• 대한전자공학회:학술대회논문집
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• 대한전자공학회 2002년도 하계종합학술대회 논문집(4)
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• pp.263-266
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• 2002

This paper describes extraction procedure for the center coordinates of steam generator tubes of Youngkwang NPP #6, which are arrayed in triangular patterns. Steam generator tube images taken with wide field-of-view lens and low-light lamp mounted on a ccd camera tend to have low contrast, because steam generator is sealed and poorly illuminated. The extraction procedures consists of two steps. The first step is to process the region with superior contrast in entire image of steam generator tubes and to extract the center points. Using the extracted coordinates in the first step and the geometrical array characteristics of tubes lined up in regular triangle forms, the central points of the rest region with low contrast are estimated. The straight lines from center point of a tube to neighbour points in horizontal and 60, 120$^{\circ}$ degree directions are derived. The intersections of straight line In horizontal direction and slant line in regular triangle direction are selected as the center coordinates of steam generator tubes. The Chi-square interpolation method is used to determine the line's coefficients in horizontal and regular triangle direction.

• Nuclear Engineering and Technology
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• 제41권3호
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• pp.287-294
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• 2009

Due to thermal hydraulics phenomena, such as thermal stratification, various events occur to the parts of a nuclear power plant during their lifetimes: e.g. cracked and dislocated pipes and thermally fatigued, bent, and damaged supports. Due to the operational characteristics of the parts of the steam generator feedwater inlet horizontal pipe, thermal stratification takes place particularly frequently. However, the thermal stress due to thermal stratification at the steam generator feedwater inlet horizontal pipe was not reflected in the design stage of old plants(Kori Unit No.1, 2, 3 and 4, Yeonggwang Unit No. 1 and 2, and Uljin Unit No. 1 and 2; referred to as old-style power plants hereinafter). Accordingly, a verification experiment was performed for thermal stratification in the horizontal inlet nozzle steam generator of old-style plants. If thermal stratification occurred in the horizontal pipe of an old-style power plant, numerical analysis of the temperature distribution of the pipes and fluids was conducted. The temperature distributions were compared at the curved part of the pipe and the horizontal pipe before and after the installation of the improved thermal sleeves designed to alleviate thermal stress due to thermal stratification. The thermal stress reduction measure was proven effective at the steam generator inlet horizontal pipe and the curved part of the pipe.

• Nuclear Engineering and Technology
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• 제42권1호
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• pp.97-108
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• 2010

Fluid-elastic instability and turbulence-induced vibration of steam generator U-tubes of a nuclear power plant are studied numerically to investigate the effect of design changes of support structures in the upper region of the tubes. Two steam generator models, Model A and Model B, are considered in this study. The main design features of both models are identical except for the conditions of vertical and horizontal support bars. The location and number of vertical and horizontal support bars at the middle of the U-bend region in Model A differs from that of Model B. The stability ratio and the amplitude of turbulence-induced vibration are calculated by a computer program based on the ASME code. The mode shape with a large modal displacement at the upper region of the U-tube is the key parameter related to the fretting wear between the tube and its support structures, such as vertical, horizontal, and diagonal support bars. Therefore, the location and the number of vertical and horizontal support bars have a great influence on the fretting wear mechanism. The variation in the stability ratios for each vibrational mode is compared with respect to Model A and Model B. Even though both models satisfy the design criteria, Model A shows substantial improvements over Model B, particularly in terms of having greater amplitude margins in the turbulence-excited vibration (especially at the inner region of the tube bundle) and better stability ratios for the fluid-elastic instability.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집B
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• pp.645-650
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• 2000

This dynamic analysis is performed about shutdown, load controlled and temperature controlled startup operating characteristics of the Horizontal drum type HRSG. This analysis was performed by constructing a dynamic model of the plant and running it through the appropriate.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2005년도 춘계학술대회논문집
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• pp.312-317
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• 2005

Fluid-elastic instability and turbulence excitation for an under developing steam generator are investigated numerically. The stability ratio and the amplitude of turbulence excitation are obtained by using the PIAT (Program for Integrity Assessment of Steam Generator Tube) code from the information on the thermal-hydraulic data of the steam generator. The aspect ratio, the ratio between the height of U-tube from the upper most tube support plate (h) and the width of two vertical portion of U-tube (w), is defined for geometric parameter study. Several aspect ratios with relocation of tube support plates are adopted to study the effects on the mode shapes and characteristics of flow-induced vibration. When the aspect ratio exceeds value of 1, most of the mode shapes at low frequency are generated at the top of U-tube. It makes very high value of the stability ratio and the amplitude of turbulent excitation as well. We can consider that the local mode shape at the upper side of U-tube will develop the wear phenomena between the tube and the anti-vibration bars such as vertical, horizontal, and diagonal strips. It turns out that the aspect ratio reveals very important parameter for the design stage of the steam generator. The appropriate value of the aspect ratio should be specified and applied.

