Reliability of Steam Generator Tubes With Axial Cracks

1996 ◽  
Vol 118 (4) ◽  
pp. 441-446 ◽  
Author(s):  
L. Cizelj ◽  
B. Mavko ◽  
P. Vencelj

An approach for estimating the failure probability of tubes containing through-wall axial cracks has already been proposed by the authors. It is based on probabilistic fracture mechanics and accounts for scatter in tube geometry and material properties, scatter in residual and operational stresses responsible for crack propagation, and characteristics of nondestructive examination and plugging procedures (e.g., detection probability, sizing accuracy, human errors). Results of preliminary tests demonstrated wide applicability of this approach and triggered some improvements. The additions to the model are extensively discussed in this paper. Capabilities are demonstrated by results of analysis of steam generator no. 1 in Slovenian nuclear power plant located in Krsˇko after the 1992 inspection and plugging campaign. First, the number of cracked tubes and the crack length distribution were estimated using data obtained by the 100-percent motorized pancake coil inspection. The inspection and plugging activities were simulated in the second step to estimate the efficiency of maintenance in terms of single and multiple-tube rupture probabilities. They were calculated as a function of maximum allowable crack length. The importance of human errors and some limitations of present nondestructive examination techniques were identified. The traditional wall thickness and crack-length-based plugging criteria are compared. The crack-length-based criterion is shown to be more efficient and more safe, especially because of strong suppression effect on probability of multiple-tube rupture. The results are considered to be important for safety and maintenance of existing plants and for further research.

2005 ◽  
Vol 297-300 ◽  
pp. 2410-2415 ◽  
Author(s):  
Dong Hak Kim ◽  
Jeong Hyun Lee ◽  
Ho Dong Kim ◽  
Ki Ju Kang

A toughness locus Jc-Q for a ductile steel, SA106 Grade C used in the main steam piping of nuclear power plants, has been experimentally evaluated. Along with the standard fracture test procedure for J-R curve, Q as the second parameter governing stress triaxiality nearby the crack tip is measured from the displacements nearby the side necking which occurs near the crack tip on the lateral surface of a fracture specimen. The displacements nearby the side necking are measured from the digital images taken during the fracture experiment based on Stereoscopic Digital Photography (SDP) and high resolution Digital Image Correlation (DIC) software. The crack length is monitored by Direct Current Potential Drop (DCPD) method and the J-R curve is determined according to ASTM standard E1737-96. The effects of crack length, specimen geometry and thickness of specimen are studied, which are included in the toughness locus Jc-Q.


Author(s):  
Sahil Gupta ◽  
Eugene Saltanov ◽  
Igor Pioro

Canada among many other countries is in pursuit of developing next generation (Generation IV) nuclear-reactor concepts. One of the main objectives of Generation-IV concepts is to achieve high thermal efficiencies (45–50%). It has been proposed to make use of SuperCritical Fluids (SCFs) as the heat-transfer medium in such Gen IV reactor design concepts such as SuperCritical Water-cooled Reactor (SCWR). An important aspect towards development of SCF applications in novel Gen IV Nuclear Power Plant (NPP) designs is to understand the thermodynamic behavior and prediction of Heat Transfer Coefficients (HTCs) at supercritical (SC) conditions. To calculate forced convection HTCs for simple geometries, a number of empirical 1-D correlations have been proposed using dimensional analysis. These 1-D HTC correlations are developed by applying data-fitting techniques to a model equation with dimensionless terms and can be used for rudimentary calculations. Using similar statistical techniques three correlations were proposed by Gupta et al. [1] for Heat Transfer (HT) in SCCO2. These SCCO2 correlations were developed at the University of Ontario Institute of Technology (Canada) by using a large set of experimental SCCO2 data (∼4,000 data-points) obtained at the Chalk River Laboratories (CRL) AECL. These correlations predict HTC values with an accuracy of ±30% and wall temperatures with an accuracy of ±20% for the analyzed dataset. Since these correlations were developed using data from a single source - CRL (AECL), they can be limited in their range of applicability. To investigate the tangible applicability of these SCCO2 correlations it was imperative to perform a thorough error analysis by checking their results against a set of independent SCCO2 tube data. In this paper SCCO2 data are compiled from various sources and within various experimental flow conditions. HTC and wall-temperature values for these data points are calculated using updated correlations presented in [1] and compared to the experimental values. Error analysis is then shown for these datasets to obtain a sense of the applicability of these updated SCCO2 correlations.


2018 ◽  
Vol 2018 ◽  
pp. 1-12
Author(s):  
Taeseok Kim ◽  
Wonjun Choi ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Hoichul Jung ◽  
...  

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.


Author(s):  
Carlos Alexandre de Jesus Miranda ◽  
Miguel Mattar Neto

A fundamental step in tube plugging management of a Steam Generator (SG), in a Nuclear Power Plant (NPP), is the tube structural integrity evaluation. The degradation of SG tubes may be considered one of the most serious problems found in PWRs operation, mainly when the tube material is the Inconel 600. The first repair criterion was based on the degradation mode where a uniform tube wall thickness corrosion thinning occurred. Thus, a requirement of a maximum depth of 40% of the tube wall thickness was imposed for any type of tube damage. A new approach considers different defects arising from different degradation modes, which comes from the in-service inspections (NDE) and how to consider the involved uncertainties. It is based on experimental results, using statistics to consider the involved uncertainties, to assess structural limits of PWR SG tubes. In any case, the obtained results, critical defect dimensions, are within the regulatory limits. In this paper this new approach will be discussed and it will be applied to two cases (two defects) using typical data of SG tubes of one Westinghouse NPP. The obtained results are compared with ‘historical’ approaches and some comments are addressed from the results and their comparison.


Author(s):  
Deok Hyun Lee ◽  
Do Haeng Hur ◽  
Myung Sik Choi ◽  
Kyung Mo Kim ◽  
Jung Ho Han ◽  
...  

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].


2005 ◽  
Vol 235 (23) ◽  
pp. 2477-2484 ◽  
Author(s):  
Seong Sik Hwang ◽  
Hong Pyo Kim ◽  
Joung Soo Kim ◽  
Kenneth E. Kasza ◽  
Jangyul Park ◽  
...  

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