High Temperature Water Environments Research Articles

Microstructural Evolution of 52M Weld Metal Near the Fusion Boundary and Oxide Films Formed in Simulated PWR Primary Water

Environmentally assisted cracking of structural materials such as dissimilar metal welds in high temperature water environments is strongly affected by the surface oxide film formed in the operating environments. In the present work, Nickel-base 52M weld metal samples are taken from two dilution zones and undiluted zone, and the microstructure was analyzed. Oxide films formed on these samples immersed in simulated PWR primary water at 325oC for 100h. The films are duplex-layered with the outer layer consisted of larger and scattered particles, and the inner layer consisted of small and dense particles. With increasing the distance from FB, the number of larger particles of outer layer decreased while the appearance of smaller particle of inner layer changed into needle-liked particles. With the increase of distance, the proportion of needle-liked particles increase, but the larger particles became less angular. The inner layer became thinner while the size of larger particles of outer layer became smaller with the distance from FB increases. The outer layer was Fe-rich while the inner layer was Cr-rich. The change of element concentration resulted in different oxidization behavior in PWR primary water. The formation of larger particles of outer layer can influence the appearance of inner oxide film. Figure 1

XPS Analysis of the Oxide Films Formed on Non-Charged and Hydrogen-Charged 316L SS in Deaerated High Temperature Water

Austenitic stainless steel components have been used in pressurized water reactors (PWRs) coolants. These austenitic stainless steels may absorb hydrogen from either the dissociation of dissolved hydrogen in PWR primary water or the corrosion processes at the steel surfaces during the immersion in PWR primary water. It is necessary to investigate the effect of hydrogen in alloys on the oxidation and resultant environmental degradation processes in high temperature water environments. In the present study, the oxide films formed in deaerated PWR primary water at 300℃on a cold worked 316L (316LCW) stainless steel with or without prior hydrogen-charging were characterized. Hydrogen-charging of 316LCW was performed by cathodic charging in sulfuric acid solution containing thiourea at room temperature. Prior to the immersion tests, the Immersion tests were performed in a 2-L Ti-autoclave with solution deaerated by purging N2 gas before heating to test temperature. The oxide films formed on 316LCW and hydrogen-charged 316L CW (316LCW-H) specimens after exposure to deaerated PWR primary water at 300 oC for168 h were observed. The XPS depth profiles of the oxide films formed on 316LCW and 316LCW-H specimens after exposed to simulated PWR primary water are compared. The thickness of the oxide film was estimated from the half height of oxygen concentration and the sputtering rate (0.1 nm/s), for determining the thickness of the oxide film that was about 180 nm on 316LCW and about 250 nm on 316LCW-H. The oxide films formed on both 316LCW and 316LCW-H were enriched in Fe and Ni in the outer layer, and enriched in Cr in the inner part of the oxide film. The Cr content in the inner oxide layer formed on 316LCW-H specimen was lower than that on 316LCW specimen. Figure 1

Environmental degradation of fracture resistance in high-temperature water environments of low-alloy reactor pressure vessel steels with high sulphur or phosphorus contents

The fracture behaviour of low-alloy reactor pressure vessel steels with various sulphur and phosphorus contents, different environmentally-assisted cracking (EAC) and temper embrittlement (TE) susceptibilities was evaluated by elastic plastic fracture mechanics tests in air and various simulated light water reactors environments. A moderate but clear reduction of fracture initiation resistance occurred in a) high-sulphur steel with high EAC susceptibility in oxygenated high-temperature water with aggressive occluded crevice environment and preceding EAC growth, and in b) high-phosphorus steel with high TE susceptibility, where the reduction of fracture resistance was most pronounced in hydrogenated high-temperature water.

Effect of dynamic strain ageing on environmental degradation of fracture resistance of low-alloy RPV steels in high-temperature water environments

Elastic plastic fracture mechanics tests on different low-alloy reactor pressure vessel (RPV) steels in air and various simulated light water reactor environments revealed a clear, but moderate reduction of the fracture initiation resistance of RPV steels with high susceptibility to dynamic strain ageing (DSA) in high-temperature water (HTW). In HTW, the reduction of fracture initiation resistance increased with decreasing loading rate and was most pronounced in hydrogenated HTW. The stronger reduction of fracture resistance in hydrogenated HTW was mainly due to the localization of plastic deformation by interaction between DSA and hydrogen.

