Analysis of Fluid-Elastic Instability for KSNP Steam Generator Tube and Its Plugging Effect at Central Region

2003 ◽  
Author(s):  
Ki-Wahn Ryu ◽  
Bong-Ho Cho ◽  
Chi-Yong Park ◽  
Su-Ki Park
Keyword(s):  
Steam Generator ◽  
Central Region ◽  
Outer Zone ◽  
Elastic Instability ◽  
Stability Ratio ◽  
Steam Generator Tube ◽  
Instability Analysis ◽  
Integrity Assessment ◽  
The Stability ◽  
Steam Generator Tubes

The characteristics of fluid-elastic instability for the KSNP steam generator tubes were investigated numerically. The information on the thermal-hydraulic data of the steam generator has been obtained by using the ATHOS3-MOD1 code and the fluid-elastic instability analysis has been conducted by using the PIAT (Program for Integrity Assessment of Steam Generator Tube) code. The KSNP steam generator has the concentrated plugging zone at the vicinity of the stay cylinder inside the steam generator. To investigate the cause of the concentrated plugging, the fluid-elastic instability analysis has been performed on various column and row number of the KSNP steam generator tubes. From the results of this study the stability ratio due to the fluid-elastic instability in the concentrated plugged zone tend to have larger values than those of the outer zone. Even though the further study will still be required, these results seem to be related with concentrated plugging inside the steam generator. And the stability ratio of plugged tube does not have any consistent advantages for all modes over the normal one. This seems to be caused by the decrease of mass, the increase of natural frequency, and the change of mode shape after plugging.

FIV Characteristics of U-Tubes Due to Relocation of the Tube Support Plate

2005 ◽  
Author(s):  
Ki-Wahn Ryu ◽  
Hyung-Jin Kim ◽  
Chi-Yong Park
Keyword(s):  
Aspect Ratio ◽  
Steam Generator ◽  
Mode Shapes ◽  
Design Stage ◽  
Parameter Study ◽  
Stability Ratio ◽  
Integrity Assessment ◽  
Support Plate ◽  
Aspect Ratios ◽  
The Stability

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.

Structural Integrity Assessment of Steam Generator Tube by the Use of Heterogeneous Finite Element Method

2008 ◽  
Vol 130 (4) ◽  
Author(s):  
Xinjian Duan ◽  
Michael J. Kozluk ◽  
Sandra Pagan ◽  
Brian Mills
Keyword(s):  
Finite Element ◽  
Steam Generator ◽  
Failure Modes ◽  
Structural Integrity ◽  
Fretting Wear ◽  
Defect Depth ◽  
Steam Generator Tube ◽  
Integrity Assessment ◽  
Failure Pressure ◽  
Steam Generator Tubes

Aging steam generator tubes have been experiencing a variety of degradations such as pitting, fretting wear, erosion-corrosion, thinning, cracking, and denting. To assist with steam generator life cycle management, some defect-specific flaw models have been developed from burst pressure testing results. In this work, an alternative approach; heterogeneous finite element model (HFEM), is explored. The HFEM is first validated by comparing the predicted failure modes and failure pressure with experimental measurements of several tubes. Several issues related to the finite element analyses such as temporal convergence, mesh size effect, and the determination of critical failure parameters are detailed. The HFEM is then applied to predict the failure pressure for use in a fitness-for-service condition monitoring assessment of one removed steam generator tube. HFEM not only calculates the correct failure pressure for a variety of defects, but also predicts the correct change of failure mode. The Taguchi experimental design method is also applied to prioritize the flaw dimensions that affect the integrity of degraded steam generator tubes such as the defect length, depth, and width. It has been shown that the defect depth is the dominant parameter controlling the failure pressure. The failure pressure varies almost linearly with defect depth when the defect length is greater than two times the tube diameter. An axial slot specific flaw model is finally developed.

Assessment of Limit Loads for Circumferential Cracks in Steam Generator Tube Considering Constraining Effect of Tube Support

2005 ◽  
Author(s):  
Hyun Su Kim ◽  
Jong Sung Kim ◽  
Tae Eun Jin ◽  
Hong Deok Kim ◽  
Han Sub Chung
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Three Dimensional ◽  
Fission Products ◽  
Limit Load ◽  
Steam Generator Tube ◽  
Integrity Assessment ◽  
Steam Generator Tubes

The steam generator in a nuclear power plant is a large heat exchanger that uses heat from reactor to generate steam to drive the turbine generators. Rupture of a steam generator tube can result in release of fission products to environment. Therefore, an accurate integrity assessment of the steam generator tubes with cracks is of great importance for maintaining safety of a nuclear power plant. The steam generator tubes are supported at periodic intervals by support plates and rotations of the tubes are constrained. Although it was reported that the limit load for a circumferential crack was significantly affected by boundary condition of the tube, existing limit load solutions do not include the constraining effect of tube supports. This paper provides detailed limit load solutions for circumferential cracks in steam generator tubes considering the actual boundary conditions to simulate the constraining effect of the tube supports. Such solutions are developed based on three dimensional (3D) 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.

PWR and WWER Steam Generator Tube Integrity Assessment

2013 ◽  
Author(s):  
Roman Krautschneider
Keyword(s):  
Steam Generator ◽  
Operating Conditions ◽  
Degradation Mechanisms ◽  
Steam Generator Tube ◽  
Integrity Assessment ◽  
Steam Generator Tubes

Paper is describing and comparing degradation mechanisms and integrity assessment of PWR and WWER type of steam generator tubes. Because of different design, different used materials and also different operating conditions, there are significant differences in degradation mechanisms. Therefore both steam generator types have their specific codes dealing with inspection, monitoring and maintenance.

