High cycle thermal fatigue Issues in PWR nuclear power plants, life time improvement of some austenitic stainless steel components

2007 ◽  
Vol 104 (3) ◽  
pp. 156-162 ◽  
Author(s):  
J. -A. Le Duff ◽  
A. Lefrançois ◽  
Y. Meyzaud ◽  
J.-Ph. Vernot ◽  
D. Martin ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Thermal Fatigue ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Life Time ◽  
High Cycle Thermal Fatigue

Study of the Crack Propagation Behaviour of 304L Austenitic Stainless Steel Under Thermal Fatigue

2011 ◽  
Author(s):  
Pauline Bouin ◽  
Antoine Fissolo ◽  
Ce´dric Gourdin
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Crack Propagation ◽  
Crack Growth ◽  
Thermal Fatigue ◽  
Nuclear Power ◽  
Power Plants ◽  
Numerical Study ◽  
Fatigue Cracks ◽  
Life Time

Thermal fatigue phenomena are long term deterioration mechanisms which become more and more important as the life-time of nuclear power plant increases. Some incidents as the incident of Civaux I and some works in thermal fatigue have disproved current methodologies and usual criteria to predict propagation of thermal fatigue cracks in nuclear power plants. This paper presents the results of the thermal fatigue tests, Fat3D, which are conducted on 304L austenitic stainless steel pipes. This experiment has been designed to study the problem of closure effect and fatigue crack growth under thermal fatigue conditions on quasi-structure specimens. The importance of the initiation and the propagation phases on a notched specimen and the evolution of the stress intensity factor according to the propagation are investigated as well. The use of different non destructive techniques to detect and follow crack propagation is also assessed. In parallel, a numerical interpretation is developed based on a material characterisation and using finite element analyses with the French Cast3M code. This combined experimental and numerical study enables to assess improvements of classical methods to accurately predict the crack growth propagation under thermal loads and to understand the influence of the main parameters concerning crack propagation in such components.

Current in-service inspections of austenitic stainless steel and dissimilar metal welds in light water nuclear power plants

1994 ◽  
Vol 151 (2-3) ◽  
pp. 539-550 ◽  
Author(s):  
Ludwig von Bernus ◽  
Werner Rathgeb ◽  
Rudi Schmid ◽  
Friedrich Mohr ◽  
Michael Kröning
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Light Water ◽  
Dissimilar Metal

scholarly journals Parameters influencing the low-cycle fatigue life of materials in pressure water reactor nuclear power plants / Parametry ovlivòující únavové chování materiálù pro tlakovodní jaderné reaktory v režimu nízkocyklové únavy

2015 ◽  
Vol 59 (3) ◽  
pp. 91-98
Author(s):  
V. Šefl
Keyword(s):  
Stainless Steel ◽  
Fatigue Life ◽  
Carbon Steel ◽  
Austenitic Stainless Steel ◽  
Literature Review ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Low Cycle Fatigue ◽  
Cycle Fatigue

Abstract In this literature review we identify and quantify the parameters influencing the low-cycle fatigue life of materials commonly used in nuclear power plants. The parameters are divided into several groups and individually described. The main groups are material properties, mode of cycling and environment parameters. The groups are further divided by the material type - some parameters influence only certain kind of material, e.g. sulfur content may decreases fatigue life of carbon steel, but is not relevant for austenitic stainless steel; austenitic stainless steel is more sensitive to concentration of dissolved oxygen in the environment compared to the carbon steel. The combination of parameters i.e. conjoint action of several detrimental parameters is discussed. It is also noted that for certain parameters to decrease fatigue life, it is necessary for other parameter to reach certain threshold value. Two different approaches have been suggested in literature to describe this complex problem - the Fen factor and development of new design fatigue curves. The threshold values and examples of commonly used relationships for calculation of fatigue lives are included. This work is valuable because it provides the reader with long-term literature review with focus on real effect of environmental parameters on fatigue life of nuclear power plant materials.