Effects of Temperature Variation on Mechanics Field at Crack Tip of Stress Corrosion Cracking

Creep strain near crack front zone is one of the main driving forces in nuclear structural materials of stress corrosion crack propagation. To understand the effect of temperature variation on mechanical field near crack tip of nuclear structure material in high temperature water environment, compact tensile specimen was taken as an research object and ABAQUS finite element analysis software was also adopted, the Mise stress, plastic strain and creep strain near SCC crack tip region was studied. Results show that the area of the high-stress region decreases gradually with the increasing of the creep temperature, and its declining rate was faster and faster. The effect of temperature variation on the plastic strain is not very obvious. Temperature only has an obvious effect on a small scale near crack tip. With the increase of the distance from the crack tip, the effect of temperature on creep is gradually reduced.

Effect of Cold Working on the Residual Stress around the Notch front of Stress Corrosion Cracking

Failures show that the austenitic stainless steel (SS) pipes with produced high weld residual stresses are more susceptible to stress corrosion cracking in the simulated high temperature water environments of pressurized water reactor. To understand the effect of cold working on the residual stress and strain of stress corrosion cracking in 304SS, the stress-strain field near the crack front of 304SS under different degree of cold working was simulated by ABAQUS. Results show that the higher tensile residual stress and the larger compressive residual stress simultaneously occurs in the lower cold working material, and cold working has greater effect on the residual stress than on the residual strain. The residual compressive stress and residual compressive strain both decrease rapidly in a small range around the notch fronts and tend to be consistent in a more larger distance. But cold working has little effect on the residual strain.

Effect of Crack Length on Mechanical Field of Stress Corrosion Crack Tip of Cold Working 316L Stainless Steel

The effect of crack length at different cold working level on stress and strain state and fracture parameters of stress corrosion cracking tip of 316L stainless steel in high temperature water environment of the nuclear pressure vessel was studied by sub-model technology of elastic plastic finite element method. It combined with the mechanical properties of 316L stainless steel under different cold working level, and the effect of crack length on the Mises stress, equivalent plastic strain, tensile stress, tensile strain and J integral of 316L stainless steel under different cold working level was compared. The results indicate that crack length of stress corrosion cracking has certain effect on stress and strain state at crack tip. With the increase of crack length, Mises stress and equivalent plastic strain, tensile stress and tensile strain increase at crack tip, at the same time, J integral increases with increase of crack length. Cold working level has changed the mechanical properties of 316L stainless steel. When the stress intensity factor K is constant, with the increase of cold working level, Mises stress increases and equivalent plastic strain decreases at crack tip, while tensile stress increases and tensile strain decreases at crack tip. The rate of change of stress strain curve from 0° to 90° is different because of cold working level. The higher the level of cold working, the faster the Mises stress and the tensile stress increase at 90° in front of the crack tip, and equivalent plastic strain and tensile strain decrease most slowly at 90° in front of the crack tip. As the degree of cold working increases, the J integral of crack tip also increases. The increase of crack length and the level of cold working aggravate the stress corrosion cracking of 316L stainless steel under high temperature and high pressure in a certain range.

Effect of Si-Addition in Stainless Steel on Surface Oxidation in High Temperature Water Environments

Being used as the internal components of pressurized water reactor (PWR), austenitic stainless steels (SSs) are susceptible to irradiation-assisted stress corrosion cracking (IASCC) that is a critical issue for the long term safe operation. Radiation induced segregation (RIS) at grain boundaries in austenitic SSs, such as the depletion of Cr and enrichment of Ni and Si, is thought to contribute to IASCC. Studying the oxidation behavior of the model alloys simulating grain boundary RIS in high temperature water could provide evidence for the effect of RIS on IASCC. In this study, the model alloys 9Cr-25Ni-1.5Mo-5Si and 9Cr-25Ni-1.5Mo-3Si were used to simulate the depletion of Cr and enrichment of Ni and Si at grain boundaries of irradiated 316L SS. The oxide film properties of 316L SS and model alloys in of high temperature water environments were investigated. After the immersion in high-temperature water environments, the surface morphologies of the oxide films formed on the immersed specimens were examined using JEOL JSM-7500F field emission scanning electron microscope (FESEM). The oxide films were analyzed using INVIA laser Raman spectrometer with laser wavelength of 514.5 nm. The compositions of oxide particles were analyzed by the energy-dispersive spectrometry (EDS) attached to the FESEM. Fig.1 shows the SEM morphologies of the 316L SS, 9Cr-25Ni-1.5Mo-3Si and 9Cr-25Ni-1.5Mo-5Si alloy surfaces after immersion in pure water at 290 oC (pH at 290 oC of about 5.6) for 120h. The surfaces of all the test specimens were covered with polyhedral particles of different sizes. Larger size outer layer oxides are formed on 9Cr-25Ni-1.5Mo-3Si and 9Cr-25Ni-1.5Mo-5Si specimen surface. On the surface of 316L SS, there were only sparsely distributed large-size particles. Si addition in the alloy matrix favors the formation of the oxide layer in this environment. Fig. 2 shows the SEM morphologies of 316L SS, 9Cr-25Ni-1.5Mo-5Si and 9Cr-25Ni-1.5Mo-3Si alloys surfaces after the immersion in high temperature water of pH 9.5 (pH adjusted with monoethanolamine) at 290 oC for 24h, 48h and 120h. Compact and homogeneous oxide particles were formed on the surfaces of 316L SS and model alloys. The size of the oxide particles on model alloys was smaller than that on 316L SS. The oxide particle size of 9Cr-25Ni-1.5Mo-5Si and 9Cr-25Ni-1.5Mo-3Si was similar. The morphologies of particles became more stereoscopic with increasing immersion time. The outer layer oxide particles formed on the 316L SS and model alloys decreased with increasing solution pH value. The pH value has a significant influence on the oxidation of 316L SS and model alloys. These results show that the effect of Si in the alloy on oxide film properties is strongly dependent on the water chemistry. Si in the alloy facilitated the formation of large-sized outer layer in 290 oC water of pH 5.6, which refined the outer layer oxide particles in 290 oC water of pH 9.5. High solubility of Si-bearing oxide in high temperature water can modify the oxide film on the surface. These results show that Si in the alloy would play a role in IASCC through the modification of oxide film. Figure 1

