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.

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.

Failure Pressure for Steam Generator Tube With Two Different Collinear Axial Through-Wall Cracks

2006 ◽  
Author(s):  
Nam-Su Huh ◽  
Yoon-Suk Chang ◽  
Young-Jin Kim
Keyword(s):  
Steam Generator ◽  
Complex Shape ◽  
Structural Integrity ◽  
Failure Behavior ◽  
Multiple Cracks ◽  
Single Crack ◽  
Steam Generator Tube ◽  
3 Dimensional ◽  
Failure Pressure ◽  
Steam Generator Tubes

To maintain the structural integrity of steam generator tubes, 40% of wall thickness plugging criterion has been developed. The approach is for the steam generator tube with single crack, so that the interaction effect of multiple cracks can not be considered. Although, recently, several approaches has been proposed to assess the integrity of steam generator tube with two identical cracks whilst actual multiple cracks reveal more complex shape. In this paper, the failure pressure of steam generator tube containing multiple cracks of different length is evaluated based on the detailed 3-dimensional elastic-plastic finite element (FE) analyses. In terms of the crack shape, two collinear axial through-wall cracks with different length were considered. Furthermore, the resulting FE failure pressures are compared with FE failure pressures and experimental results for two identical collinear axial through-wall cracks to quantify the effect of crack length ratio on failure behavior of steam generator tube with multiple cracks.

U.S. Operating Experience With Thermally Treated Alloy 600 Steam Generator Tubes

2002 ◽  
Author(s):  
April Smith ◽  
Kenneth J. Karwoski
Keyword(s):  
Steam Generator ◽  
Operating Experience ◽  
Alloy 600 ◽  
Steam Generators ◽  
Pressurized Water ◽  
Steam Generator Tube ◽  
Water Reactors ◽  
1960S And 1970S ◽  
The 1960S ◽  
Thermally Treated

Steam generators placed in service in the 1960s and 1970s were primarily fabricated from mill-annealed Alloy 600. Over time, this material proved to be susceptible to stress corrosion cracking in the highly pure primary and secondary water chemistry environments of pressurized-water reactors. The corrosion ultimately led to the replacement of steam generators at numerous facilities, the first U.S. replacement occurring in 1980. Many of the steam generators placed into service in the 1980s used tubes fabricated from thermally treated Alloy 600. This tube material was thought to be less susceptible to corrosion. Because of the safety significance of steam generator tube integrity, this paper evaluates the operating experience of thermally treated Alloy 600 by looking at the extent to which it is used and recent results from steam generator tube examinations.

Evaluation of Creep Properties of Alloy 690 Steam Generator Tubes at High Temperature Using Tube Specimen

2019 ◽  
Author(s):  
Jongmin Kim ◽  
Woogon Kim ◽  
Minchul Kim
Keyword(s):  
High Temperature ◽  
Steam Generator ◽  
Creep Test ◽  
Rupture Life ◽  
Creep Rupture ◽  
Severe Accident ◽  
High Temperature Creep ◽  
Creep Life ◽  
Temperature Creep ◽  
Alloy 690

Abstract Thermally induced steam generator (SG) tube failures caused by hot gases from a damaged reactor core can result in a containment bypass event and may lead to release of fission products to the environment. A typical severe accident scenario is a station blackout (SBO) with loss of auxiliary feedwater. Alloy 690 which has increased the Cr content has been replaced for the SG tube due to its high corrosion resistance against stress corrosion cracking (SCC). However, there is lack of research on the high temperature creep rupture and life prediction model of Alloy 690. In this study, creep test was performed to estimate the high temperature creep rupture life of Alloy 690. Based on reported creep data and creep test results of Alloy 690 in this study, creep life extrapolation was carried out using Larson-Miller Parameter (LMP), Orr-Sherby-Dorn (OSD), Manson-Haferd Parameter (MHP), and Wilshire’s approach. And a hyperbolic sine (sinh) function to determine master curves in LMP, OSD and MHP methods was used for improving the creep life estimation of Alloy 690 material.

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.

Coalescence Criterion of Part-Through Wall Cracks in Steam Generator Tubes of Nuclear Power Plants

2004 ◽  
Author(s):  
Jeries Abou-Hanna ◽  
Timothy McGreevy ◽  
Saurin Majumdar ◽  
Amit J. Trivedi ◽  
Ashraf Al-Hayek
Keyword(s):  
Finite Element ◽  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Local Instability ◽  
Element Analysis ◽  
Failure Pressure ◽  
Ligament Failure ◽  
Steam Generator Tubes

In scheduling inspection and repair of nuclear power plants, it is important to predict failure pressure of cracked steam generator tubes. Nondestructive evaluation (NDE) of cracks often reveals two neighboring cracks. If two neighboring part-through cracks interact, the tube pressure, under which the ligament between the two cracks fails, could be much different than the critical burst pressure of an individual equivalent part-through crack. The ability to accurately predict the ligament failure pressure, called “coalescence pressure,” is important. The coalescence criterion, established earlier for 100% through cracks using nonlinear finite element analyses [1–3], was extended to two part-through-wall axial collinear and offset cracks cases. The ligament failure is caused by local instability of the radial and axial ligaments. As a result of this local instability, the thickness of both radial and axial ligaments decreases abruptly at a certain tube pressure. Good correlation of finite element analysis with experiments (at Argonne National Laboratory’s Energy Technology Division) was obtained. Correlation revealed that nonlinear FEM analyses are capable of predicting the coalescence pressure accurately for part-through-wall cracks. This failure criterion and FEA work have been extended to axial cracks of varying ligament width, crack length, and cases where cracks are offset by axial or circumferential ligaments. The study revealed that rupture of the radial ligament occurs at a pressure equal to the coalescence pressure in the case of axial ligament with collinear cracks. However, rupture pressure of the radial ligament is different from coalescence pressure in the case of circumferential ligament, and it depends on the length of the ligament relative to crack dimension.

scholarly journals Micro-Galvanic Corrosion of Steam Generator Materials within Pores of Magnetite Flakes in Alkaline Solutions

Metals ◽  
2018 ◽  
Vol 8 (11) ◽  
pp. 899 ◽  
Author(s):  
Soon-Hyeok Jeon ◽  
Geun Song ◽  
Do Hur
Keyword(s):  
Steam Generator ◽  
Galvanic Corrosion ◽  
Area Ratio ◽  
Corrosion Products ◽  
Alkaline Solutions ◽  
Galvanic Coupling ◽  
Pressurized Water ◽  
Steam Generator Tube ◽  
Alloy 690 ◽  
Water Reactors

In secondary coolant system of the pressurized water reactors, the reduced corrosion products such as metallic Cu and Pb particles were accumulated in the pores of the magnetite flakes and electrically contacted to the steam generator materials. The micro-galvanic corrosion behavior of steam generator materials (steam generator tube materials: Alloy 600 and Alloy 690, steam generator tube sheet materials: SA508 Gr.3) contacted to the corrosion products (magnetite, Cu, and Pb) was investigated in an alkaline solution. The steam generator materials considered in this study were all the anodic elements of the galvanic pair because their corrosion potentials were lower than those of the corrosion products. The corrosion rate of the steam generator materials was increased by the galvanic coupling with the each corrosion products, and was more accelerated with increasing the area ratio of the corrosion products to the steam generator materials. Among the corrosion products, Cu has the largest galvanic effect on steam generator materials in the pores when area ratio of cathode to anode is 10.

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