Effects of cold deformation on electrochemical corrosion behaviors of 304 stainless steel

2017 ◽  
Vol 64 (2) ◽  
pp. 252-262 ◽  
Author(s):  
Jiamei Wang ◽  
Le Fu Zhang
Keyword(s):  
Stainless Steel ◽  
Nuclear Power ◽  
Cold Deformation ◽  
Water Environment ◽  
Electrochemical Corrosion ◽  
Pressurized Water ◽  
Content Type ◽  
304 Austenitic Stainless Steel ◽  
Primary Water ◽  
Corrosion Behaviors

Purpose The purpose of this study was to investigate the effects of deformation-induced martensite on electrochemical corrosion behaviors of 304 austenitic stainless steel in a simulated primary water environment of a pressurized water reactor nuclear power plant with boric acid and lithium hydroxide contaminated with chloride by potentiodynamic polarization, electrochemical impedance spectroscopy (EIS), Mott–Schotty curves and X-ray photoelectron spectroscopy (XPS). Design/methodology/approach The effects of deformation-induced martensite transformation on electrochemical corrosion behaviors of 304 austenitic stainless steel was investigated in a simulated primary water environment of a pressurized water reactor nuclear power plant with boric acid and lithium hydroxide contaminated with 0.1 M Cl− by potentiodynamic polarization, EIS, Mott–Schotty curves and XPS in this paper. Findings The results revealed that the martensitic phase contents increased with the level of cold deformation. The general corrosion current density and the corrosion potential increased and decreased, respectively, with the increase of cold deformation degree. However, the pitting potential decreased as the cold deformation increased up to 20 per cent, then a slight increase was observed at 35 per cent cold working. It was found from Mott–Schottky curves and XPS analysis that as the cold deformation degree increased from 0 to 35 per cent, the doping concentrations of the oxide films increased; however, the film thickness decreased, which indicates that both density and integrity of the films are degraded significantly as the deformation degree increases, and this ultimately contributes to the significant increment of the general corrosion rate and reduction of the pitting corrosion resistance. Originality/value The effects of deformation-induced martensite transformation on electrochemical corrosion behaviors of 304 austenitic stainless steel was investigated in a simulated primary water environment of a pressurized water reactor nuclear power plant with boric acid and lithium hydroxide contaminated with 0.1 M Cl− by potentiodynamic polarization, EIS, Mott–Schotty curves and XPS in this paper.

Study on electrochemical behavior of 690 alloy with corrosion products in simulated PWR primary water environment

2018 ◽  
Vol 65 (6) ◽  
pp. 616-625
Author(s):  
Xuequn Cheng ◽  
Xiao Mei Zuo
Keyword(s):  
High Pressure ◽  
High Temperature ◽  
Electrochemical Behavior ◽  
Water Environment ◽  
Corrosion Products ◽  
Inhibition Mechanism ◽  
Pressurized Water ◽  
Content Type ◽  
Primary Water ◽  
Immersion Testing

Purpose The purpose of this paper is to study the electrochemical behavior of 690 alloy with corrosion products in simulated pressurized water reactor (PWR) primary water environment. Design/methodology/approach This paper opted for a laboratory study using simulation of high temperature and high pressure environment immersion testing. The electrochemical behavior was studied by potentiodynamic polarization, electrochemical impedance spectroscopy, scanning Kelvin probe microscopy (SKP). Moreover, the corrosion products were analyzed by X-ray photoelectron spectroscopy. Findings The results demonstrated that the particle majority in the 690 alloy corrosion products subsequent to high temperature and high pressure immersion testing were mainly oxides of Fe and Ni, which protected the matrix. As the immersion testing duration increased, the corrosion potential of the 690 alloy apparently increased, and the corrosion current density de'creased, while the corrosion resistance Rf increased gradually along with the density. The SKP demonstrated that the EKP increased by nearly 400 mV from −0.42 to −0.03 V following the immersion testing, indicating that the corrosion product film played an apparent protective role on the substrate. Originality/value This paper provides a theoretical basis for the corrosion behavior and inhibition mechanism of 690 alloy in PWR primary water environment.

Mean Stress Effect on Fatigue Properties of Type 316 Stainless Steel in Pressurized Water Reactors Primary Water Environment

2019 ◽  
Vol 141 (5) ◽  
Author(s):  
Masayuki Kamaya
Keyword(s):  
Stainless Steel ◽  
Fatigue Life ◽  
Water Environment ◽  
Stress Effect ◽  
Mean Stress ◽  
Pressurized Water ◽  
316 Stainless Steel ◽  
Mean Stress Effect ◽  
Primary Water ◽  
The Mean

The mean stress effect on the fatigue life of type 316 stainless steel was investigated in simulated pressurized water reactor (PWR) primary water and air at 325 °C. The tests in air environment have revealed that the fatigue life was increased with application of the positive mean stress for the same stress amplitude because the strain range was decreased by hardening of material caused by increased maximum peak stress. On the other hand, it has been shown that the fatigue life obtained in simulated PWR primary water was decreased compared with that obtained in air environment even without the mean stress. In this study, type 316 stainless steel specimens were subjected to the fatigue test with and without application of the positive mean stress in high-temperature air and PWR water environments. First, the mean stress effect was discussed for high-temperature air environment. Then, the change in fatigue life in the PWR water environment was evaluated. It was revealed that the change in the fatigue life due to application of the mean stress in the PWR water environment could be explained in the same way as for the air environment. No additional factor was induced by applying the mean stress in the PWR water environment.