• 한국소음진동공학회논문집
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• 제15권9호
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• pp.1009-1015
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• 2005

Fluid-elastic instability and turbulence excitation for an under developing steam generator are investigated numerically. The stability ratio and the amplitude of turbulence excitation are obtained by using the $PIAT^{(R)}$ (program for integrity assessment of steam generator tube) code from the information on the thermal-hydraulic data of the steam generator. The aspect ratio, the ratio between the height of U-tube from the upper most tube support Plate (h) and the width of two vertical portion of U-tube (w), is defined for geometric parameter study. Several aspect ratios with relocation of tube support plates are adopted to study the effects on the mode shapes and characteristics of flow-induced vibration. When the aspect ratio exceeds value of 1, most of the mode shapes at low frequency are generated at the top of U-tube. It makes very high value of the stability ratio and the amplitude of turbulent excitation as well. We can consider that the local mode shape at the upper side of U-tube will develop the wear phenomena between the tube and the anti-nitration bars such as vortical, horizontal, and diagonal strips. It turns out that the aspect ratio reveals very important parameter for the design stage of the steam generator. The appropriate value of the aspect ratio should be specified and applied.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.92-102
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• 2017

The analysis of steam generators as an interface between primary and secondary circuits in light water nuclear power plants is crucial in terms of safety and design issues. VVER-1000 nuclear power plants use horizontal steam generators which demand a detailed thermal hydraulics investigation in order to predict their behavior during normal and transient operational conditions. Two phase flow field simulation on adjacent tube bundles is important in obtaining logical numerical results. However, the complexity of the tube bundles, due to geometry and arrangement, makes it complicated. Employment of porous media is suggested to simplify numerical modeling. This study presents the use of porous media to simulate the tube bundles within a general-purpose computational fluid dynamics code. Solved governing equations are generalized phase continuity, momentum, and energy equations. Boundary conditions, as one of the main challenges in this numerical analysis, are optimized. The model has been verified and tuned by simple two-dimensional geometry. It is shown that the obtained vapor volume fraction near the cold and hot collectors predict the experimental results more accurately than in previous studies.

• 한국압력기기공학회 논문집
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• 제7권1호
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• pp.35-40
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• 2011

If a plugged tube in a steam generator is broken, it may damage nearby intact tubes. To prevent this damage, it is recommended that a stabilizer is installed into the plugged tube. However, the installation cost of a stabilizer is very high. So studies are required to determine the conditions on which the installation is necessary. For this purpose static loads and dynamic loads applied on a tube should be known to estimate the residual strength and remaining fatigue and wear life of a plugged tube. Two-dimensional and three-dimensional computational fluid dynamics (CFD) analyses are performed to obtain the drag coefficient for cross flow to a tube. Using the obtained drag coefficient, the static load can be estimated and the residual strength of a plugged tube can be calculated. An inclined flow problem is also analyzed and the vertical and horizontal forces are obtained and discussed.

• 대한기계학회논문집B
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• 제22권9호
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• pp.1267-1276
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• 1998

Falling-film evaporation experiments with aqueous lithium bromide (LiBr) solution were performed to investigate the heat transfer characteristics of enhanced copper tubes. Enhanced tubes (a knurled tube, a spirally grooved tube, and a tube coated with $20{\mu}m$ aluminum particles) and a bare tube were selected as test specimens. Averaged evaporation fluxes of water were obtained from horizontal tubes with various film Reynolds numbers, system pressures, LiBr concentrations and degrees of wall superheat. The enhanced performance of steam generation was compared between tubes with varying parameters. The knurled tube geometry showed the most excellent performance among the tubes tested. The specified enhanced tubes were more useful for generating steam on a low grade heat source such as waste heat.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4561-4569
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• 2023

A correct understanding of vibration-based degradation is crucial from the standpoint of maintenance for Steam Generators (SG) as crucial mechanical equipment in nuclear power plants. This study has established a novel approach to developing a model for investigating tube bundle degradation according to crack growth caused by two-phase Flow-Induced Vibration (FIV). An important step in the approach is to calculate the two-phase flow field parameters between the SG tube bundles in various zones using the porous media model to determine the velocity and vapor volume fraction. Afterward, to determine the vibration properties of the tube bundles, the Fluid-Solid Interaction (FSI) analysis is performed in eighteen thermal-hydraulic zones. Tube bundle degradation based on crack growth using the sixteen most probable initial cracks and within each SG thermal-hydraulic zone is performed to calculate useful lifetime. Large Eddy Simulation (LES) model, Paris law, and Wiener process model are considered to model the turbulent crossflow around the tube bundles, simulation of elliptical crack growth due to the vibration characteristics, and estimation of SG tube bundles degradation, respectively. The analysis shows that the tube deforms most noticeably in the zone with the highest velocity. As a result, cracks propagate more quickly in the tube with a higher height. In all simulations based on different initial crack sizes, it was observed that zone 16 experiences the greatest deformation and, subsequently, the fastest degradation, with a velocity and vapor volume fraction of 0.5 m/s and 0.4, respectively.