Simulated Pressurized Water Reactor Research Articles

Quantification of SCC mechanisms in austenitic alloys under PWR primary water conditions

Abstract In order to achieve a full mechanistic understanding of stress corrosion cracking (SCC), the key operating mechanisms need to be identified but also quantified. In this study, we summarize and rationalize key findings from the last 15 years of high-resolution characterization of SCC in our group. A comprehensive characterization of a set of austenitic alloys with different Ni content and constant Cr level, tested under simulated pressurized water reactor (PWR) primary water conditions at various temperatures, has revealed evidence for at least two operating mechanisms: one diffusion-related and the other deformation-related. For their relevance to the nuclear industry, two additional alloys with increased Cr content were also studied (A800 and A690). Key precursors for SCC initiation and propagation are identified and their effect on alloy degradation discussed. A list of key materials’ properties that ensure low SCC susceptibility is proposed.

Hydrogen contribution to low temperature embrittlement during PWR exposure of an alloy 82 weld metal

The embrittlement of alloy 82 welds exposed to a simulated pressurised water reactor (PWR) environment was analysed and compared with that observed for cathodically hydrogen-precharged specimens. Similarities in mechanical behaviour, i.e. the evolution of the constitutive equations with a decrease in flow stress compared with non-precharged specimens, and in the fracture modes, with brittle zones in the outer layer, enabled us to conclude that the embrittlement observed in simulated PWR environment corresponded to a two-step hydrogen-assisted cracking mechanism. The first step corresponded to hydrogen absorption into the surface layer leading to crack initiation, and the second step to hydrogen localisation ahead of the crack tip promoting crack propagation.

Effect of microstructure and bending angle in cold bend pipe of SA 106B on flow accelerated corrosion behavior in simulated pressurized water reactor secondary circuit water environment

The flow-accelerated corrosion (FAC) behavior at different bending angles in bend pipe in simulated secondary circuit water environment of pressurized water reactor (PWR) was investigated. The results indicated that the FAC rates at different bending angles within the bend pipe strongly depended on the ferrite nanocrystallization degree and the mass transfer coefficients. Ferrite nanocrystallization played a dominant effect on FAC, and the mass transfer coefficient played a secondary role on FAC, causing the maximum FAC rate in bend pipe was transferred to the 45° position on the intrados.

Effects of pressurized water reactor environment and cyclic loading parameters on the low cycle fatigue behavior of 304L stainless steel

Austenitic stainless steels used in light water reactor coolant environments are susceptible to environmentally assisted fatigue due to non-monotonic loading conditions, primarily associated with load-follow, thermal transients, or intermittent plant shutdowns and start-ups. This study investigates the effects of a high-temperature pressurized water reactor (PWR) water environment and cyclic loading parameters on the low cycle fatigue behavior of austenitic 304L stainless steel. Prolonged exposure to a PWR environment and cyclic loading conditions such as a lower strain rate or a higher fraction of slow strain rate enhances the initiation and accelerates the crack growth rate of fatigue cracks, resulting in decreased fatigue life. The deformation-induced α'-martensite is observed in proximity to fatigue crack tips primarily in specimens tested in simulated PWR primary water, while cellular dislocation structures are more frequently observed near crack tips in specimens tested in high-temperature air. The deformation-induced martensitic transformation from γ-austenite to α'-martensite, occurring via the precursor ε-martensite phase, contributes to the accelerated fatigue crack growth rate in a PWR environment with hydrogen.

Short communication: “Effect of dissolved hydrogen on the early corrosion behavior of 316 stainless steel in simulated PWR environments”

In simulated pressurized water reactor environments, the corrosion test on 316 stainless steel was conducted for 40 to 4000 s, with 25 and 50 cc/kg-H2O dissolved hydrogen concentrations (NH and HH). It was found that the external Fe-rich oxide particles and the inner Cr-rich layer were simultaneously formed at the beginning of corrosion. However, the first formed Cr-rich layer was mainly Cr(OH)3 in the NH condition, while Cr2O3 in the HH condition. As exposure time increased, fraction of Cr2O3 increased due to transformation from Cr(OH)3 in both conditions. In addition, for HH condition, partial transformation to FeCr2O4 was observed.

