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Exposure and slow strain rate tensile tests were conducted in a simulated pressurized water reactor (PWR) primary water to investigate the oxidation resistance and stress corrosion cracking (SCC) susceptibility of 308L and 309L stainless steel (SS) cladding layers. A double-layer structure oxide layer grown on 308L SS and 309L SS contained the Cr-enriched nanocrystalline internal layer and the Fe-enriched spinel oxide in the external layer. Ni-enrichment at the matrix/oxide boundary was observed. The internal oxide film on 309L SS was thicker and had a lower Cr content than that on 308L SS. Preferential dissolution of inclusions led to pits on 308L SS and 309L SS surfaces during the exposure tests. More inclusions in 309L would decrease its SCC resistance due to the pits’ ability to act as the SCC initiation site. 308L SS had a lower susceptibility to SCC than 309L SS in PWR primary water. Lower ferrite content and higher strength/hardness reduced the oxidation and SCC resistance of 309L SS cladding. The effect of ferrite on oxidation and SCC of the SS claddings is discussed.

Double loop electrochemical potentiokinetic reactivation (DL-EPR) was applied to evaluate thermal aging of 308L stainless steel weld metal. It was found that the activation and reactivation peaks of DL-EPR curve were induced by dissolution of austenite and δ-ferrite, respectively. Before saturation of hardness, the linear relationship between reactivation ratio and hardness could be used for assessing the thermal aging-induced hardening. In the following thermal aging process, the reactivation ratio is applicable to assess the occurrence of the saturation of thermal aging-induced hardening. The results demonstrated that the DL-EPR test is applicable to assess the evolution of thermal aging.

Although highly cold-worked UNS N06690 has been shown to be susceptible to stress corrosion crack growth in pressurized water reactor primary water, it is unclear whether cold-work promotes stress ...

The stress corrosion crack (SCC) initiation and propagation behavior of three heats of Alloy 600 in simulated pressurized water reactor (PWR) primary water have been studied. The SCC initiation time could be effectively detected using constant load tests with in situ direct current potential drop (DCPD) method. Materials with lower SCC initiation time also showed higher SCC crack growth rate (CGR) under constant K condition. The relationship between SCC initiation time and CGR of Alloy 600 was qualitatively built, and the SCC initiation time could be roughly predicted once the CGR was measured.

The electrochemical corrosion behavior of Cr, Ni, and Mo in Fe-Cr-Ni alloys in simulated pressurized water reactor water was investigated by thermodynamic calculation, potentiodynamic polarization, and electrochemical impedance spectra analysis. The results revealed that the corrosion potential of Fe-Cr-Ni alloys in Ar-deaerated water was between the and , affected most by the Fe anodic dissolution and water reduction. The oxidation current for Fe3O4 to Fe2O3 can be only detected when the potential crosses the Fe3O4/Fe2O3 phase boundary during potentiodynamic polarization. For both stainless steels and nickel-based alloys, rapid chromium oxide dissolution may be an important reason for the short stress corrosion cracking (SCC) initiation times and higher SCC susceptibility measured in oxygenated pure water. Mo and Cr additions in Fe-Cr-Ni alloys reduce the general corrosion rate of stainless steels in the passive region, but no effect is observed in the transpassive region where chromium oxide is dissolving. Mo may stabilize the inner compact Cr/Ni-rich layer oxides of Mo-containing stainless steels.

Forged Type 316NG austenitic stainless steel is commonly used to fabricate primary coolant pipe in advanced pressurized water reactor systems, but it is subject to corrosion fatigue cracking in primary coolant environments. In this study, the fatigue crack growth of Type 316NG austenitic stainless steel in 325°C water was investigated. The corrosion fatigue effect in 325°C water was mainly correlated to the crack growth rate and load frequency. A maximum Fen (environment corrected factor) value up to 100 was observed at f = 0.005 Hz, ΔK = 13 MPa√m. The dissolved oxygen tended to have little influence on the fatigue crack growth. The crack path was deflected and branched. Most secondary cracks tended to be initiated from grain boundaries.