Influence of PWR Primary Water on LCF Behavior of Type 304L Austenitic Stainless Steel at 300°C: Comparison With Results Obtained in Vacuum or in Air

2012 ◽  
Author(s):  
Laurent De Baglion ◽  
José Mendez ◽  
Jean-Alain Le Duff ◽  
André Lefrancois
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Strain Amplitude ◽  
Environmental Effects ◽  
Water Environment ◽  
Dynamic Strain ◽  
Strain Rates ◽  
Pressurized Water ◽  
Fatigue Lifetime ◽  
Primary Water

Nowadays, it is well known that the low cycle fatigue (LCF) life of austenitic stainless steels can be affected in specific conditions of temperature, strain rate, strain amplitude or dissolved oxygen concentration by the effect of Pressurized Water Reactor (PWR) primary coolant environment. Nevertheless, questions remain about the best methodology that must be used to consider environmental effects for nuclear power plant licensing and for operating lifetime extensions. These environmental effects are most commonly evaluated from a mean fatigue curve based on tests conducted in air at room temperature. However, it is well established that air is not a neutral environment for metallic alloys and its effect can be highly dependent on the temperature level. Thus, in order to evaluate the intrinsic fatigue resistance of a 304L austenitic stainless steel at 300°C and the importance of complex fatigue – environment interactions in air or in PWR water, LCF tests were performed in both environments and specifically designed ones were conducted in secondary vacuum. Tests were performed on 304L cylindrical specimens at 20 or 300°C in vacuum or in air and only at 300°C in PWR water, under total axial strain control using a triangular waveform at strain amplitudes of ±0.3 or ±0.6% and strain rates of 4 × 10−3, 1 × 10−4 or 1 × 10−5 s−1. It was found that compared with vacuum, air is responsible for a strong decrease in fatigue lifetime in this steel, especially at 300°C and low strain amplitude. The PWR water coolant environment is still more active than air and leads mainly to increased damage kinetics, with slight effects on initiation sites or propagation modes. More precisely, the decreased fatigue life in PWR water is essentially attributed to an enhancement of both crack initiation and “short crack” micropropagation stages. Furthermore, a detrimental influence of low strain rates on the fatigue lifetime at 300°C was observed in PWR water environment or in air, but also in vacuum without environmental effects, and was in the last case exclusively attributed to the occurrence of the dynamic strain aging (DSA) phenomenon. So, the use of data obtained in a neutral environment as a reference allows the evaluation of the intrinsic effect of each environmental or loading condition. Moreover, in an active environment such as air or PWR primary water, damage evolutions as well as fatigue lives cannot be predicted by a simple multiplication of each parameter effect taken separately because they are the result of numerous interactions. The last conclusion is supported by complementary results showing that the PWR water environment effect as well as the ground surface finish effect can be attenuated when LCF tests are performed with a more representative loading signal shape.

Over Limit Oxygen Effect on Crack Propagation of WWER 440 RPV Steel in Primary Circuit Water

2009 ◽  
Author(s):  
Miroslava Ernestova ◽  
Anna Hojna
Keyword(s):  
Crack Propagation ◽  
Nuclear Power ◽  
Power Plants ◽  
Water Environment ◽  
Nuclear Research ◽  
Compact Tension ◽  
Oxygen Effect ◽  
Primary Circuit ◽  
Nuclear Research Institute ◽  
Primary Water

Experience with operating nuclear power plants worldwide reveals that many failures may be attributed to fatigue associated with mechanical loading due to vibration and with corrosion effect due to exposure to high-temperature environment. In order to clarify the simultaneous influence on reactor pressure vessel (RPV) material testing of ferritic steel 15Ch2MFA used for RPV of WWER 440 was performed at Nuclear Research Institute (NRI) autoclaves. Cyclic and constant loadings were applied to Compact Tension (CT) specimens in WWER primary water environment at 290°C and simultaneous effect of different oxygen levels (< 20 ppb, 200 ppb, 2000 ppb) on crack propagation has been evaluated. Obtained crack growth rates are compared with ASME XI Code and VERLIFE curves and crack behaviour is discussed.

Intergranular oxidation of 316L stainless steel in the PWR primary water environment

2017 ◽  
Vol 125 ◽  
pp. 175-183 ◽  
Author(s):  
R.P. Matthews ◽  
R.D. Knusten ◽  
J.E. Westraadt ◽  
T. Couvant
Keyword(s):  
Stainless Steel ◽  
316L Stainless Steel ◽  
Water Environment ◽  
Intergranular Oxidation ◽  
Primary Water

Continuity of Environmentally Assisted Fatigue and Stress Corrosion Cracking Based on Short Crack Growth Behavior of 316 Stainless Steel in Simulated PWR Primary Water

2017 ◽  
Author(s):  
Choongmoo Shim ◽  
Yoichi Takeda ◽  
Tetsuo Shoji
Keyword(s):  
Stainless Steel ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Water Environment ◽  
Growth Behavior ◽  
Corrosion Cracking ◽  
Short Crack ◽  
Crack Growth Behavior ◽  
Primary Water

Environmental correction factor (Fen) is one of the parameters to evaluate the effect of a pressurized high temperature water environment. It has been reported that Fen for stainless steel saturates at a very low strain rate. However, the relationship between environmentally assisted fatigue (EAF) and stress corrosion cracking (SCC) is still unclear. The aim of this study is to investigate the short crack growth behavior and possible continuity of EAF and SCC at very low strain rates. Short crack initiation and propagation have similar behaviors, which retard the crack growth between 100–200 μm in depth. We find that the striation spacing correlates well with the maximum crack growth rate (CGR) data. Based on the correlation, it is clarified that the local CGR on an intergranular facet was faster than that on a transgranular facet. Furthermore, the overall maximum and average CGR from the EAF data is well interpreted and compared with the SCC data.

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