrogenated HTW, the reduction of fracture resistance correlated fairly well with the DSA susceptibility of different steels and sulphur content played no or a minor role. In oxygenated HTW, the reduction of fracture resistance increased with steel sulphur content for steels with low DSA susceptibility. Stable ductile transgranular tearing by micro-void coalescence dominates in both air and HTW environments. The fracture surface of specimens tested in air was mainly ductile. In contrast, specimens tested in HTW environments, varying amounts (a few %) of secondary cracking, macro-voids, quasi-cleavage and intergranular cracking were observed on fracture surface. The observed fracture modes and morphology suggest a combination of hydrogen-enhanced local plasticity and hydrogen-enhanced strain-induced vacancies mechanisms with minor contributions of hydrogen-enhanced decohesion embrittlement mechanism. The main reason for the moderate degradation effects is the low hydrogen availability in HTW together with a high density of (fine-dispersed and strong) hydrogen traps in RPV steels. In addition, the environmental reduction and softening by hydrogen and HTW environments was partially compensated by the toughness increase due to DSA. Although the environmental effects are moderate, effects can be more critical for RPV materials with low initial upper shelf toughness under unfavourable combinations (high sulphur content, increased strength, high EAC, TE and DSA susceptibilities) or old plants with small margins regarding irradiation embrittlement

A review of stress corrosion cracking of austenitic stainless steels in PWR primary water

Initial cases of stress corrosion cracking (SCC) in pressurized water reactors (PWRs) occurred mostly but not exclusively in stagnant areas like dead-legs, but recently more extensive IGSCC has occurred in normal free-flowing PWR primary water. Operational experience and laboratory data reveal that the main parameters in IGSCC include cold work and weld residual strain, oxygen, and residual and applied stress. Residual strain, which arises from manufacturing, surface grinding, and welding, should be limited by optimizing manufacturing procedures, minimizing alignment and fit-up stresses and using high-quality weld procedures. Preventing oxygen ingress in the make-up water should be pursued. Stresses created by thermal fluctuations (thermal mixing, low-leakage core operation, and start-ups) deserve more attention. Weld residual stress, fit-up stresses and local stresses from load follow must be maintained below the annealed yield stress. IGSCC should be considered in aging management and in-service inspection. Detection techniques capable of identifying IGSCC should be employed

Mechanistic understanding of the localized corrosion behavior of laser powder bed fused 316L stainless steel in pressurized water reactor primary water

The laser powder bed fused (LPBFed) stainless steels showed anomalous and localized corrosion behavior in the nuclear reactor high-temperature water compared to their wrought counterparts, which affects their performance during plant operation. In this study, advanced microstructural characterization was performed on LPBFed 316 L sample along with wrought 316 L sample after corrosion tests to understand the underlying mechanisms. The results showed that an inhomogeneous/discontinuous inner oxide layer formed on LPBFed 316 L, in contrast to the continuous inner oxide layer on the wrought 316 L specimen. This discontinuous inner oxide layer was identified to consist of Cr-enriched nano-sized spinel oxide and the barrier layer features a Ni-enriched hexagonal close-packed Laves phase. Localized/preferential oxidation was found to occur along the cellular walls which were tangled with high density dislocations and decorated with Mn and Si-enriched nano-sized precipitates, and the nano-precipitates were observed in the core of dispersed Cr-enriched inner oxide crystals

Effect of hydrogen on electrochemical behavior of additively manufactured 316L in pressurized water reactor primary water

The electrochemical behavior of laser powder bed fusion (LPBF) 316 L stainless steel subject to different heat-treatments (solution annealing and hot isostatic pressing) is compared to nuclear-grade wrought 316 L in pressurized water reactor primary water at 288 °C (with and without dissolved hydrogen) using current-time transients, cyclic voltammetry and electrochemical impedance spectroscopy (EIS). Analysis of spectra by the Mixed-Conduction Model revealed slower corrosion rates of LPBF 316 L than wrought 316 L, the effect being more pronounced in the presence of dissolved hydrogen. The characteristics of the barrier layer and the oxide film/coolant interface were irreversibly altered upon removal of dissolved hydrogen

Effect of hydrogen on electrochemical behavior of additively manufactured 316L in pressurized water reactor primary water

The electrochemical behavior of laser powder bed fusion (LPBF) 316 L stainless steel subject to different heat-treatments (solution annealing and hot isostatic pressing) is compared to nuclear-grade wrought 316 L in pressurized water reactor primary water at 288 °C (with and without dissolved hydrogen) using current-time transients, cyclic voltammetry and electrochemical impedance spectroscopy (EIS). Analysis of spectra by the Mixed-Conduction Model revealed slower corrosion rates of LPBF 316 L than wrought 316 L, the effect being more pronounced in the presence of dissolved hydrogen. The characteristics of the barrier layer and the oxide film/coolant interface were irreversibly altered upon removal of dissolved hydrogen