The primary purpose of this research is to examine the stress corrosion cracking (SCC) resistance of Alloy 800NG in pressurized water reactor (PWR) primary water and pressurized heavy water reactor (PHWR) primary water. Rates of SCC growth of 20% cold-worked (CW) Alloy 800NG measured over the temperature range between 270°C and 360°C were compared with previously reported results for 20% CW Alloy TT690 and 20% CW Alloy 600 in order to consider which material is the most SCC resistant among materials presently being used for steam generator (SG) tubing worldwide. The secondary purpose is to examine the effect of chromium addition on SCC growth in PWR primary water of a series of alloys based on the Alloy 800 composition. SCC growth measurements were performed in PWR primary water over the chromium concentration range from 16% to 27% to obtain fundamental knowledge useful for considering a future alternative SCC-resistant material for SG tubing in extended life PWRs and PHWRs. The third objective is to examine the rate of cavity formation of 20% CW Alloy 800NG to obtain basic knowledge of one possible mechanism for SCC initiation after long-term operation. Measured rates of cavity formation in 20% CW Alloy 800NG were compared with previously reported results of 20% CW Alloy TT690 to compare the rate of SCC initiation caused by cavity formation. Four important patterns were observed. First, excellent SCC growth resistance was observed for 20% CW Alloy 800NG compared to 20% CW Alloy TT690 at 320°C, 340°C, and 360°C. Second, an inverse temperature dependence on SCC growth was observed in Alloy 800NG. The rate of SCC growth increased with decreasing temperature which was completely different from the trend for Alloy 600. Third, a significant beneficial effect by chromium addition in 800 series alloys on SCC growth resistance was observed in PWR primary water in the operating temperature range of PWRs and PHWRs. The rate of SCC growth decreased with increasing chromium concentration in the chromium concentration range between 16% and 27% chromium at 270°C, 290°C, and 320°C. However, no beneficial effect of chromium addition in these alloys was observed at 340°C and 360°C. Finally, a more than 10 times slower rate of cavity formation was observed in 20% CW Alloy 800NG than for 20% CW Alloy TT690. Results suggested that because of cavity formation, a more than 10-fold faster crack initiation occurred in Alloy TT690 than in Alloy 800NG. Further, carbide coverage and grain size significantly affected the rate of cavity formation. Detailed and comprehensive studies of long-term SCC initiation are necessary to ensure the future reliability of life-extended PWRs and PHWRs.

The fatigue crack propagation behavior on the surface of machining notches of Alloy 690TT (thermally treated) was investigated in room-temperature air environment and in simulated pressurized water reactor (PWR) environment with saturated air at 325°C. The results showed that more than one crack formed on the notch surface and the crack propagation paths were transgranular with some branching. The grain size also had some effect on the crack propagation path but had no significant effects on the crack growth rates (CGRs). The CGR vs. a curve could be divided into two stages in air environment and three stages in simulated PWR environment. The corrosive environment promotes crack propagation in the macroscopic crack propagation stage. Different load frequencies express similar crack propagation process, but decreasing frequency can assist the corrosion fatigue crack nucleation and propagation on the notch surface.

Stress corrosion crack initiation of two thermally-treated, cold-worked (CW) Alloy 690 (UNS N06690) materials was investigated in 360°C simulated pressurized water reactor primary water using constant load tensile (CLT) and blunt notch compact tension (BNCT) tests equipped with direct current potential drop (DCPD) for in situ detection of cracking. Stress corrosion cracking initiation was not detected by DCPD for either the 21% or 31% CW CLT specimens loaded at their yield stress for ∼9,220 h; however, intergranular (IG) precursor damage and isolated surface cracks were observed on the specimens. The two 31% CW BNCT specimens loaded at constant stress intensity after several cyclic loading ramps showed DCPD-indicated crack initiation after 10,400 h of exposure, which resulted from significant growth of IG cracks. The 21% CW BNCT specimens only exhibited isolated small IG surface cracks and showed no apparent DCPD change throughout the test. Post-test cross-section examinations revealed many grain boundary...

Environmentally assisted cracking (EAC) susceptibilities of Alloys 600, 690, and their weld metals with and without weld defects were benchmarked in pure and doped hydrogenated steam at 400°C and pressurized water reactor (PWR) water at 360°C on four-point bend (0.35% to 1.1% strain) and U-bend specimens (5% strain). Contrary to the expectations, no EAC initiation from existing weld defects was observed. As expected, Alloy 600 and its weld metals were much more prone to EAC initiation than Alloy 690 and its weld metals. Oxide structures remain the same in doped and pure steam and PWR water, although the environment has an impact on the EAC initiation times and oxidation rates. Because of the similar oxide structures, the EAC mechanism is assumed to be the same in all studied environments.

