scholarly journals Fatigue Damage Evaluation of Stainless Steel Pipes in Nuclear Power Plants Using Positron Annihilation Lineshape Analysis

2002 ◽  
Vol 66 (7) ◽  
pp. 740-744 ◽  
Author(s):  
Yasuhiro Kawaguchi ◽  
Noriko Nakamura ◽  
Satoru Yusa
2014 ◽  
Vol 487 ◽  
pp. 54-57 ◽  
Author(s):  
Meng Yu Chai ◽  
Li Chan Li ◽  
Wen Jie Bai ◽  
Quan Duan

304 stainless steel and 316L stainless steel are conventional materials of primary pipeline in nuclear power plants. The present work is to summarize the acoustic emission (AE) characteristics in the process of pitting corrosion of 304 stainless steel, intergranular corrosion of 316L stainless steel and weldments of 316L stainless steel. The work also discussed the current shortcomings and problems of research. At last we proposed the coming possible research topics and directions.


2006 ◽  
Vol 321-323 ◽  
pp. 743-746 ◽  
Author(s):  
Jong Ho Park ◽  
Joon Hyun Lee ◽  
Gyeong Chul Seo ◽  
Sang Woo Choi

In carbon steel pipes of nuclear power plants, local wall thinning may result from erosion-corrosion or flow-accelerated corrosion(FAC) damage. Local wall thinning is one of the major causes for the structural fracture of these pipes. Therefore, assessment of local wall thinning due to corrosion is an important issue in nondestructive evaluation for the integrity of nuclear power plants. In this study, laser-generated ultrasound technique was employed to evaluate local wall thinning due to corrosion. Guided waves were generated in the thermoelastic regime using a Q-switched pulsed Nd:YAG laser with a linear slit array. . In this paper, time-frequency analysis of ultrasonic waveforms using wavelet transform allowed the identification of generated guided wave modes by comparison with the theoretical dispersion curves. Modes conversion and group velocity were employed to detect thickness reduction.


2000 ◽  
Vol 2000 (0) ◽  
pp. 441-442
Author(s):  
Katsumi HOSAKA ◽  
Choji ARATA ◽  
Hiroshi UEDA ◽  
Yasuhide ASADA

Author(s):  
Francesco Bertoncini ◽  
Mauro Cappelli ◽  
Francesco Cordella ◽  
Marco Raugi

On-line monitoring for installed piping in Nuclear Power Plants (NPPs), as well as for Oil & Gas and other kind of plants, is crucial to early detect local ageing effects and locate single defects before they may result in critical failures. All the actions able to prevent failures are of great value especially if non-invasive and allowing an In-Service Inspection (ISI). In particular the Long Term Operation (LTO) and Plant Life Extension (PLEX) may be invalidated from radiation, thermal, mechanical stresses besides their own ageing. Hence on-line monitoring techniques are of much interest especially if they assure the required safety levels and at the same time are simple and cost-effective. Guided Waves (GW) satisfy these requirements since they are structure-borne ultrasonic waves that propagate themselves without interfering along the same pipe structure, which in turns through its geometric boundaries serves as a confining structure for the GW used to test its integrity. The frequencies used for GW testing extend up to 250 kHz, thus allowing a long-range inspection of pipes (tens of meters in favorable circumstances). The experimental conditions (e.g. temperature, complex piping structure, wall thickness, materials) have to be considered since they strongly affect the results but GW generated through magnetostrictive sensors are expected to overcome such issues due to their robustness and positioning ease. In this paper, new experimental tests conducted using the proposed methodology for steel pipes having different types of structural complexity are described.


1994 ◽  
Vol 151 (2-3) ◽  
pp. 539-550 ◽  
Author(s):  
Ludwig von Bernus ◽  
Werner Rathgeb ◽  
Rudi Schmid ◽  
Friedrich Mohr ◽  
Michael Kröning

Author(s):  
Wei Tang ◽  
Maxim Gussev ◽  
Zhili Feng ◽  
Brian Gibson ◽  
Roger Miller ◽  
...  

Abstract The mitigation of helium induced cracking in the heat affected zone (HAZ), a transition metallurgical zone between the weld zone and base metal, during repair welding is a great challenge in nuclear industry. Successful traditional fusion welding repairs are limited to metals with a maximum of a couple of atomic parts per million (appm) helium, and structural materials helium levels in operating nuclear power plants are generally exceed a couple of appm after years of operations. Therefore, fusion welding is very limited in nuclear power plants structural materials repairing. Friction stir welding (FSW) is a solid-state joining technology that reduces the drivers (temperature and tensile residual stress) for helium-induced cracking. This paper will detail initial procedural development of FSW weld trials on irradiated 304L stainless steel (304L SS) coupons utilizing a unique welding facility located at one of Oak Ridge National Laboratory’s hot cell facilities. The successful early results of FSW of an irradiated 304L SS coupon containing high helium are discussed. Helium induced cracking was not observed by scanning electron microscopy in the friction stir weld zone and the metallurgical zones between the weld zone and base metal, i.e. thermal mechanical affected zone (TMAZ) and HAZ. Characterization of the weld, TMAZ and HAZ regions are detailed in this paper.


2015 ◽  
Vol 59 (3) ◽  
pp. 91-98
Author(s):  
V. Šefl

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.


Author(s):  
Uijeong Ro ◽  
Jeong Hwan Kim ◽  
Hoomin Lee ◽  
Seok Jun Kang ◽  
Moon Ki Kim

The Sodium Fast-cooled Reactor (SFR), are generation IV nuclear power plants, have a target operating temperature of 550°C which makes creep-fatigue behavior more critical than a generation III nuclear power plants. So it is important to understand the nature of creep-fatigue behavior of the piping material, Grade 91 steel. The creep-fatigue damage diagram of Grade 91 steel used in ASME-NH was derived using a conventional time-fraction testing method which was originally developed for type 300 stainless steels. Multiple studies indicate that the creep-fatigue damage diagram of Grade 91 steel developed using this testing method has excessive conservatism in it. Therefore, an alternative testing method was suggested by separating creep and fatigue using interrupted creep tests. The suggested method makes it possible to control creep life consumption freely which was difficult with the previous method. It also makes it easier to observe the interaction between creep and fatigue mechanisms and microstructural evolution. In conclusion, an alternative creep-fatigue damage diagram for Grade 91 steel at 550°C was developed using an interrupt creep fatigue testing method and FE model simulation.


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