Oxidation of Alloy 690 in Simulated Pressurized Water Reactor Primary Coolant – Experiments and Modeling

During normal operation of the primary loop of a water cooled-water moderated energy reactor (WWER), the associated water chemistry evolves from an initially high-boron high-potassium (beginning-of-cycle, 1200 ppm B and 8 ppm K) to significantly lower concentrations (ca. 400 ppm B and 3.5 ppm K) at the end of the respective campaign. In addition to that, the concentration of Li produced by transmutation of B reaches ca. 0.7 – 0.9 ppm. To the best of our knowledge, the effect of such water chemistry changes on pressurized water reactor (PWR) internals such as Alloy 690 have not been studied yet. Such a study appears relevant also within the frames of a discussion for a possible transition from B-Li to B-K water chemistry in PWRs.In the present paper, the effect of evolution of primary water chemistry during operation on the corrosion rate and conduction mechanism of oxide films on Alloy 690 is studied by in-situ impedance spectroscopy at 280 °C / 8 MPa during 1-week exposures in an autoclave connected to a re-circulation loop. Impedance measurements were performed in three-electrode mode featuring a Pt sheet counter electrode and a Pd electrode negatively polarized with a current of -10 µA vs. another Pt to approximate the reversible hydrogen electrode (RHE). Spectra were recorded at the corrosion potential in a frequency range 10 kHz – 0.1 mHz using an ac amplitude of 50 mV (rms). At the end of exposure, the samples were polarized in the range of potentials -0.1 to 0.5 V vs. RHE to evaluate the stability of the passive oxide. Separate samples of Alloy 690 were simultaneously exposed in the autoclave and subsequently analyzed by Glow-Discharge Optical Emission Spectroscopy (GDOES) in order to estimate the thickness and the in-depth composition of the formed oxides.Impedance data were quantitatively interpreted using the Mixed-Conduction Model for Oxide Films (MCM) to estimate the rates of metal oxidation at the alloy/oxide interface, oxide dissolution and restructuring at the film/coolant interface and ion transport in the protective layer. The effect of water chemistry evolution on corrosion rate and conduction mechanism in the oxide on Alloy 690 in primary coolant is discussed on the basis of the obtained parameters.Acknowledgments: This study is funded by the European Union-NextGenerationEU, through the National Recovery and Resilience Plan of the Republic of Bulgaria, project № BG-RRP-2.004-0002 (BiOrgaMCT).

The single effect of microstructure, residual strain and geometric structure on the stress corrosion cracking behavior of scratched alloy 690TT

Scratched area on alloy 690TT were treated by relief annealing (RA) and solution annealing plus thermal treatment (SATT), respectively, and then detailed characterization and immersion tests in simulated pressurized water reactor environment were carried out. RA reduces residual strain but not changes microstructure stratification, which causes stress corrosion cracking (SCC). SATT eliminates both strain and microstructure stratification, preventing SCC. Meanwhile, the independent effects of geometric structure, strain and microstructure on SCC behavior were investigated. Specifically, geometric structure is necessary but not sufficient for crack initiation. While the residual strain mainly enhances propagation, and the microstructure promotes both initiation and propagation.

Effects of Matrix and Dendrite Boundary Composition on Stress Corrosion Cracking Performance of Alloy 52M Buttering Layer in Simulated Pressurized Water Reactor Primary Water

The oxidation and stress corrosion cracking (SCC) behaviors of Alloy 52M buttering (52Mb) layer were studied in simulated pressurized water reactor (PWR) primary water after microstructure characterization. The 52Mb layer was first postweld heat treatment (1st-PWHT) and follow-up PWHT (FU-PWHT). Two dilution zones (DZs) were found in both the 1st-PWHT and FU-PWHT 52Mb samples. FU-PWHT decreased the Cr content in the first DZ, enhancing the oxidation rate and accelerating the formation of the local oxidation penetration zone at the oxide/substrate interface. Lower Cr content and high Fe content in the DZ were detrimental to the SCC resistance of the 52Mb in PWR primary water.

Effect of material chemical composition on oxidation and stress corrosion cracking of the submerged arc welding Alloy 52M overlay in a simulated PWR primary water

Compact tension (CT) specimens from submerged arc welding Alloy 52 M overlay are used to determine the stress corrosion cracking (SCC) resistance of the dilution zones and bulk overlay metal on the 52 M side in a simulated pressurized water reactor (PWR) primary water. SEM-EDS and X-ray fluorescence spectroscopy (XRF) results show three dilution zones (DZs) on the 52 M side. Oxidation test results show that the thickness of the compact oxide layer and maximum local oxidation penetration depth decrease for 1st DZ, 2nd DZ, and bulk 52 M, which are consistent with the average SCC growth rate and the number of local intergranular SCC cracks in these regions. The dilution zones with low Cr but high Fe content possess lower oxidation and SCC resistance than the bulk 52 M.