Evaluation of the Creep Rupture Behavior of Alloy 690 Steam Generator Tubes Considering the Pressure Ramp Rate

2021 ◽  
Author(s):  
Jongmin Kim ◽  
Min-Chul Kim ◽  
Joonyeop Kwon
Keyword(s):  
Steam Generator ◽  
Creep Rupture ◽  
Parameter Method ◽  
Alloy 600 ◽  
Steam Generator Tube ◽  
Failure Pressure ◽  
Alloy 690 ◽  
Rupture Model ◽  
Temperature And Pressure ◽  
Steam Generator Tubes

Abstract The materials used previously for steam generator tubes around the world have been replaced and will be replaced by Alloy 690 given its improved corrosion resistance relative to that of Alloy 600. However, studies of the high- temperature creep and creep-rupture characteristics of steam generator tubes made of Alloy 690 are insufficient compared to those focusing on Alloy 600. In this study, several creep tests were conducted using half tube shape specimens of the Alloy 690 material at temperatures ranging from 650 to 850C and stresses in the range of 30 to 350 MPa, with failure times to creep rupture ranging from 3 to 870 hours. Based on the creep test results, creep life predictions were then made using the well-known Larson Miller Parameter method. Steam generator tube rupture tests were also conducted under the conditions of a constant temperature and pressure ramp using steam generator tube specimens. The rupture test equipment was designed and manufactured to simulate the transient state (rapid temperature and pressure changes) in the event of a severe accident condition. After the rupture test, the damage to the steam generator tubes was predicted using a creep rupture model and a flow stress model. A modified creep rupture model for Alloy 690 steam generator tube material is proposed based on the experimental results. A correction factor of 1.7 in the modified creep rupture model was derived for the Alloy 690 material. The predicted failure pressure was in good agreement with the experimental failure pressure.

Limit Loads for Circumferential Cracks in the Steam Generator Tube under Combined Pressure and Bending Loads

2007 ◽  
Vol 345-346 ◽  
pp. 1357-1360
Author(s):  
Hyun Su Kim ◽  
Tae Eun Jin ◽  
Hong Deok Kim ◽  
Han Sub Chung ◽  
Yoon Suk Chang ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Plastic Material ◽  
Material Behavior ◽  
Limit Loads ◽  
Integrity Assessment ◽  
Bending Loads ◽  
Steam Generator Tubes

Steam generator in a nuclear power plant is huge heat exchanger that transfers heat from reactor to make steam to drive turbine-generator. Failure of the steam generator tubes can result in the release of fission products to the secondary side. Therefore, accurate integrity assessment of the cracked steam generator tubes is of great importance for maintaining the safety of the nuclear power plant. This paper provides limit loads for circumferential through-wall cracks in steam generator tubes under combined internal pressure and bending loads. Such limit loads are developed on the basis of three dimensional finite element analyses assuming elastic-perfectly plastic material behavior. As for the crack location, both the top of the tubesheet and U-bend regions are considered. The analysis results can be directly applied to the practical integrity assessment of cracked steam generator tubes, because the comparison between experimental data and FE results shows a very good agreement.

Experimental Investigation of Subcooled Choking Flow in a Steam Generator Tube Crack

2016 ◽  
Author(s):  
Hung Nguyen ◽  
Mark Brown ◽  
Shripad T. Revankar ◽  
Jovica Riznic
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Test Facility ◽  
Experimental Program ◽  
Steam Generators ◽  
Steam Generator Tube ◽  
Loss Of Coolant ◽  
Steam Generator Tubes

Steam generator tubes have a history of small cracks and even ruptures, which lead to a loss of coolant from the primary side to the secondary side. These tubes have an important role in reactor safety since they serve as one of the barriers between radioactive and non-radioactive materials of a nuclear power plant. A rupture then signifies the loss of the integrity of the tube itself. Therefore, choking flow plays an integral part not only in the engineered safeguards of a nuclear power plant, but also to everyday operation. There is limited data on actual steam generators tube wall cracks. Here experiments were conducted on choked flow of subcooled water through two samples of axial cracks of steam generator tubes taken from US PWR steam generators. The purpose of the experimental program was to develop database on critical flow through actual steam generator tube cracks with subcooled liquid flow at the entrance. The knowledge of this maximum flow rate through a crack in the steam generator tubes of a pressurized water nuclear reactor will allow designers to calculate leak rates and design inventory levels accordingly while limiting losses during loss of coolant accidents. The test facility design is modular so that various steam generator tube cracks can be studied. Two sets of PWR steam generators tubes were studied whose wall thickness is 1.285 mm. Tests were carried out at stagnation pressure up to 6.89 MPa and range of subcoolings 16.2–59°C. Based on these new choking flow data, the applicability of analytical models to highlight the importance of non-equilibrium effects was examined.

Plastic Limit Load Solutions for Surface Crack in the Steam Generator Tubes

2005 ◽  
Vol 297-300 ◽  
pp. 1704-1712
Author(s):  
Ouk Sub Lee ◽  
Hyun Su Kim ◽  
Jong Sung Kim ◽  
Tae Eun Jin ◽  
Hong Deok Kim ◽  
...  
Keyword(s):  
Steam Generator ◽  
Safety Margin ◽  
Local Limit ◽  
Limit Load ◽  
Material Behavior ◽  
Surface Cracks ◽  
Steam Generators ◽  
Integrity Assessment ◽  
Global And Local ◽  
Steam Generator Tubes

Operating experience of steam generators has shown that cracks of various morphologies 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, because the comparison between experimental data and FE solutions shows good agreement.

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