Environmental Degradation Effect of High-Temperature Water and Hydrogen on the Fracture Behavior of Low-Alloy Reactor Pressure Vessel Steels

Structural integrity of reactor pressure vessel (RPV) in light water reactors (LWR) is of highest importance regarding operation safety and lifetime. The fracture behaviour of low-alloy RPV steels with different dynamic strain aging (DSA) & environmental assisted cracking (EAC) susceptibilities in simulated LWR environments was evaluated by elastic plastic fracture mechanics tests (EPFM) and by metallo- and fractographic post-test analysis. Exposure to high temperature water (HTW) environments at LWR temperatures revealed only moderated reductions in the fracture initiation and tearing resistance of low alloy RPV steels with high DSA or EAC susceptibility, accompanied with a moderate, but clear change in fracture morphology, which indicates the potential synergies of hydrogen/HTW embrittlement with DSA and EAC under suitable conditions. The most pronounced degradation effects occurred in a) RPV steels with high DSA susceptibility, where the fracture initiation and tearing resistance reduction increased with decreasing loading rate and were most pronounced in hydrogenated HTW and b) high sulphur steels with high EAC susceptibility in aggressive occluded crevice environment and with preceding fast EAC crack growth in oxygenated HTW. The moderate effects are due to the low hydrogen availability in HTW together with high density of fine-dispersed hydrogen traps in RPV steels. Stable ductile transgranular tearing by microvoid coalescence was the dominant failure mechanism in all environments with additional varying few % of secondary cracks, macrovoids and quasi-cleavage in HTW. The observed behavior suggests a combination of plastic strain localisation by the Hydrogen-enhanced Local Plasticity (HELP) mechanism, in synergy with DSA, and Hydrogen-enhanced Strain-induced Vacancies (HESIV) mechanism with additional minor contributions of Hydrogen-enhanced Decohesion Embrittlement (HEDE) mechanism.

Study on Stress Corrosion Behavior of Domestic Forged 316L Stainless Steel in High Temperature and High Pressure Water Environment

Study on Stress Corrosion Behavior of Domestic Forged 316L Stainless Steel in High Temperature and High Pressure Water Environment

The role of grain boundary microchemistry in irradiation-assisted stress corrosion cracking of a Fe-13Cr-15Ni alloy

A novel explanation based on the variation of grain boundary composition is proposed to elucidate the localized nature of irradiation-assisted stress corrosion cracking, i.e., intergranular cracking is observed only at specific sites of random high-angle grain boundaries, in a Fe-13Cr-15Ni austenitic model alloy subjected to proton irradiation and straining in a high-temperature water environment. Specifically, this work presents electron microscopy characterization of multiple cracked and un-cracked grain boundary sites and the neighboring oxides. The depletion of Cr and enrichment of Ni, as expected due to radiation-induced segregation, is observed at grain boundaries far from the crack tip (for cracked sites) or the metal surface (for un-cracked sites), and such modification of grain boundary composition is enhanced in the vicinity of the corrosion reaction front. Unexpectedly, grain boundary sites beyond the crack tip always present a lower Ni content and higher Cr content than the un-cracked sites on the same grain boundary. Overall, it is proposed that the site-specific susceptibility to stress corrosion cracking is governed by the grain boundary microchemistry, which determines not only the quantity of Cr, but also the efficiency of Cr transport to the reaction front and thus the protectiveness of the inner oxide. These effects of local composition may be further coupled with the local structure of and the local stresses interacting with the random high-angle grain boundaries.