The structure dependence of grain boundary passivation of Alloy 690 in high temperature water

The oxidation behavior of grain boundaries (GBs) was investigated on Alloy 690 after exposure to simulated pressurized water reactor primary water. The tendency to form passive film above GB is positively related to Cr diffusivity along GB which is determined by GB structure. Highly-ordered GB like twin boundary is subject to penetrative oxidation as Cr diffusivity is very low. GBs with misorientation angle θ ≤17.4° can't support fast diffusion of Cr either and also show penetrative oxidation. When θ ranges between 17.4° and 28.7°, Cr diffusivity is enhanced a little and GB shows unstable passivation. Meanwhile, the Cr diffusivity along GB is also related to the atomic packing density of GB planes. GB with higher atom packing density exhibits lower Cr diffusivity and deeper intergranular oxide depth. As θ increases above 28.7°, Cr diffusion is fast enough to support the formation of protective film above GB.

Decreasing Hydrogen Content within Zirconium Using Au and Pd Nanoparticles as Sacrificial Agents under Pressurized Water at High Temperature.

Decreasing hydride-induced embrittlement of zirconium-based cladding is a significant challenge for the successful dry storage of spent nuclear fuel. Herein, to radically minimize hydride-induced embrittlement, we used nanoparticles as sacrificial agents with a greater affinity than zirconium for hydrogen. Corrosion experiments in the presence of gold (Au) and palladium (Pd) nanoparticles under simulated pressurized water reactor (PWR) conditions revealed that the hydrogen content of the zirconium samples was remarkably reduced, with a maximum decrease efficiency of 53.9% using 65 nm Au and 53.8% using 50 nm Pd nanoparticles. This approach provides an effective strategy for preventing hydride-induced embrittlement of zirconium-based cladding.

Frequency induced fretting corrosion mechanism evolution of Alloy 690 exposed to simulated secondary water

The effect of frequency (5–30 Hz) on the fretting corrosion behavior of Alloy 690 tube was investigated in simulated pressurized water reactor secondary water. The results show that the wear volume of the tube increases by ∼19 times with the frequency increases, and the wear mechanism changes from adhesive wear to delamination wear and abrasive wear. The increasing fretting frequency could affect the contact temperature and accelerates the change of contact stress, which cause a shift in the wear mechanism. Under the synergic effect of mechanical wear and corrosion, a gradient multi-layer structure is formed on the wear area, which is composed of the ultra-fine grain three-body layer, the tribologically transformed structure layer and matrix distributed from top to bottom. The three-body layer is mainly composed of Fe-rich spinel oxides and a small amount of Fe3O4 and Cr2O3, and its thickness increases with the increase of the fretting frequency.

Irradiation assisted stress corrosion cracking of 347 stainless steel in simulated PWR primary water containing lithium hydroxide or potassium hydroxide

The susceptibility of type 347 stainless steel to irradiation assisted stress corrosion cracking (IASCC) in pressurized water reactor (PWR) primary water with two water chemistries was evaluated. Miniature four-point bend tests were used to assess the effect of potassium hydroxide (KOH) in comparison to lithium hydroxide (LiOH) on the stress to initiate IASCC. Four-point bend specimens were machined from baffle-former bolts removed from the Salem unit 1 and D.C. Cook unit 2 reactors with accumulative doses ranging from 7.8 to 26.4 dpa. It was found that the pseudo-threshold stress to initiate IASCC in this alloy was 50–60% of the irradiated yield strength irrespective of water chemistry. Intergranular cracking was the dominant fracture mode in all specimens and the fracture morphology was characterized by a tortuous fracture path through the oxide and was common to both water chemistries. Grain boundary oxidation combined with localized stress enhancement at deformation band – grain boundary sites were precursors for IASCC initiation in both water chemistries. The pseudo-threshold stress for IASCC initiation is in good agreement with O-ring test results on both 316SS and 304SS. Overall, the stress to initiate cracks in neutron-irradiated 347SS was indistinguishable between simulated primary water containing LiOH or KOH.