Development of Load Multipliers (Z-Factors) in Elastic-Plastic Fracture Mechanics Evaluation in Rules on Fitness-for-Service for Nuclear Power Plants

2012 ◽  
Author(s):  
Seiji Asada ◽  
Masao Itatani ◽  
Naoki Miura ◽  
Hideo Machida
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Fracture Mechanics ◽  
Ferritic Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Elastic Plastic ◽  
Surface Flaws ◽  
Steel Piping

Not only nonmandatory Appendix C, “Evaluation of Flaws in Piping,” in ASME Boiler & Pressure Vessel Code Section XI but also Appendix E-9, “Elastic-Plastic Fracture Mechanics Evaluation,” in the JSME Rules on Fitness-for-Service for Nuclear Power Plants use the load multiplier Z-factor that is applied to elastic-plastic fracture mechanics evaluation for a circumferential flaw of austenitic stainless steel piping and ferritic steel piping. The Z-factor is defined as the ratio of the limit load to the load at fracture load. Basically, the Z-factor equations were conservatively formulated by using the Z-factors for circumferential through-wall flaws. However, the Codes require flaw evaluation for circumferential surface flaws. Accordingly, Z-factors for circumferential surface flaws should be developed to have the consistency. Therefore Z-factor equations of austenitic stainless steel piping and ferritic steel piping have been developed for circumferential surface flaws.

Phased Array Ultrasonic Inspection Technique for Cast Austenitic Stainless Steel Parts of Nuclear Power Plants

2016 ◽  
Author(s):  
Setsu Yamamoto ◽  
Jun Semboshi ◽  
Azusa Sugawara ◽  
Makoto Ochiai ◽  
Kentaro Tsuchihashi ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Ultrasonic Testing ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Phased Array ◽  
Ultrasonic Inspection ◽  
Inspection Time ◽  
Accuracy Measurement

For safety operation of nuclear power plants, soundness assurance of structures has been strongly required. In order to evaluate properties of inner defects at plant structures quantitatively, non-destructive inspection using ultrasonic testing (UT) has performed an important role for plant maintenances. At nuclear power plants, there are many structures made of cast austenitic stainless steel (e.g. casings, valve gages, pipes and so on). However, UT has not achieved enough accuracy measurement at cast stainless steels due to the noise from large grains. In order to overcome the problem, we have developed comprehensively analyzable phased array ultrasonic testing (PAUT) system. We have been noticing that dependency of echo intensity from defect is different from grain noises when PAUT conditions (for example, ultrasonic incident angles and focal depths) were continuously changed. Analyzing the tendency of echoes from comprehensive PAUT conditions, defect echoes could be distinguished from the noises. Meanwhile, in order to minimize the inspection time on-site, we have developed the algorithms and the full matrix capture (FMC) data acquisition system. In this paper, the authors confirmed the detectability of the PAUT system applying cast austenitic stainless steel (316 stainless steel) specimens which have sand-blasted surface and 3 slits which made by electric discharge machining (EDM).

Analysis of the FATHER Experiment With an Engineering Method Devoted to High Cycle Thermal Fatigue

2011 ◽  
Author(s):  
Romain Beaufils ◽  
Stephan Courtin
Keyword(s):  
Fatigue Damage ◽  
Thermal Fatigue ◽  
Turbulent Mixing ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Industry ◽  
Engineering Method ◽  
Different Temperatures ◽  
High Cycle Thermal Fatigue

Nuclear power plants necessarily include connections of branches conveying fluids at different temperatures. Thermo-hydraulic fluctuations arising from the turbulent mixing of the flows may affect the inner wall of the pipes and lead to fatigue damage. The FATHER experiment was carried out in order to better understand this phenomenon of High Cycle Thermal Fatigue (HCTF). The aim of this paper is to present an analysis of the FATHER experiment with a simplified engineering method which is used in French nuclear industry to identify and classify zones presenting a HCTF risk.

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