• 한국전산유체공학회:학술대회논문집
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• 2008.08a
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• pp.39.3-39.3
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• 2008

• Transactions of the Korean Society of Mechanical Engineers A
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• v.31 no.2 s.257
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• pp.277-284
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• 2007

The steam generator in a nuclear power plant is a large heat exchanger that uses heat from a reactor to generate steam to drive the turbine generator. Rupture of a steam generator tube can result in release of fission products to environment outside. Therefore, an accurate integrity assessment of the steam generator tubes with cracks is of great importance for maintaining the safety of a nuclear power plant. The steam generator tubes are supported at regular intervals by support plates and rotations of the tubes are restrained. Although it has been reported that the limit load for a circumferential crack is significantly affected by boundary condition of the tube, existing limit load solutions do not consider the restraining effect of support plate correctly. In addition, there are no limit load solutions for circumferential cracks in U-bend region with the effect of the support plate. This paper provides detailed limit load solutions for circumferential cracks in top of tube sheet and the U-bend regions of the steam generator tube with the actual boundary conditions to simulate the restraining effect of the support plate. Such solutions are developed based on three dimensional finite element analyses. The resulting limit load solutions are given in a polynomial form, and thus can be simply used in practical integrity assessment of the steam generator tubes.

• International Journal of Air-Conditioning and Refrigeration
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• v.15 no.4
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• pp.169-174
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• 2007

The characteristics of nine existing condensation frictional pressure drop correlations for microfin tubes were evaluated with geometries, vapor quality, mass flux, and refrigerants. The $M\ddot{u}ller-Steinhagen$ and Heck [17] smooth tube frictional pressure drop correlation was utilized to evaluate the pressure drop penalty factor (PF). Except the Nozu et al. [2], the Kedzierski and Goncalves [3], the Choi et al. [10], and the Cavallini et al. [7], other pressure drop correlations did not consider the effect of tube geometry. The prediction values for R407C by pressure drop correlations show discrepancy with previous researcher's experimental trend. Additional efforts on the development of reliable condensation pressure drop correlation for microfin tubes are still required with the systematic investigation of various effects like geometries and working conditions.

• Journal of the Korean Society for Precision Engineering
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• v.14 no.10
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• pp.85-90
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• 1997

The study presents a computer-aided analysis and its design for the forming process of .OMEGA. -type bellows tubes. Finite element analysis was carried out to perform the process simulation. Bsed on the analytic results of various conditions, the forming conditions used for angled U-type bellows tubes were determined. The 3-D modeling was constructed by I-DEAS and the process simulation was constructed by PAM- STAMP. It is concluded that the difference of height between die and bellows during the forming process causes a non-uniform shape of the bellows and also influences .OMEGA. -shape. These results can be utilized for the process design.

• Nuclear Engineering and Technology
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• v.42 no.1
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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.

• Nuclear Engineering and Technology
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• v.30 no.2
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• pp.112-127
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• 1998

The present study is to assess the applicability of the best-estimate thermal-hydraulic codes, RELAP5/MOD3.2 and CATHARE2V1.3U, to the analysis of thermal-hydraulic behavior in PWRs during midloop operation following the loss of RHRS. The codes simulate an integral test, BETHSY 6.94, which was conducted in the large scale test facility of BETHSY in France. The test represents the accident where the loss of RHRS occurs during midloop operation with the pressurizer and upper head vents open and the sight level indicator broken. Besides, the hot legs are half filled with water and the upper parts of the primary cooling system are filled with nitrogen, with a letdown line open and only one SG available. The purposes of this study are to understand the physical phenomena associated with reflux cooling in the 5G U-tubes when noncondensable gas is present under low pressure and to assess the applicability of the codes to simulate the loss of RHRS event by comparing the predictions with the test results. The results of the study may contribute to actual applications for plant safety evaluation and description of the emergency operating procedure.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.24 no.2 s.173
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• pp.279-289
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• 2000

Residual stress induced in U-bending and tube-to-tubesheet joint processes of PWR's row-1 heat exchanger tube was measured by X-ray method and Hole-Drilling Method(HDM). Compressive residual stresses(-) at the extrados surface were induced in U-bending, and its maximum value reached -319 MPa in axial direction at the position of $\psi$ = $0^{\circ}$. Tensile residual stresses(+) of $\sigma_{zz}$ = 45 MPa and $\sigma_{\theta\theta}$ = 25 MPa were introduced in the intrados surface at the position of $\psi$ = $0^{\circ}$. Maximum tensile residual stress of 170 MPa was measured at the flank side at the position of $\psi$ = $90^{\circ}$, i.e., at apex region. It was observed that higher stress gradient was generated at the irregular transition regions (ITR). The trend of residual stress induced by U bending process of the tubes was found to be related with the change of ovality. The residual stress induced by the explosive joint method was found to be lower than that by the mechanical roll method. The gradient of residual stress along the expanded tube was highest at the transition region (TR), and the residual stress in circumferential direction was found to be higher than the residual stress in axial direction.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.177-177
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.143.2-143.2
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• 2003

• 한국정보디스플레이학회:학술대회논문집
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• 2009.10a
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• pp.1269-1271
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• 2009

We had proposed fast thermal stress estimation methodology for the components on system board when the system is stationary within specific ambient air temperature. Now, we will propose one efficient thermal stress estimation methodology, block adaptive filtering methodology, for the FPD electronic system board which is enclosed by mechanical cover.