The effects of dissolved gas and cold work on the electrochemical behaviors of Type 304 stainless steel were investigated in simulated pressurized water reactor primary water environment by potentiodynamic polarization and electrochemical impedance spectroscopy. Results reveal that both dissolved gas and cold work have a significant influence on anodic passivation properties and corrosion resistance of the oxide film. In Ar deaerated water and oxygenated water environments, the corrosion potential of stainless steel and Pt is controlled by the dissolved oxygen concentration, and the corrosion resistance of the inner barrier layer is much higher. However, in H2 deaerated water, the corrosion potential is depressed by the H2/H2O redox potential and the inner oxide film formed in H2 deaerated water is less stable. In addition, the passive region of polarization curves on 304 stainless steel in oxygenated and Ar deaerated water is controlled by metal dissolution instead of water oxidation, and the passive cur...

In order to provide background for a review of stress corrosion cracking (SCC) in nuclear power plants, a review is first presented of the history of occurrence of SCC in fossil fired power plants and a few related applications. The historical development of SCC in nuclear power plants is then described. The many materials and components that have been affected by SCC over the years are identified, as are the material, stress, and environmental factors involved. A particular focus of this monograph is the question of how it happened that materials with significant susceptibility to SCC were used for so many important structural applications in nuclear power plants. To address this question, the historical development of both pressurized water reactor designs and boiling water reactor designs is reviewed. The review covers factors such as the material types selected for test and use, the operating temperatures used as the designs evolved, and the qualification tests and prior operating experience that were relied upon for selection of materials. Some conclusions as to the factors involved in the choice of materials that turned out to be susceptible to SCC are presented, together with some suggestions for how to minimize such problems in the future.

Synchrotron x-ray diffraction analytical techniques have been used to investigate the structure of oxide films formed on Type 316L (UNS S31603) austenitic stainless steel in simulated pressurized water reactor primary water. An in situ technique for investigating the layer structures of oxide films has been developed using this measurement method. The observed layer structures of the oxide films changed depending on the dissolved hydrogen concentration (DH) in PWR primary water. In two cases, where DH = 5 cm3/kg (H2O) or 30 cm3/kg (H2O), a (NixFe(1−x))Fe2O4-type spinel oxide was observed as the outer oxide, and a FeCr2O4-type spinel oxide was detected as the thin inner oxide. When DH = 30 cm3/kg (H2O), the Fe:Ni ratio in the (NixFe(1−x))Fe2O4 outer spinel oxide was much larger than when DH = 5 cm3/kg (H2O). In addition, sequential in situ measurements when the hydrogen concentration varied from 5 cm3/kg (H2O) to 30 cm3/kg (H2O) also demonstrated that the oxide layer structure seemed to adjust its characte...

The primary water stress corrosion cracking (PWSCC) susceptibility of Ni-based alloys, such as Alloy 600 (UNS N06600), has been much higher than that of Fe-based alloys, such as Type 316 stainless steel (UNS S31600). This phenomenon is generally recognized; however, the effects of Fe, Ni, and Cr are not fully understood. In this study, the effect of Cr and Ni in Ni-Cr-Fe alloys on PWSCC susceptibility was investigated using alloys with a wide range of Ni content. The role of these elements is discussed from the standpoint of grain boundary oxidation. According to the stress corrosion cracking (SCC) test results, high-Ni alloys with low Cr contents had high SCC susceptibility, and the SCC susceptibility in high Ni-content alloys was improved by increasing Cr content. On the other hand, low Ni-content alloys had low SCC susceptibility, irrespective of Cr content. These features agree with the field experience in pressurized water reactor plants. In comparison with the results of SCC tests and immersion test...

Two unique facilities for isothermal low-cycle fatigue (LCF) and in-phase (IP) or out-of-phase (OP) thermo-mechanical fatigue (TMF) testing of tubular specimens under boiling water reactor (BWR) and pressurized water reactor (PWR) coolant conditions were set up and successfully tested. The systems allow both strain- and stress-controlled fatigue experiments with very small strain amplitudes or complex stress/strain profiles with superimposed rather rapid temperature changes (100°C to 340°C) under flowing conditions. The present article introduces the new test facility and briefly discusses the first results of these experiments. In contrast to air, a strong effect of temperature on LCF and TMF lives was observed in deoxygenated and hydrogenated high-temperature water environment, where both lives decrease with increasing temperature above 100°C. The TMF life is between that of the isothermal LCF tests at minimum and maximum temperature. The IP TMF life is shorter than that of OP TMF and close to the LCF l...

Grain boundary microstructures and microchemistries are examined in cold-rolled Alloy 690 (UNS N06690) materials and comparisons are made to intergranular stress corrosion cracking (IGSCC) behavior in pressurized water reactor (PWR) primary water. Chromium carbide precipitation is found to be a key aspect for materials in both the mill-annealed and thermally treated conditions. Cold rolling to high levels of reduction was discovered to produce small IG voids and cracked carbides in alloys with a high density of grain boundary carbides. The degree of permanent grain boundary damage from cold rolling was found to depend directly on the initial IG carbide distribution. For the same degree of cold rolling, alloys with few IG precipitates exhibited much less permanent damage. Although this difference in grain boundary damage appears to correlate with measured SCC growth rates, crack tip examinations reveal that cracked carbides appeared to blunt propagation of IGSCC cracks in many cases. Preliminary results su...

