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.

The Geometry of Steam Generator Tube and Its Relevance to the Occurrence of Stress Corrosion Cracking in Operating Nuclear Power Plants

2010 ◽  
Author(s):  
Deok Hyun Lee ◽  
Do Haeng Hur ◽  
Myung Sik Choi ◽  
Kyung Mo Kim ◽  
Jung Ho Han ◽  
...  
Keyword(s):  
Stress Corrosion ◽  
Steam Generator ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Corrosion Cracking ◽  
Location Type ◽  
Steam Generator Tubes ◽  
Non Destructive

Occurrences of a stress corrosion cracking in the steam generator tubes of operating nuclear power plants are closely related to the residual stress existing in the local region of a geometric change, that is, expansion transition, u-bend, ding, dent, bulge, etc. Therefore, information on the location, type and quantitative size of a geometric anomaly existing in a tube is a prerequisite to the activity of a non destructive inspection for an alert detection of an earlier crack and the prediction of a further crack evolution [1].

A New Eddy Current 3-D Profilometry for a Quantitative Measurement of Geometric Anomaly in Steam Generator Tubes of Nuclear Power Plants

2006 ◽  
Vol 321-323 ◽  
pp. 426-429
Author(s):  
Deok Hyun Lee ◽  
Myung Sik Choi ◽  
Do Haeng Hur ◽  
Jung Ho Han ◽  
Myung Ho Song ◽  
...  
Keyword(s):  
Steam Generator ◽  
Eddy Current ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Three Dimensional ◽  
Accurate Information ◽  
Profile Measurement ◽  
Root Cause ◽  
Steam Generator Tubes

Most of the corrosive degradations in steam generator tubes of nuclear power plants are closely related to the residual stress existing in the local region of a geometric change, that is, an expansion transition, u-bend, ding, dent, bulge, etc. Therefore, accurate information on a geometric anomaly in a tube is a prerequisite to the activity of a non destructive inspection for a precise and earlier detection of a defect in order to prevent a failure during an operation, and also for a root cause analysis of a failure. In this paper, a newly developed eddy current technique of a three-dimensional profilometry is introduced and the proof for the applicability of the technique to a plant inspection is provided. The quantitative profile measurement using a new eddy current probe was performed on steam generator expansion mock-up tubes with various geometric anomalies typically observed in the operating power plants, and the accuracy of the measured data was compared with those from the laser profilometry.

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.

Effects of curative and preventive chemical cleaning processes on fouled steam generator tubes in nuclear power plants

2017 ◽  
Vol 323 ◽  
pp. 120-132 ◽  
Author(s):  
C. Goujon ◽  
Th. Pauporté ◽  
A. Bescond ◽  
C. Mansour ◽  
S. Delaunay ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Chemical Cleaning ◽  
Steam Generator Tubes
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