• Nuclear Engineering and Technology
• /
• v.40 no.7
• /
• pp.607-614
• /
• 2008

Since pipes with wall-thinning defects can collapse at fluid pressure that are lower than expected, the collapse moment of wall-thinned pipes should be determined accurately for the safety of nuclear power plants. Wall-thinning defects, which are mostly found in pipe bends and elbows, are mainly caused by flow-accelerated corrosion. This lowers the failure pressure, load-carrying capacity, deformation ability, and fatigue resistance of pipe bends and elbows. This paper offers a support vector regression (SVR) model further enhanced with a fuzzy algorithm for calculation of the collapse moment and for evaluating the integrity of wall-thinned piping systems. The fuzzy support vector regression (FSVR) model is applied to numerical data obtained from finite element analyses of piping systems with wall-thinning defects. In this paper, three FSVR models are developed, respectively, for three data sets divided into extrados, intrados, and crown defects corresponding to three different defect locations. It is known that FSVR models are sufficiently accurate for an integrity evaluation of piping systems from laser or ultrasonic measurements of wall-thinning defects.

• Nuclear Engineering and Technology
• /
• v.46 no.2
• /
• pp.225-234
• /
• 2014

Wall-thinned defects caused by accelerated corrosion due to fluid flow in the inner pipe appear in many structures of the secondary systems in nuclear power plants (NPPs) and are a major factor in degrading the integrity of pipes. Wall-thinned defects need to be managed not only when the NPP is under maintenance but also when the NPP is in normal operation. To this end, a test technique was developed in this study to detect such wall-thinned defects based on the temperature difference on the surface of a hot pipe using infrared (IR) thermography and a cooling device. Finite element analysis (FEA) was conducted to examine the tendency and experimental conditions for the cooling experiment. Based on the FEA results, the equipment was configured before the cooling experiment was conducted. The IR camera was then used to detect defects in the inner pipe of the pipe specimen that had artificially induced defects. The IR thermography developed in this study is expected to help resolve the issues related to the limitations of non-destructive inspection techniques that are currently conducted for NPP secondary systems and is expected to be very useful on the NPPs site.

• Nuclear Engineering and Technology
• /
• v.52 no.11
• /
• pp.2678-2685
• /
• 2020

The pipe bends and elbows in nuclear power plants (NPPs) are vulnerable to degradation mechanisms and can cause wall-thinning defects. As it is difficult to detect both the defects generated inside the wall-thinned pipes and the preliminary signs, the wall-thinning defects should be accurately estimated to maintain the integrity of NPPs. This paper proposes a deep fuzzy neural network (DFNN) method and estimates the collapse moment of wall-thinned pipe bends and elbows. The proposed model has a simplified structure in which the fuzzy neural network module is repeatedly connected, and it is optimized using the least squares method and genetic algorithm. Numerical data obtained through simulations on the pipe bends and elbows with extrados, intrados, and crown defects were applied to the DFNN model to estimate the collapse moment. The acquired databases were divided into training, optimization, and test datasets and used to train and verify the estimation model. Consequently, the relative root mean square (RMS) errors of the estimated collapse moment at all the defect locations were within 0.25% for the test data. Such a low RMS error indicates that the DFNN model is accurate in estimating the collapse moment for wall-thinned pipe bends and elbows.

• Nuclear Engineering and Technology
• /
• v.55 no.3
• /
• pp.947-957
• /
• 2023

Curved cylindrical structures such as elbows have a non-uniform thickness distribution due to their fabrication process, and as a result have a number of complex mode shapes, including circumferential and axial nodal patterns. In nuclear power plants, material degradation is induced in pipes by flow accelerated erosion and corrosion, causing the wall thickness of carbon steel elbows to gradually thin. The corresponding frequencies of each mode shape vary according to the wall thinning state. Therefore, the thinning state can be estimated by monitoring the varying modal characteristics of the elbow. This study investigated the varying modal characteristics of artificially thinned carbon steel elbows for each thinning state using numerical simulation and experimental methods (MRIT, Multiple Reference Impact Test). The natural frequencies of specified mode shapes were extracted, and results confirmed they linearly decreased with increasing thinning. In addition, by comparing single FRF (Frequency Response Function) data with the results of MRIT, a concise and cost effective thinning estimation method was suggested.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.36 no.3
• /
• pp.313-318
• /
• 2012

In CANDU, feeder pipes supply heavy water to pressure tube and steam generator. Under service conditions, Flow-Accelerated Corrosion (FAC) produces local wall-thinning in the feeder pipes. The wall-thinning in these pipes affects the integrity of the piping system, as verified in previous research. This paper provides limit loads for wallthinned feeder pipes with $45^{\circ}$ and $60^{\circ}$ bend angles, and proposes an equation that predicts the limit loads for wallthinned feeder pipes with arbitrary bend angles. On the basis of finite element limit analyses, limit loads are obtained for wall-thinned feeder pipes under in-plane bending and internal pressure. There are two cases of in-plane bending: the in-plane closing direction and the in-plane opening direction. The material is considered the effect of the large deformation, so an elastic-perfectly-plastic material is assumed in the calculations.