This paper describes the development of both facilities and capabilities for conducting crack growth rate tests on neutron-irradiated alloys in light water reactor (LWR) environments. The facilitie...

The rate of stress corrosion cracking (SCC) was measured for nonsensitized, cold-worked Type 316 (UNS S31600) and Type 304 (UNS S30400) in both hydrogenated pressurized water reactor (PWR)...

Pitting corrosion was the primary cause of the Alloy 600 (UNS N06600) steam generator tube degradation in a Korean pressurized water reactor (PWR) plant. Pulled tube examinations and remedial measures were carried out to mitigate the pitting. Based on the destructive examinations, the main causes of pitting corrosion were considered to be the following: accumulated sludge with a high copper content due to corrosion of copper alloys in the secondary system, acidic crevice conditions caused by chloride from condenser leakage, and ingress of air during layup. Countermeasures such as copper alloy replacement, water chemistry control, and chemical cleaning were implemented to mitigate the pitting. Chemical cleaning was evaluated as the most effective.

The work in this article is part of a project to develop a quantitative description of stress corrosion cracking (SCC) on the secondary side of pressurized water reactor (PWR) steam generator tubing based on existing information from operating plants and from laboratory experiments. This work is the second step in developing a predictive model for SCC on the secondary side. The first step involved developing a statistical framework into which dependencies of the various submodes of SCC can be inserted. The results of the present work will lead to quantitative descriptions of corrosion processes that, in turn, will be incorporated into the statistical framework. The chemistry of heat-transfer crevices will then be assessed to determine the proper inputs to the dependencies of the various submodes of SCC, and these will be connected to bulk environments. The modeling here is directed toward predicting the early occurrence of SCC that is too shallow to be detected by nondestructive examination (NDE)...

The work in this article is part of a project to develop a quantitative description of stress corrosion cracking (SCC) on the secondary side of pressurized water reactor (PWR) steam genera...

The work in this article is part of a project to develop a quantitative description of stress corrosion cracking (SCC) on the secondary side of pressurized water reactor (PWR) steam generator tubing based on existing information from operating plants and from laboratory experiments. This work is the second step in developing a predictive model for SCC on the secondary side. The first step involved developing a statistical framework into which dependencies of the various submodes of SCC can be inserted. The results of the present work will lead to quantitative descriptions of corrosion processes that, in turn, will be incorporated into the statistical framework. The chemistry of heat-transfer crevices will then be assessed to determine the proper inputs to the dependencies of the various submodes of SCC, and these will be connected to bulk environments. The modeling here is directed toward predicting the early occurrence of SCC that is too shallow to be detected by nondestructive examination (NDE)...

The electrochemical corrosion potential (ECP) of a metal is influenced by a large number of variables, such as oxygen, hydrogen, and hydrogen peroxide (H2O2) concentrations, the surface chemistry, impurities, hydrodynamic flow conditions, type of reference electrode, etc. The state of reference electrode development and qualification, water chemistry conditioning procedures, surface history, etc. is such that achieving total agreement from all international investigators is difficult. However, identification of key factors, preferred techniques, and sensible ranges in potentials for several standard conditions is possible. The purpose of this paper is to identify factors that influence ECP measurements in high-temperature water, propose recommended techniques, and characterize the expected range of corrosion potentials for specific materials, dissolved gas chemistries, water chemistry/pH, and temperature associated with boiling water reactor (BWR), pressurized water reactor (PWR), and related hig...

The aging of water-cooled and moderated nuclear steam supply systems inevitably has given rise to many material corrosion problems of which stress corrosion cracking (SCC) is one of the mo...

The effect of zinc addition to pressurized water reactor (PWR) primary water on primary water stress corrosion cracking (PWSCC) susceptibility of mill-annealed Alloy 600 (UNS N06600, 600MA...

Intergranular (IG) attack regions and stress corrosion cracks in Alloy 600 (UNS N06600) U-bend samples tested in 330°C pressurized water reactor water were characterized by analytical tran...

Electrochemical potential noise measurements on sensitized stainless steel pressure tubes under pressurized water reactor (PWR) conditions were performed for the first time. Very short potential spikes, believed to be associated to crack initiation events, were detected when stressing the sample above the yield strength and increased in magnitude until the sample broke. Sudden increases of plastic deformation, as induced by an increased tube pressure, resulted in slower, high-amplitude potential transients, often accompanied by a reduction in noise level.