Effects of zinc injection on stress corrosion cracking of cold worked austenitic stainless steel in high-temperature water environments

In the nuclear power industry, few data exist to demonstrate the effectiveness of Zn injection in reducing the crack growth rate of primary component materials. This work shows that Zn injection reduced the stress corrosion cracks growth rate of cold worked 316L stainless steel (SS) by 3 times. ZnCr2O4 oxide was detected both on the crack walls and at the crack tip, showing that Zn substitution in the spinel-type oxide occurred and mitigated stress corrosion cracking.

Corrosion Fatigue Crack Growth Behavior of Alloy 690 in High-Temperature Pressurized Water

Corrosion fatigue crack growth behavior of Alloy 690 was investigated in air and high-temperature pressurized water environments. The stress intensity factor amplitude threshold (ΔKth) was obtained, and the effects of stress intensity factor (ΔK), cyclic frequency (f), loading ratio (R), and material orientation on the corrosion fatigue crack growth rates were studied. The observed environmental acceleration on corrosion fatigue crack growth was quantitatively analyzed, and an empirical formula and the Ford-Andresen model were utilized to describe and explain the corrosion fatigue behavior of Alloy 690.

Effect of Grain Size on Corrosion Fatigue Behaviors of Primary Coolant Pipes of Nuclear Power Plants

Corrosion fatigue is one of the most important failure modes of primary coolant pipe used in nuclear power plant, due to the thermal cyclic stress caused by a variety of start-up and shutdown as well as transients when severed in the high temperature and pressure water environment. And 316LN stainless steel is one of the main materials used in pressurized water reactor in nuclear power plant. The mechanisms of fatigue crack initiation and propagation were elucidated by investigating the composition properties of oxide films formed in simulated service environment. The effects of grain size on the fatigue life and crack initiation mechanism of 316LN stainless steel in 320°C water environment were also investigated. The results indicated that the specimens with fine grains (30μm) own the longest fatigue lives than those with intermediate (80μm) and coarse grains (210μm). The fatigue stress amplitudes of the specimens increased with the grain refinement, although the fatigue lives of the specimens with intermediate and coarse grains were close. Fatigue cracks were often initiated at the persistent slip bands on the surfaces of the specimens with intermediate and coarse grains, while it was not easy to be initiated on the fine grain specimens. Finally, combined with the experimental results and the actual production of industry, a suggestion for the production of the primary coolant pipe is put forward.

Influence of Nickel-Based Alloys' Mechanical Properties on Mechanochemical Effect at Crack Tip in High Temperature Water Environments

Mechanochemical effect which is the mechanical and chemical interaction can accelerate the stress corrosion cracking (SCC) in nickel-based alloys used in nuclear power plants. The mechanical property heterogeneity in weld joints will influence the mechanochemical effect indirectly. The influences of yield strength and hardening exponent of nickel-based alloy 600 on the mechanochemical effect of crack tip surface in high temperature water environment were studied by adopting one inch compact tension specimen and finite element method. The influence of elastic and plastic deformation on the mechanochemical effect at crack tip was discussed. The results indicate that the mechanochemical effect is affected by the yield strength. In contrary, the hardening exponent change of alloy 600 has an insignificant influence on mechanochemical effect.

Effects of Thermal Aging on Material Properties, Stress Corrosion Cracking, and Fracture Toughness of AISI 316L Weld Metal

Thermal aging and consequent embrittlement of materials are ongoing issues in cast stainless steels, as well as duplex, and high-Cr ferritic stainless steels. Spinodal decomposition is largely responsible for the well-known “748 K (475 °C) embrittlement” that results in drastic reductions in ductility and toughness in these materials. This process is also operative in welds of either cast or wrought stainless steels where δ-ferrite is present. While the embrittlement can occur after several hundred hours of aging at 748 K (475 °C), the process is also operative at lower temperatures, at the 561 K (288 °C) operating temperature of a boiling water reactor (BWR), for example, where ductility reductions have been observed after several tens of thousands of hours of exposure. An experimental program was carried out in order to understand how spinodal decomposition may affect changes in material properties in Type 316L BWR piping weld metals. The study included material characterization, nanoindentation hardness, double-loop electrochemical potentiokinetic reactivation (DL-EPR), Charpy-V, tensile, SCC crack growth, and in situ fracture toughness testing as a function of δ-ferrite content, aging time, and temperature. SCC crack growth rates of Type 316L stainless steel weld metal under simulated BWR conditions showed an approximate 2 times increase in crack growth rate over that of the unaged as-welded material. In situ fracture toughness measurements indicate that environmental exposure can result in a reduction of toughness by up to 40 pct over the corresponding at-temperature air-tested values. Material characterization results suggest that spinodal decomposition is responsible for the degradation of material properties measured in air, and that degradation of the in situ properties may be a result of hydrogen absorbed during exposure to the high-temperature water environment.

