Failure Analysis of Steam Generator Tubes With Dented and Wastage Configurations

1978 ◽  
Vol 100 (4) ◽  
pp. 354-359
Author(s):  
M. Reich ◽  
S. Prachuktam ◽  
T. Y. Chang

The occurrence of PWR steam generator tube cracking, denting, and wastage has been reported in the recent literature [1–5]. As indicated by its title, this paper concerns itself with the inelastic structural response of the tubes that result from various assumed monotonic as well as cyclic loading conditions, which ultimately could lead to the tube failure.

Author(s):  
Jongmin Kim ◽  
Min-Chul Kim ◽  
Joonyeop Kwon

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.


Author(s):  
Hung Nguyen ◽  
Mark Brown ◽  
Shripad T. Revankar ◽  
Jovica Riznic

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.


2008 ◽  
Vol 130 (4) ◽  
Author(s):  
Xinjian Duan ◽  
Michael J. Kozluk ◽  
Sandra Pagan ◽  
Brian Mills

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.


2021 ◽  
Vol 182 ◽  
pp. 106696
Author(s):  
H.C. Ho ◽  
Y.B. Guo ◽  
M. Xiao ◽  
T.Y. Xiao ◽  
H. Jin ◽  
...  

2012 ◽  
Vol 249 ◽  
pp. 132-139 ◽  
Author(s):  
Jinbiao Xiong ◽  
Seiichi Koshizuka ◽  
Mikio Sakai ◽  
Hiroyuki Ohshima

Author(s):  
Hyun Su Kim ◽  
Jong Sung Kim ◽  
Tae Eun Jin ◽  
Hong Deok Kim ◽  
Han Sub Chung

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.


Author(s):  
Ki-Wahn Ryu ◽  
Bong-Ho Cho ◽  
Chi-Yong Park ◽  
Su-Ki Park

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.


Author(s):  
Roman Krautschneider

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.


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