Outer-diameter stress corrosion cracking (ODSCC) of alloy 600 (UNS N06600) tubings in steam generators of the Kori-1 pressurized-water reactor (PWR) caused an unscheduled outage in 1994. Failure analysis and remedy development studies were undertaken to avoid a recurrence. Destructive examination of a removed tube indicated axial intergranular cracks developed at the top of sludge caused by a boiling crevice geometry. A high ODSCC propagation rate was attributed to high local pH and increased corrosion potential resulting from oxidized copper presumably formed during the maintenance outage and plant heatup. Remedial measures included: (1) crevice neutralization by crevice flushing with boric acid (H3BO3) and molar ratio control using ammonium chloride (NH4Cl), (2) corrosion potential reduction by hydrazine (H2NNH2) soaking and suppression of oxygen below 20 ppb to avoid copper oxide formation, (3) titanium dioxide (TiO2) inhibitor soaking, and (4) temperature reduction of 5°C. Since application o...

Stress corrosion cracking (SCC) initiation tests were carried out on alloy 600 (UNS N06600) in primary water at 325°C and 360°C and in steam with hydrogen at 400°C to gain some insight int...

A study was undertaken to better understand the lead-induced corrosion mechanism of nickel-based alloys used for steam generator tubing materials (alloys 600 and 690 [UNS N06600 and N06690...

The trend in pressurized water reactors (PWR) toward higher burnups, increasing lithium concentrations, and higher coolant temperatures imposes a demand for better fuel cladding corrosion and hydriding properties. There is a lack of reliable and fast in-situ techniques to investigate zirconium alloys in high-temperature water environments. The contact electric resistance (CER) technique was used to measure the electric resistance of the oxide growing on a zirconium-based fuel cladding material. Lithium hydroxide (LiOH) decreased electric resistance of the oxide when LiOH was in excess of ∼ 70 ppm in PWR water at 300°C. Electric resistance of the oxide was dependent upon LiOH concentration and was shown to correlate inversely with the effect of LiOH on weight gain. Kinetics of the decrease of electric resistance indicated the mechanism of degradation was a phase transformation rather than a diffusion-limited process.

The susceptibility of alloy 600 (UNS N06600) to stress corrosion cracking (SCC) in hydrogenated high-temperature water was studied as a function of the degree of cold work (CW) of the allo...

The steam generator in a pressurized water reactor (PWR) of a nuclear power plant consists mainly of a shell made of carbon (C) steel and tubes made of alloy 600 (UNS N06600). However, all...

Since lead may be one of the potential contributing causes of intergranular corrosion/stress corrosion cracking, which is found in alloy 600 (UNS N06600) steam generator tubing in PWR seco...

Environmentally assisted cracking (EAC) of different heat treatments of alloy X-750 has been studied in simulated pressurized water reactor (PWR) primary water from 50 to 340°C. The alloy, with single- (704°C/20 h) and two-stage (885°C/24 h + 704°C/20 h) aging treatments, was increasingly susceptible to EAC with decreasing temperature below ∼250°C. This susceptibility was due to hydrogen embrittlement, as indicated by the sensitivity to the hydrogen content of the water and behavior in hydrogen gas. At higher testing temperatures, a second mechanism, possibly anodic dissolution, was dominant for alloys with either a lower solution heat treatment temperature (980°C/1 h) or two-stage aging treatment. This second mechanism was suppressed by a higher temperature solution treatment (∼1100°C/1 h) and a single-stage aging treatment, which improved alloy performance.

The frequently reported effect of the partial pressure of hydrogen on the alloy 600 susceptibility to intergranular stress corrosion cracking (IGSCC) in high-temperature water is explained in this paper. On the basis of thermodynamic calculations, it was demonstrated that the susceptibility of alloy 600 to IGSCC is connected with the stability of NiO in this environment. The lower the free energy of formation of NiO, the higher the susceptibility of alloy 600 to IGSCC.

A basic problem of austenitic stainless steels applied in pressurized water reactors (PWR) is degradation by corrosion fatigue. This paper proposes a prediction model (Nuclear Power Institute of China [NPIC] model) to estimate corrosion fatigue lives of austenitic stainless steels in PWR environments. The NPIC model is built on a corrosion fatigue mechanism and indicates the corrosion fatigue processes better than the ANL model proposed by Argonne National Laboratory. Both the NPIC and ANL models are somewhat conservative in their predictions.