Micro-mechanical State at Tip of Environmentally Assisted Cracking in Nickel-based Alloy

Nickel-based alloys and austenitic stainless steels are widely used in the structures of primary circuit of nuclear power plants. Environmentally assisted cracking (EAC) of these materials is one of the most significant potential safety hazards in the primary circuit of nuclear power plants. Researches show that EAC in nickel-based alloy is a process of oxide film rupture and reform at the tip of EAC in the high temperature water environment of nuclear power plants. To understand the micro-mechanical state at the tip of EAC, the stress-strain in the oxide film and the base metal at the EAC tip was simulated and discussed using a commercial finite element analysis code, which provides a foundation to improve the quantitative predication accuracy of EAC growth rate of nickel-based alloys and austenitic stainless steels in the important structures of nuclear power plants.

Effect of hydrogen on tensile behavior of low alloy steel in the regime of dynamic strain ageing

Low alloy steels typically used for reactor pressure vessel (RPV) in light water reactors may undergo different degradations and ageing mechanisms during service like fatigue, strain-induced corrosion cracking and corrosion fatigue or irradiation embrittlement, the latter being recognized as life limiting factor. There is growing concern that hydrogen, absorbed from the high temperature water environment and corrosion reactions, may potentially reduce toughness of RPV steels in synergy (or competition) with other embrittlement mechanisms like irradiation embrittlement, thermal ageing or dynamic strain aging (DSA). Strain rate, temperature and occurrence of DSA in these steels may affect the severity of the effect of hydrogen on toughness. The present investigation was envisaged to characterize the effect of hydrogen on tensile and fracture behavior of low alloy RPV steels at different strain rates and temperatures with a special emphasis on the synergy between DSA and hydrogen embrittlement. For this reason, tensile tests were carried out with as-received and hydrogen pre-charged specimens between 25 and 400 °C and at strain rates between 10-1 and 10-6 s-1. The fracture mode was evaluated by detailed post-test fractography in the scanning electron microscope. DSA in these steels was established by the occurrence of serrations, negative strain-rate sensitivity and a maximum/minimum in strength/ductility at intermediate temperatures and strain rates. The DSA peak and range were found to be shifted to lower temperatures with decreasing strain rates and vice-versa. The hydrogen pre-charging resulted in marginal softening and strain-rate dependent reduction in ductility at 250/288 °C. The hydrogen embrittlement and reduction in ductility were more pronounced and the strain rate range for hydrogen embrittlement significantly extended in the RPV steel with higher DSA susceptibility demonstrating some synergy between DSA and hydrogen effects, probably due to the localization of plastic deformation. In presence of hydrogen, shear dominated ductile fracture (microvoid coalescence) with varying amounts of quasi-cleavage regions and secondary cracking along the prior austenite grain boundaries were observed. A detailed investigation on these aspects and a tentative mechanistic explanation is presented in this paper.

Tribotester for High-Temperature and High-Pressure Water Environment

Tribotester for High-Temperature and High-Pressure Water Environment

Estimation of Stress Corrosion Cracking Initiation and Propagation in High-Pressure, High-Temperature Water Environment Utilizing Acoustic Emission

The stress corrosion cracking (SCC) of type 304 stainless steel during a creviced bent beam (CBB) test in a high-temperature, highpressure water environment was monitored by an optical fiber acoustic emission (AE) sensor. By examining the number and type of cracks formed by different durations of CBB testing an SCC initiation and propagation behavior model was developed. This model was then used to examine the AE event rate, which was found to agree well with the observed changes in the number and size of the cracks formed. Through analysis of the relationship between crack size and AE, the weaker AE signals were assumed to be produced by the initiation of small cracks, whereas the coalescence of cracks produces moderate and large AE signals. On the basis of these findings, it is concluded that the SCC propagation behavior can be reliably estimated from the event rate and amplitude of AE signals. [doi:10.2320/matertrans.M2014357]