Decoupling the roles of grain boundary oxidation and stress in IASCC of neutron-irradiated 304L stainless steel

Irradiation assisted stress corrosion cracking (IASCC) is known to be a combination of applied stress, and a corrosive environment in irradiated materials, but an appropriate understanding of how each of these components affects cracking is not yet clear. In this work, we isolate the role of grain boundary (GB) oxidation in IASCC. For this purpose, solution-annealed 304L stainless steel irradiated in reactor to 5.4 or 69 dpa was studied using a miniaturized four-point bend technique. Two modes of bend tests were conducted; straining in simulated pressurized water reactor primary water (PW) at 320 °C in an incremental manner, and pre-oxidation in PW at 320 °C without application of stress followed by dynamic straining in purified Ar at the same temperature until crack initiation. Exposure of the 5.4 dpa sample for 210 h in high temperature water initiated cracks in Ar at 60% of the yield stress (0.6YS), whereas the companion sample of the same dose exposed for 1010 h cracked in Ar at 0.5YS. The long-term exposure in water led to GB oxidation that ultimately lowered the crack initiation stress. Dynamic straining in water resulted in larger crack lengths and greater crack depths indicating stress accelerated oxidation and cracking. Dislocation pile-ups at dislocation channel-grain boundary sites provide an amplification of the applied stress to a level consistent with the fracture stress of the grain boundary oxide, providing an explanation for IASCC that occurs at applied stresses well below the irradiated yield stress of the alloy. There was no evidence of cracking upon straining of an unoxidized 69 dpa sample in Ar to well above yield, confirming that the irradiated state is not inherently susceptible to intergranular (IG) cracking and that oxidized GBs are responsible for initiation of IG cracking.

The effect of elastic and plastic strain on surface and intergranular oxidation of alloy 600 in simulated PWR primary water

The effects of deformation and stress on the surface and grain boundary oxidation of Alloy 600, exposed to primary water of Pressurized Water Reactors (PWRs), were investigated by Time of Flight Secondary Ion Mass Spectrometry (ToF-SIMS) and Scanning Electron Microscopy (SEM). It was found that tensile straining prior to oxidation does not modify the surface oxide thickness while the intergranular oxide depth increases when the strain reaches a critical value of 15%. The oxide growth rates at surface and grain boundaries are enhanced under stress. The effect of stress is more significant when the specimen is under elastic strain than plastically deformed.

The structure dependence of oxidation behavior of high-angle grain boundaries of alloy 600 in simulated pressurized water reactor primary water

The intergranular degradation of seven different types of high-angle grain boundaries (HAGBs) were investigated on Alloy 600 after exposure to simulated pressurized water reactor primary water. All boundaries are susceptible to preferential intergranular oxidation (PIO) except for ideal coherent twin boundary. Diffusion induced grain boundary migration (DIGM) normally occurs and its depth is positively correlated with the PIO extent. Interestingly, the PIO susceptibility is independent on the grain boundary misorientation angle or Σ value, but related to the grain boundary atom packing density (GBAPD). Grain boundaries with higher GBAPD values show higher PIO resistance as the element diffusion is slower.

High-resolution characterization of the fretting corrosion of Alloy 690 in the simulated secondary water of pressurized water reactor

The worn scars on Alloy 690 after the fretting corrosion testing in simulated pressurized water reactor (PWR) secondary water have been comprehensively analyzed by scanning transmission electron microscopy (STEM) and transmission Kikuchi diffraction (TKD). The high-quality characterization results experimentally show that the fretting wear accelerates the corrosion of Alloy 690 in two approaches. The first one is to break the integrity of the oxide scale by introducing cavities at the oxide grain boundaries. The second one is to alter the microstructure of the underneath matrix, forming a nano-grained matrix layer. The increased grain boundary density in this layer can accelerate the consumption of Cr in the near-surface matrix. The loss of oxide scale integrity and the accelerated Cr consumption are believed to contribute to the deteriorated corrosion resistance of Alloy 690 during the fretting corrosion process.

