• Journal of the Korean Society of Systems Engineering
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• v.10 no.1
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• pp.49-56
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• 2014

Since steam generator (SG) tubes are located in the boundary between the primary and secondary systems of nuclear power plant (NPP), the SG is one of the most important components in the aspects of the safety of NPP. The magnetite ($Fe_30_4$) deposition, so-called fouling, is generally known as a major aging mechanism of CANDU SGs, and this aging mechanism makes the heat transfer efficiency between the primary and secondary systems of NPP reduced. Therefore, the development of SG safety assessment system which can evaluate the effect of the SG aging degradation mechanism should be needed for safety of NPP. In this study, through the suggestion of the guideline for SG safety assessment, it is possible to strengthen the basic of establishing the effective SG aging management technique. The SG safety assessment is carried out by CATHENA(Canadian Algorithm for THErmalhydraulic Network Analysis). It is possible to determine the integrity of SGs by identifying the main safety parameters which can be changed by the aging degradation of CANDU SGs.

• 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.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.763-775
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• 2021

The thermal hydraulic performances of Steam Generator (SG) under both steady and transient operation conditions are of great importance for the safety and economy in nuclear power plants. In this paper, based on our self-developed SG thermal hydraulic analysis code STAF (Steam-generator Thermalhydraulic Analysis code based on Fluent), an improved new version STAF-CT (fully Coupling and Transient) is developed and introduced. Compared with original STAF, the new version code STAF-CT has two main functional improvements including "Transient" and "Fully Three Dimensional Coupling" features. In STAF-CT, a three dimensional energy transferring module is established which can achieve energy exchange computing function at the corresponding position between two sides of SG. The STAF-CT is validated against the international benchmark experiment data and the results show great agreement. Then the U-shaped SG in AP1000 nuclear power plant is modeled and simulated using STAF-CT. The results show that three dimensional flow fields in the primary side make significant effect on the energy source distribution between two sides. The development of code STAF-CT in this paper can provide an effective method for further SG high fidelity research in the nuclear reactor system.

• 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.

• 한국정보통신설비학회:학술대회논문집
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• 2009.08a
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• pp.54-58
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• 2009

In this paper, the ECT(eddy current testing) signal analysis of eddy current array probe for inspection of SG(steam generator) tube in NPP(nuclear power plant) using electromagnetic FEM(finite element method) was performed. To obtain the electromagnetic characteristics of probes, the governing equation was derived from Maxwell's equation, and the problem was solved by using the 3-dimensional FEM. The types of defects were FBH(flat bottomed hole) and OD groove, Spiral groove, natural defects(pitting, SCC, multiple SCC, wear). The depth of FBH defects were 20%, 40%, 60%, 80%, 100 of SG tube thickness, and it was assumed that the defects were located on the tube outside. And the operation frequency of 100kHz, 300kHz and 400kHz were used. Material of specimen was Inconel 600 which is usually used for SG tubes in NPP. The signal difference could be observed according to the variation of size and depth on FBH defects and operation frequencies. The results in this paper can be helpful when the ECT signals from EC array probe are evaluated and analyzed.

• 한국정보통신설비학회:학술대회논문집
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• 2006.08a
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• pp.241-245
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• 2006

In the examination of Steam Generator (SG) tube in Nuclear Power Plant (NPP) Eddy Current Testing (ECT) probes play an Important role in detecting the defects. Bobbin probe and Rotating Pancake Coil (RPC) probe is usually used for the inspection of SG tube. Bobbin probe is good at high speed inspection, but ability of detection of circumferential defect is very weak. On the contrary RPC probe, which moves for inspection in the direction of axial and circumferential simultaneously, has very slow inspection speed, but it was excellent detection capability fur small cracks, which is hardly detected by bobbin probe. Many examinations of SG tube examination of NPP are achieved during short period. Therefore, solution about this must develop probe of new form for examination performance and examination time shortening of other probe. In this paper, analyzed technological present condition of Bob-bin probe and RPC probe been using in Nondestructive Testing (NDT) for SG tube defect detection and Appeared about background theory and view of developed probe newly.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.4 no.2
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• pp.34-39
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• 2008

Steam Generator(SG) tubes are classified as one of the key components in nuclear power plants, and they should be periodically examined by the intensified management program for the assurance and diagnosis of their structural integrity. In this study, we use the optimum analysis technique to draw the detection and categorization of bowing(BOW) signals; abnormal tube-to-tube proximity in the SG upper bundle free span area. The locations in which BOW signals are detected likely have latent degradation of ODSCC(Outer Diameter Stress Corrosion Cracking). For the sake of timely and correct detection of BOW signals and diagnosis of ODSCC, we carried out the experimental demonstrations using a reduced mock-up. And we validated the MRPC(Motorized Rotating Pancake Coil) analysis technique is better than the bobbin. Hence, it comes to conclusion that the optimum analysis technique can be a good alternative for the reliable SG tube examination.

• 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.3 no.1
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• pp.30-33
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• 2004

It has been reported that some events of a rupture of seam generator tube have occurred in nuclear power plants around the world. Main causes of the leakage are from various types of corrosion in the steam generator(SG) tubings. Primary water stress corrosion cracking(PWSCC) of steam generator tubings have occurred in many tubes in Korean plant, and they were repaired using sleeves or plugs, In order to develop proper repair criteria, it is necessary to ascertain the leak behavior of the tubings. A high pressure leak and burst testing system was manufactured. Various types of Electro Discharged Machined (EDM) notches were developed on the SG tubes. Leak rate and burst pressure were measured on the tubes at room temperature. Burst pressure of the part through wall defected tubes depends on the defect depth, Water flow rates after the burst were independent of the t1aw types; tubes having 20 to 60 mm long EDM notches showed similar flow rates regardless of the defect depth. A fast pressurization rate gave the tube a lower burst pressure than the case of a slow pressurization.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1381-1389
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• 2019

Stress corrosion cracking (SCC) widely found in both primary and secondary sides of steam generator (SG) tubing in pressurized water reactors (PWR) has become an important safety issue. Using eddy-current tests (ECTs), non-destructive evaluations are performed for the integrity management of SG tubes against intergranular SCC. To enhance the reliability of ECT, this study investigates the effects of oxide films on ECT's detection capabilities for SCC in laboratory-degraded SG tubing in high temperature and high pressure aqueous environment.