• Proceedings of the KSME Conference
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• 2004.04a
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• pp.2047-2052
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• 2004

Inconel-600 alloy has been used as steam generator tube material for current pressurized water reactors (PWRs). The long-term operation of steam generators showed that the use of this material induced localized corrosion damages and increased tube wear of steam generator. To protect these problems, steam generator tube material is being changed to Inconel-690 alloy. Based on the current trend, we have chosen Inconel 690 as the Advanced Power Reactor 1400 (APR1400) steam generator(SG) tube material and performed the design optimization of preventive measure against tube fretting wear for the APR1400 steam generator. In this paper, we examined the technical consideration in this modification : the selection of material, wear characteristics, effect of the Egg-crate Flow Distribution Plate installation, and effect analysis of vertical strip installation.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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• pp.64-70
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• 2019

In recent years, attention has been paid to the integrity of steam generator (SG) tubes due to severe accident and beyond design basis accident conditions. In these transient conditions, steam generator tubes may be damaged by high temperature and pressure, which might result in a risk of fission products being released to the environment due to the failure. Alloy 690 which has increased the Cr content has been replaced for the SG tube due to its high corrosion resistance against stress corrosion cracking (SCC). However, there is lack of research on the high temperature creep rupture and life prediction model of Alloy 690. In this study, creep test was performed to estimate the high temperature creep rupture life of Alloy 690 using tube specimens. Based on manufacturer's creep data and creep test results performed in this study, creep life prediction was carried out using the Larson-Miller (LM) Parameter, Orr-Sherby-Dorn (OSD) parameter, Manson-Haford (MH) parameter, and Wilshire's approach. And a hyperbolic sine (sinh) function to determine master curves in LM, OSD and MH parameter methods was used for improving the creep life estimation of Alloy 690 material.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.20 no.1
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• pp.16-24
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• 2024

Denting of steam generator (SG) tube is defined as the reduction in tube diameter due to the stresses exerted by the corrosion products formed on the outer diameter surface. This phenomenon is mostly observed in the crevices between SG tube and the top-of tubesheet or tube support plate. Despite the replacement of SG tube with Alloy 690, which has better corrosion resistance than Alloy 600, the denting of SG tube still remains a potential problem that could decrease the SG integrity. Deformation of SG tube by denting phenomenon can affect the surface properties and microstructure of SG tube. In this study, the effects of plastic deformation on surface properties and microstructure of Alloy 690 thermally treated (TT) tube was investigated by using the various analysis techniques. The plastic deformation of Alloy 690 increased the surface roughness and area. Many surface defects such as ripped surface and micro-cracks were observed on the deformed Alloy 690TT specimen. Based on the electron backscatter diffraction analysis, the dislocation density of deformed SG tube increased compared to non-deformed SG tube. In addition, the effects of changes in surface properties and microstructure of SG tube on general corrosion behavior were discussed.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.49-55
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• 2020

Creep rupture data for Alloy 690 steam generator tubes in a pressurized water reactor are essentially needed to demonstrate a severe accident scenario on thermally-induced tube failures caused by hot gases in a damaged reactor core. The rupture data were obtained using the tube specimens under different applied-stress levels at 650℃, 700℃, 750℃, 800℃, and 850℃. Important creep constants were proposed using various creep laws in terms of Norton power law, Monkman-Grant (M-G) relation, damage tolerance factor (λ), and Zener-Hollomon parameter (Z). In addition, a creep activation energy (Q) value for Alloy 690 tube was reasonably determined using experimental data. Creep behaviors such as creep strength, creep rates, rupture elongation showed the results of temperature dependence well. Modified M-G plot improved a correlation of the creep rate and rupture life. Damage tolerance factor for Alloy 690 tubes was found to be λ =2.20 in an average value. Creep activation energy for Alloy 690 tube was optimized for Q=350 (kJ/mol). A plot of Z parameter obeyed a good linearity, and the same creep mechanism was inferred to be operative in the present test conditions.

• Corrosion Science and Technology
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• v.20 no.1
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• pp.22-25
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• 2021

In the past, Alloy 600 nickel-based alloys have been widely used in steam generators. However, most of them have been replaced by thermally treated alloy 690 tubes in recent years because mill annealed alloy 600 materials are known to be susceptible to stress corrosion cracking. Unlike this general perception, some steam generators using mill annealed alloy 600 tubes show excellent performance even though they are designed, manufactured, and operated in the same way. Therefore, various analyses were carried out to determine causes for the degradation of steam generators. Based on the general stress corrosion cracking mechanism, tube material susceptibility, residual stress, and sludge deposits of steam generators were compared to identify factors affecting stress corrosion cracking. It was found that mill annealed alloy 600 steam generator tubes showed higher resistance to stress corrosion cracking when the amount of sludge deposits on tube surface was smaller and residual stress generated during the fabrication was lower.

