Estimation of Plastic Collapse Load of Steam Generator Tube With Two Parallel Axial Through-Wall Cracks

2003 ◽  
Author(s):  
S. I. Moon ◽  
Y. J. Kim ◽  
J. H. Lee
Keyword(s):  
Steam Generator ◽  
Local Failure ◽  
Plastic Collapse ◽  
Steam Generator Tube ◽  
Coalescence Model ◽  
Collapse Model ◽  
Failure Models ◽  
Optimum Model ◽  
Steam Generator Tubes

The 40% of wall criterion, which is generally used for the plugging of steam generator tubes, is applied only to a single crack. In the previous study, a total number of 9 failure models were introduced to estimate the local failure of the ligament between cracks and the optimum coalescence model of multiple collinear cracks was determined among these models. It is, however, known that parallel axial cracks are more frequently detected during an in-service inspection than collinear axial cracks. The objective of this study is to determine the plastic collapse model which can be appfied to the steam generator tube containing two parallel axial through-wall cracks. Three previously proposed local failure models were selected as the candidates. Subsequently interaction effects between two adjacent cracks were evaluated to screen them. Plastic collapse tests for the plate with two parallel through-wall cracks and finite element analyses were performed for the determination of the optimum plastic collapse model. By comparing the test results with the prediction results obtained from the candidate models, a COD base model was selected as an optimum model.

Optimum Global Failure Prediction Model of Steam Generator Tube With Two Parallel Axial Through-Wall Cracks

2005 ◽  
Vol 127 (2) ◽  
pp. 123-128 ◽  
Author(s):  
Seong-In Moon ◽  
Young-Jin Kim ◽  
Jin-Ho Lee ◽  
Myung-Ho Song ◽  
Youn-Won Park
Keyword(s):  
Steam Generator ◽  
Crack Opening ◽  
Local Failure ◽  
Plastic Collapse ◽  
Opening Displacement ◽  
Steam Generator Tube ◽  
Collapse Model ◽  
Failure Models ◽  
Optimum Model

The 40% of wall criterion, which is generally used for the plugging of steam generator tubes, is applied only to a single crack. In the previous study (Moon et al. (2002)), a total number of 9 failure models were introduced to estimate the local failure of the ligament between cracks and the optimum coalescence model of multiple collinear cracks was determined among these models. It is, however, known that parallel axial cracks are more frequently detected during an in-service inspection than collinear axial cracks. The objective of this study is to determine the plastic collapse model which can be applied to the steam generator tube containing two parallel axial through-wall cracks. Three previously proposed local failure models were selected as the candidates. Subsequently interaction effects between two adjacent cracks were evaluated to screen them. Plastic collapse tests for the plate with two parallel through-wall cracks and finite element analyses were performed for the determination of the optimum plastic collapse model. By comparing the test results with the prediction results obtained from the candidate models, a crack opening displacement (COD) base model was selected as an optimum model.

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.

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.

scholarly journals A Bayesian updating of crack distributions in steam generator tubes

2021 ◽  
Vol 30 (2) ◽  
pp. 33-44
Author(s):  
Alexandre Santos Francisco ◽  
Tiago Simões
Keyword(s):  
Steam Generator ◽  
Failure Probability ◽  
Nuclear Power ◽  
Power Plants ◽  
Plastic Collapse ◽  
Structural Failure ◽  
Maintenance Strategies ◽  
Threshold Size ◽  
Steam Generator Tubes ◽  
Good Agreement

The structural failure of steam generator tubes is a common problem that can a ect the availability and safety of nuclear power plants. To minimize the probability of occurrence of failure, it is needed to implement maintenance strategies such as periodic nondestructive inspections of tubes. Thus, a tube is repaired or plugged whenever it has detected a crack which a threshold size is overtaken. In general, uncertainties and errors in crack sizes are associated with the nondestructive inspections. These uncertainties and errors should be appropriately characterized to estimate the actual crack distribution. This work proposes a Bayesian approach for updating crack distributions, which in turn allows computing the failure probability of steam generator tubes at current and future times. The failure criterion is based on plastic collapse phenomenon, and the failure probability is computed by using the Monte-Carlo simulation. The failure probability at current and future times is in good agreement with the ones presented in the literature.

Evaluation of Interaction Effects on Part-Through Cracks in Steam Generator Tubes

2013 ◽  
Author(s):  
Yong-Seok Kang ◽  
Hong-Deok Kim ◽  
Kuk-Hee Lee ◽  
Jai-Hak Park
Keyword(s):  
Steam Generator ◽  
Interaction Effect ◽  
Structural Integrity ◽  
Structural Parameters ◽  
Interaction Effects ◽  
Multiple Cracks ◽  
Performance Limit ◽  
Steam Generator Tubing ◽  
Steam Generator Tubes

Degraded steam generator tubing can affect its safety functions. Therefore, its integrity should be maintained for each degradation form and all detected degradation must be assessed to verify that if adequate integrity is retained. Determination of tube integrity limits includes identifying acceptable structural parameters such as flaw length, depth, and amplitude of signals. If we consider just single-cracked tubes, short and deep flaws are not likely to threaten structural integrity of tubes. But if it has multiple-cracks, we have to consider interaction effects of multiple adjacent cracks on its burst pressure. Because adjacent multiple cracks can be merged due to the crack growth then it can challenge against the structural performance limit. There are some studies on the interaction effects of adjacent cracks. However, existing works on the interaction effect consider only through-wall cracks. No study has been carried out on the interaction effects of part-through cracks. Most cracks existing in real steam generator tubing are not through-wall cracks but part-through cracks. Hence, integrity of part-through cracks is more practical issue than that of through-wall cracks. This paper presents experimental burst test results with steam generator tubing for evaluation of interaction effects with axial oriented two collinear and parallel part-through cracks. The interaction effect between two adjacent cracks disappeared when the distance exceeds about 2 mm.

Evaluation of Plastic Collapse Pressure for Steam Generator Tube with Non-Aligned Two Axial Through-Wall Cracks

2005 ◽  
Vol 29 (8) ◽  
pp. 1070-1077 ◽  
Author(s):  
Seong-In Moon ◽  
Yoon-Suk Chang ◽  
Jin-Ho Lee ◽  
Myung-Ho Song ◽  
Young-Hwan Choi ◽  
...  
Keyword(s):  
Steam Generator ◽  
Plastic Collapse ◽  
Steam Generator Tube ◽  
Collapse Pressure

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.

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