• Nuclear Engineering and Technology
• /
• v.52 no.9
• /
• pp.2092-2099
• /
• 2020

Many damages due to flow-accelerated corrosion and cracking have been observed during recent in-service inspections of nuclear power plants. To determine the operability or repair for damaged pipes, an integrity evaluation related to the damaged piping system should be performed by using already proven code and standards. One of them, the ASME Code Case is most popularly used to integrity assessment in nuclear power plants. However, the recent version of CC N-513 still recommends the simplified method which means a damaged elbow is assumed as an equivalent straight pipe. In addition, to enhance the accuracy integrity assessment in elbow, several previous studies recommend that the SIF and elastic COD values for an elbow with relatively large crack could be predicted by an interpolation technique. However, those estimates for elbow with relatively large crack might be derived to inaccurate results for crack growth analysis, such as for the allowable crack size and life estimation. Therefore, in this paper, the effect of crack length (0.3≤θ₁/π≤0.5) on SIF and elastic COD for elbow is systematically investigated. Then, for large crack in elbow, accurate estimates for SIF and elastic COD, which are widely used to assess the integrity of elbows, are proposed. Those proposed solutions are expected to be the technical basis for revisions of CC N-513-4 through the validation.

• Journal of the Korea Academia-Industrial cooperation Society
• /
• v.11 no.8
• /
• pp.3112-3118
• /
• 2010

The pH control agent in PWRs, to insure the integrity of steam generator, was changed from ammonia to ethanolamine(ETA) which decreased pH at condensate system and low pressure feedwater heater drain system, so that several amines were investigated for the selection of the optimum amine. There was no single alternative amine to meet the optimum condition. The more volatile ammonia provides the higher pH in condensate, while the less volatile ETA increases the pH in wet steam area. Thus, the combined amine of ammonia and ETA is able to equally raise the pH in both region so that the flow accelerated corrosion be reduced in the every system of the secondary side and the integrity of steam generator be also improved in pressurized water reactors (PWRs).

• Proceedings of the Korean Vacuum Society Conference
• /
• 1999.07a
• /
• pp.216-216
• /
• 1999

The usage of a stationary plasma thruster (SPT) ion source, invented previously for space application in Russia, in experiments with surface modifications and film deposition systems is reported here. Plasma in the SPT is formed and accelerated in electric discharge taking place in the crossed axial electric and radial magnetic fields. Brief description of the construction of specific model of SPT used in the experiments is presented. With gas flow rate 39ml/min, ion current distributions at several distances from the source are obtained. These was equal to 1~3 mA/$\textrm{cm}^2$ within an ion beam ejection angle of $\pm$20$^{\circ}$with discharge voltage 160V for Ar as a working gas. Such an extremely high ion current density allows us to obtain the Ti metal films with deposition rate of $\AA$/sec by sputtering of Ti target. It is shown a possibility of using of reactive gases in SPT (O2 and N2) along with high purity inert gases used for cathode to prevent the latter contamination. It is shown the SPT can be operated at the discharge and accelerating boltages up to 600V. The results of presented experiments show high promises of the SPT in sputtering and surface modification systems for deposition of oxide thin films on Si or polymer substrates for semiconductor devices, optical coatings and metal corrosion barrier layers. Also, we have been tried to establish in application of the modeling expertise gained in electric and ionic propulsion to permit numerical simulation of additional processing systems. In this mechanism, it will be compared with conventional DC sputtering for film microstructure, chemical composition and crystallographic considerations.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.34 no.9
• /
• pp.1293-1298
• /
• 2010

In most power plants, wall thinning in carbon-steel pipes due to flow-accelerated corrosion is one of the major aging phenomena, and it reduces the load-carrying capacity of the piping system. Various types of wall-thinning defects were manufactured in real-scale elbows, and monotonic in-plane bending tests were performed under internal pressure to evaluate the failure behavior of the elbows. In this paper, the behavior of elastic and plastic limit leads of locally thinned elbows in a real-scale failure test is presented. The loads determined on the basis of TES (twice elastic slope) were considered to be the limit loads of locally thinned elbows so that the integrity of the thinned elbows could be maintained, even when a small amount of plastic deformation might have occurred.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.35 no.10
• /
• pp.1097-1103
• /
• 2011

Liquid droplet impingement erosion (LDIE) is caused by the impact of high-velocity droplets entrained in steam or air on metal. The degradation caused by the LDIE has been experienced in steam turbine internals and high-velocity airplane components (particularly canopies). Recently, LDIE has also been observed in the pipelines of nuclear plants. LDIE among the pipelines occurs when two-phase steam experiences a high pressure drop (e.g., across an orifice in a line to the condenser). In 2011, a nuclear power plant in Korea experienced a steam leak caused by LDIE in a pipe through which a two-phase fluid was flowing. This paper describes a study on the design change of a pipe affected by LDIE in order to mitigate the damage. The design change has been reviewed in terms of fluid dynamics by using the FLUENT code.