• Nuclear Engineering and Technology
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• v.30 no.6
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• pp.555-567
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• 1998

Alloy 600 and 690 steam generator tubes fabricated in Korea were evaluated by flare tests according to ASTM standards. The stress acting in the tube elements during the tests was predicted. All the tubes, including alleys 600 and 690, satisfied the requirement of a 30% or 35% O.D expansion. Flow curves obtained from the flare test were found to be higher in alloy 690 tubes than in alloy 600 ones. The difference between alloy 600 and 690 tubes increased gradually with flaring percentage (F.P,%). An effective stress corresponding to mean yield stress was introduced and calculated. It showed that the prediction values were in good agreement with the measured ones for all the 690 and 600 alloy tubes. It became possible to predict the amount of acting stresses within tubes during expansion process.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.917-925
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• 2022

Steam generator (SG) tubes in a nuclear power plant can undergo rapid changes in pressure and temperature during an accident; thus, an accurate model to predict short-term creep damage is essential. The theta (𝜃) projection method has been widely used for modeling creep-strain behavior under constant stress. However, many creep test data are obtained under constant load, so creep rupture behavior under a constant load cannot be accurately simulated due to the different stress conditions. This paper proposes a novel methodology to obtain the creep curve under constant stress using a modified 𝜃 projection method that considers the increase in true stress during creep deformation in a constant-load creep test. The methodology is validated using finite element analysis, and the limitations of the methodology are also discussed. The paper also proposes a creep-strain model for alloy 690 as an SG material and a novel creep hardening rule we call the damage-fraction hardening rule. The creep hardening rule is applied to evaluate the creep rupture behavior of SG tubes. The results of this study show its great potential to evaluate the rupture behavior of an SG tube governed by creep deformation.

• Corrosion Science and Technology
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• v.8 no.3
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• pp.126-131
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• 2009

The interaction between wear and corrosion can significantly increase total material losses in water chemistry environment. The corrosive wear tests of a PWR steam generator tube material (Alloy 690) against the anti vibration bar material (409 SS) were performed at room temperature. The tests were performed in alkaline water chemistry conditions. NaOH solution was selected for test condition to investigate the corrosive wear effect of steam generator tube material in alkaline pH condition without other factors. The flow induced vibration can caused tube damage and the corrosion can be occurred by water chemistry. The test results showed that, in the alkaline solution at pH 13.9, the corrosion current density was increased about ten times than that in the distilled water. And wear rate at pH 13.9 was increased about ten times from that at neutral condition. However, the wear rate was decreased with time. The decrease would be attributed to the change in roughness of specimen or sub-layer of the worn surface with time. From microstructure observation, severe abrasive shape and several wear debris were found. From those results, it could infer that the oxide film on Alloy 690 changed to easily breakable one in the alkaline water, and then abrasion with corrosion became the main wear mechanism.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1570-1578
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• 2022

During a severe accident, steam generator (SG) tubes undergo rapid changes in the pressure and temperature. Therefore, an appropriate creep model to predict a short term creep damage is essential. In this paper, a novel creep model for Alloy 690 SG tube material was proposed. It is based on the theta (θ) projection method that can represent all three stages of the creep process. The original θ projection method poses a limitation owing to its inability to represent experimental creep curves for SG tube materials for a large strain rate in the tertiary creep region. Therefore, a new modified θ projection method is proposed; subsequently, a master curve for Alloy 690 SG material is also proposed to optimize the creep model parameters, θ_i (i = 1-5). To adapt the implicit creep scheme to the finite element code, a partial derivative of incremental creep with respect to the stress is necessary. Accordingly, creep model parameters with a strictly linear relationship with the stress and temperature were proposed. The effectiveness of the model was validated using a commercial finite element analysis software. The creep model can be applied to evaluate the creep rupture behavior of SG tubes in nuclear power plants.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.5 no.2
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• pp.1-6
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• 2009

Nuclear power plants (NPP) using Alloy 600 (Ni 75wt%, Cr 15wt%, Fe 10wt%) as a heat exchanger tube of the steam generator (SG) have experienced various corrosion problems by ageing such as pitting, intergranular attack (IGA) and stress corrosion cracking (SCC). In spite of much effort to reduce the material degradations, SCC is still one of important problems to overcome. Especially lead is known to be one of the most deleterious species in the secondary system that cause SCC of the alloy. Even Alloy 690 (Ni 60wt%, Cr 30wt%, Fe 10wt%) as an alternative of Alloy 600 because of outstanding superiority to SCC is also susceptible to leaded environment. An oxide on SG tubing materials such as Alloy 600 and Alloy 690 is formed and modified expanding to complex sludge throughout hideout return (HOR) of various impurities including Pb. Oxide formation and breakdown is requisite for SCC initiation and propagation. Therefore it is expected that an oxide property such as a passivity of an oxide formed on steam generator tubing materials is deeply related to PbSCC and an inhibitor to hinder oxide modification by lead efficiently can be found. In the present work, the SCC susceptibility obtained by using a slow strain rate test (SSRT) in aqueous solutions with and without lead was discussed in view of the oxide property. The oxides formed on Alloy 600 and Alloy 690 in aqueous solutions with and without lead were examined by using a transmission electron microscopy (TEM), an energy dispersive x-ray spectroscopy (EDXS), an x-ray photoelectron spectroscopy (XPS) and an electrochemical impedance spectroscopy (EIS).