• Nuclear Engineering and Technology
• /
• 제42권1호
• /
• pp.97-108
• /
• 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
• /
• 제30권4호
• /
• pp.377-385
• /
• 1998

Many nuclear power plants have experienced unscheduled shutdown due to the leakage of steam generator tubes. The leakages are normally due to the crack, possibly stress corrosion cracking (SCC) near the tube expansion at the top of tubesheet or at the tangential point of the row-1 U-bend region. The conventional eddy current technique, which makes use of a differential bobbin coil, has been found to be inadequate for the early detection of SCC. During the in-service inspection, therefore, it is a general practice that the rotating pancake coil (RPC) is used for detecting the cracks. Even in using RPC, however, it is difficult to determine the depth of the cracks quantitatively. This paper attempts to determine the detectability and sizing ability of RPC technique for axial or circumferential cracks at the tube expansion region. The simulated cracks with various dimensions were fabricated by electro-discharge machining (EDM) method. Experimental results are discussed with theoretical calculations.

• 한국소음진동공학회:학술대회논문집
• /
• 한국소음진동공학회 2003년도 추계학술대회논문집
• /
• pp.84-91
• /
• 2003

Flow-Induced Vibration of steam generator tubes may result in fretting wear damage at the tube-to-support locations. KSNP(Korean Standard Nuclear Power plant) steam generators experienced fretting wear in the upper part of U-bend above the central cavity region of steam generators. This region has conditions susceptible to the flow-induced vibration, such as high flow velocity, high void fraction, and longer unsupported span. To improve its performance, APR1400 steam generator is designed with additional supports in this region to reduce unsupported span and to reduce peak velocity in the central cavity region. In this paper, we examined its performance improvement using ATHOS code. The thermal-hydraulic condition in the region of secondary side of APR1400 steam generator is obtained using the ATHOS3 code. The effective mass for modal analysis is calculated using the void fraction, enthalpy, and operating pressure information from ATHOS3 code result. With the effective mass distribution along the tube, natural frequency and mode shape is obtained using ANSYS code. Finally, stability ratios and real mean squared displacements for selected tubes of the APR1400 steam generator are computed. From these results, the current design of the APR1400 steam generator are examined.

• 한국소음진동공학회:학술대회논문집
• /
• 한국소음진동공학회 2004년도 춘계학술대회논문집
• /
• pp.115-118
• /
• 2004

In this study, amplitudes of turbulence excitation are obtained for selected tubes inside the KSNP SG and their normal work-rates are investigated to estimate the magnitude of fretting-wear. From the results of numerical calculation, row 40&41 tubes show the maximum work-rates. Up to this row number, the work-rates inside the row 41 have much larger values than those of outside tubes. This phenomenon reveals the particular central one which has larger normal work-rate than that of outside zone. It turns out that both of the higher local mode at the U-bend region and the larger value of effective mass in the central region Increase the normal work-rate enormously.

• 대한기계학회논문집A
• /
• 제22권1호
• /
• pp.238-246
• /
• 1998

Surface residual stresses as well as wall thickness and ovality changes after U-bending process on UNS N06690 row-1 heat exchanger tubes, were estimated. Surface residual stresses were measured by Hole Drilling Method(HDM), calculating the stresses from relieved strains of 3 rosette strain gages. After bending of the tubes, dimensional tolerances for wall thickness and ovality were satisfied with ASTM requirements. Residual stresses at the extrados were introduced with compressive stress(-) by bending operations, and its maximum value reached-319 MPa in axial direction at ${\phi}=0^{\circ}$ in position. Tensile residual stresses(+) of ${\sigma}_zz=45$ MPa,${\sigma}_zz=25$ MPa were introduced in the intrados surface at position of ${\phi}=0^{\circ}$ Maximum tensile residual stress of 170 MPa was detected on the flank side at position of ,${\phi}=95^{\circ}$i.e., at apex region. It appeared that higher stress gradients were generated at the irregular transition regions. In the trend of residual stress changes with U-bend position, the extrados is related with the changes of ovality and the intrados is related with the changes of wall thickness.

• Corrosion Science and Technology
• /
• 제17권3호
• /
• pp.116-122
• /
• 2018

The secondary coolant of a nuclear power plant has small amounts of various impurities (S, Pb, and Cl, etc.) introduced during the initial construction, maintenance, and normal operation. While the concentration of impurities in the feed water is very low, the flow of the cooling water is restricted, so impurities can accumulate on the Top of Tubesheet (TTS). This environment is chemically very complicated and has a very wide range of pH from acidic to alkaline. In this study, the characteristics of the oxide and the mechanism of stress corrosion cracking (SCC) are investigated for Alloy 690 TT in alkaline solution containing Pb, Cl, and S. Reverse U-bend (RUB) specimens were used to evaluate the SCC resistance. The test solution comprises 3m NaCl + 500ppm Pb + 0.31m $Na_2SO_4$ + 0.45m NaOH. Experimental results show that Alloy 690 TT of the crevice environment containing Pb, S, and Cl has significant cracks, indicating that Alloy 690 is vulnerable to stress corrosion cracking under this environment.

• 대한기계학회논문집A
• /
• 제30권8호
• /
• pp.993-1000
• /
• 2006

Operating experience of steam generators has shown that cracks of various morphology frequently occur in the steam generator tubes. These cracked tubes can stay in service if it is proved that the tubes have sufficient safety margin to preclude the risk of burst and leak. Therefore, integrity assessment using exact limit load solutions is very important for safe operation of the steam generators. This paper provides global and local limit load solutions for surface cracks in the steam generator tubes. Such solutions are developed based on three-dimensional (3-D) finite element analyses assuming elastic-perfectly plastic material behavior. For the crack location, both axial and circumferential surface cracks, and for each case, both external and internal cracks are considered. The resulting global and local limit load solutions are given in polynomial forms, and thus can be simply used in practical integrity assessment of the steam generator tubes.

• Journal of Mechanical Science and Technology
• /
• 제18권1호
• /
• pp.74-81
• /
• 2004

The KSNP Steam Generators (Youngkwang Unit 3 and 4, Ulchin Unit 3 and 4) have a problem of U-tube fretting wear due to Flow Induced Vibration (FIV). In particular, the wear is localized and concentrated in a small area of upper part of U-bend in the Central Cavity region. The region has some conditions susceptible to the FIV, which are high flow velocity, high void fraction, and long unsupported span. Even though the FIV could be occurred by many mechanisms, the main mechanism would be fluid-elastic instability, or turbulent excitation. To remedy the problem, Eggcrate Flow Distribution Plate (EFDP) was installed in the Central Cavity region or Ulchin Unit 5 and 6 steam generators, so that it reduces the flow velocity in the region to a certain level. However, the cause of the FIV and the effectiveness of the EFDP was not thoroughly studied and checked. In this study, therefore the Stability Ratio (SR), which is the ratio of the actual velocity to the critical velocity, was compared between the value before the installation of EFDP and that after. Also the possibility of fluid-elastic instability of KSNP steam generator and the effectiveness of EFDP were checked based on the ATHOS3 code calculation and the Pettigrew's experimental results. The calculated results were plotted in a fluid-elastic instability criteria-diagram (Pettigrew, 1998, Fig. 9). The plotted result showed that KSNP steam generator with EFDP had the margin of Fluid-Elastic Instability by almost 25％.