Investigation via Electron Microscopy and Electrochemical Impedance Spectroscopy of the Effect of Aqueous Zinc Ions on Passivity and the Surface Films of Alloy 600 in PWR PW at 320 °C

Aqueous zinc ions lower the corrosion rate of Alloy 600, which helps lower the radiation dose rate in pressurized water reactors (PWRs). The influence of zinc on the electrochemical behavior of Alloy 600 in PWR primary water (PW) at 320 °C was investigated using a combination of electron microscopy and electrochemical impedance spectroscopy (EIS). Secondary electron microscopy (SEM) and scanning transmission electron microscopy (STEM)/energy-dispersive X-ray spectroscopy (EDS) indicated duplex surface films were formed on the Alloy 600 in PWR PW with and without 100 ppb of zinc. There was no effect of zinc on the chromium-rich inner layer (IL) (of Cr2O3 and/or CrOOH). Zinc had a significant effect on the outer layer (OL). In the absence of zinc, a highly porous OL formed that was mostly composed of nickel oxide whiskers. In the presence of zinc, a zinc-containing, denser OL of oxide was formed. The EIS data were acquired in laboratory simulated PWR PW at 320 °C with and without 100 ppb zinc. The spectra were measured at nine different values of potential that spanned a 500 mV-wide range. The EIS indicated there was no effect of zinc on the oxidation rate of metals at the alloy/IL interface nor on the transport of ions through the IL. Zinc lowered the corrosion rate because the dense OL inhibited the release of nickel ions from the IL into the solution.

Performance and evolution of cold spray Cr-coated optimized ZIRLO™ claddings under simulated loss-of-coolant accident conditions

The performance of Cr-coated Optimized ZIRLO™ as accident tolerant fuel cladding material for pressurized water reactors (PWRs) is assessed. The coating oxidation mechanisms, oxide stability, and the transformation of the Cr-coating/Optimized ZIRLO™ interface are among the studied phenomena. For this purpose, samples were exposed at 1200°C in steam for 3 min, 20 min and 40 min. As-fabricated coated claddings, plus specimens tested in autoclave at 415°C for 90 days in simulated PWR water chemistry were employed for comparison. Characterization techniques such as scanning electron microscopy, energy dispersive x-ray spectroscopy, electron backscattered diffraction, and transmission electron microscopy were used to determine the chemistry and crystalline structure of the various phases formed during the different exposures. When exposed to loss-of-coolant accident (LOCA) conditions for 40 min, a layer of Cr2O3 up to 8 µm thick was measured on the outer surface of the Cr-coating. No significant oxidation of the underlaying Optimized ZIRLO™ alloy occurred, and the applied coating appears to be very effective at delaying the cladding degradation under accident conditions. At the coating-substrate interface, a 1–2 µm thick layer of (Cr,Fe)2Zr Laves phase was found. The presence of this phase appears to have no detrimental effects on the coating performance, and it might play a role in slowing down the dissolution of the coating into the substrate. ZrO2 particles were frequently found at grain boundaries in the coating after exposure to LOCA conditions. For longer exposure time, these particles are expected to grow into a ZrO2-network, creating a fast diffusion path for O, and compromising the oxidation protection offered by the coating.

Using Direct Current Potential Drop Technique to Estimate Fatigue Crack Growth Rates in Solid Bar Specimens under Environmental Assisted Fatigue in Simulated Pressurized Water Reactor Conditions

The direct current potential drop (DCPD) technique may be used in crack propagation tests to measure the crack growth rate (CGR). Potential probes attached to the specimen allow the variation of the crack length to be estimated. In this research, the DCPD technique using one single potential probe was applied to solid bar specimens (i.e., without any initial notch or crack) subjected to low-cycle fatigue testing in a simulated pressurized water reactor (PWR) environment. This particular analysis had two associated difficulties, the first one being the fact that crack initiation sites are not known beforehand, and the second one consisting in the experimental difficulties and conditioning factors associated with the simulation of the PWR environment. Nine solid bar specimens were tested to fatigue failure under different strain amplitudes and frequencies, while also measuring the corresponding DCPD signal during the fatigue process. It was observed that the initiation of multiple cracks was detected by the DCPD measurements. Moreover, as fatigue continued, one of the cracks became dominant and progressed to cause the specimen failure. The DCPD technique allowed the average CGR of the dominant crack to be estimated. Finally, the obtained average CGRs were validated by comparing them with average CGRs derived from striation spacing measurements, obtained from scanning electron microscopy (SEM) and from literature values gathered in the NUREG/CR-6909 document.