Investigation of the processes of fatigue and corrosion-fatigue destruction of pseudo-α titanium alloy

2021 ◽  
Vol 7 ◽  
pp. 37-48
Author(s):  
A. A. Murashov ◽  
◽  
N. N. Berendeyev ◽  
A. V. Nokhrin ◽  
E. A. Galaeva ◽  
...  
Keyword(s):  
Titanium Alloy ◽  
Corrosion Fatigue ◽  
Nuclear Power ◽  
Power Plants ◽  
Fatigue Cracks ◽  
Cyclic Fatigue ◽  
Fatigue Tests ◽  
Corrosive Environment ◽  
Β Phase ◽  
Loading Scheme

The paper describes the results of fatigue and corrosion-fatigue tests of the pseudo α titanium alloy PT-3V, which is actively used in nuclear engineering for the manufacture of heat exchange equipment for modern nuclear power plants. Alloy PT-3V has an inhomogeneous coarse-crystalline structure with precipitates of β-phase particles along the grain boundaries of the lamellar shape. It is shown that smooth specimens tested according to the bending-with-rotation loading scheme do not show a noticeable decrease in the cyclic fatigue life when exposed to a neutral corrosive environment (3 % aqueous NaCl solution). However, specimens with a notch (stress concentrator) tested according to the cantilever bending loading scheme demonstrate sensitivity to the action of a corrosive environment at the stages of initiation and growth of fatigue cracks, as evidenced by a significant decrease in the number of cycles before crack initiation, as well as before specimen failure, in comparison with tests in air. Fractographic analysis of fractures of smooth specimens and specimens with a concentrator after fatigue and corrosion-fatigue tests has been carried out. The main stages of the initiation and growth of fatigue cracks are revealed. It has been established that a decrease in the resistance to the initiation and propagation of corrosion-fatigue cracks during testing of notched specimens may be due to the effect of hydrogen embrittlement, accelerated by stress concentration.

scholarly journals Interaction of Cyclic Loading (Low-Cyclic Fatigue) with Stress Corrosion Cracking (SCC) Growth Rate

2020 ◽  
Vol 2020 ◽  
pp. 1-10
Author(s):  
Rehmat Bashir ◽  
He Xue ◽  
Rui Guo ◽  
Yueqi Bi ◽  
Muhammad Usman
Keyword(s):  
Growth Rate ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Cyclic Fatigue ◽  
Joint Effect ◽  
Corrosion Cracking

The structural integrity analysis of nuclear power plants (NPPs) is an essential procedure since the age of NPPs is increasing constantly while the number of new NPPs is still limited. Low-cyclic fatigue (LCF) and stress corrosion cracking (SSC) are the two main causes of failure in light-water reactors (LWRs). In the last few decades, many types of research studies have been conducted on these two phenomena separately, but the joint effect of these two mechanisms on the same crack has not been discussed yet though these two loads exist simultaneously in the LWRs. SCC is mainly a combination of the loading, the corrosive medium, and the susceptibility of materials while the LCF depends upon the elements such as compression, moisture, contact, and weld. As it is an attempt to combine SCC and LCF, this research focuses on the joint effect of SCC and LCF loading on crack propagation. The simulations are carried out using extended finite element method (XFEM) separately, for the SCC and LCF, on an identical crack. In the case of SCC, da/dt(mm/sec) is converted into da/dNScc (mm/cycle), and results are combined at the end. It has been observed that the separately calculated results for SCC da/dNScc and LCF da/dNm of crack growth rate are different from those of joint/overall effect,  da/dNom. By applying different SCC loads, the overall crack growth is measured as SCC load becomes the main cause of failure in LWRs in some cases particularly in the presence of residual stresses.

Environmental Fatigue Test of CF8M With a Small Autoclave Simulating PWR Conditions

2008 ◽  
Author(s):  
Il-Seok Jeong ◽  
Gag-Hyeon Ha ◽  
Tae-Ryoung Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Low Cycle Fatigue ◽  
Test Procedure ◽  
Cyclic Strain ◽  
Fatigue Tests ◽  
Operating Conditions ◽  
Measuring Instruments ◽  
Pressurized Water Reactor ◽  
Pressurized Water

To develop a fatigue design curve of cast stainless steel CF8M used in primary piping material of nuclear power plants, low-cycle fatigue tests have been conducted by Korea Electric Power Research Institute (KEPRI). A small autoclave simulated the environment of a pressurized water reactor (PWR), 15 MPa and 315 °C. Fatigue life was measured in terms of the number of cycles with the variation of strain amplitudes at 0.04%/s strain rate. A small autoclave of 1 liter and cylindrical solid fatigue specimens were used for the strain-controlled low cycle environmental fatigue tests to make the experiments convenient. However, it was difficult to install displacement measuring instruments at the target length of the specimens inside the autoclave. To mitigate the difficulty displacement data measured at the shoulders of the specimen were calibrated based on the data relation of the target and shoulder length of the specimen during hot air test conditions. KEPRI developed a test procedure to perform low cycle environmental fatigue tests in the small autoclave. The procedure corrects the cyclic strain hardening effect by performing additional tests in high temperature air condition. KEPRI verified that the corrected test result agreed well with that of finite element method analysis. The process of correcting environmental fatigue data would be useful for producing reliable fatigue curves using a small autoclave simulating the operating conditions of a PWR.

Comparison of Strain Range Measures and Environmental Fatigue Calculation Methodologies for the Stepped Pipe Tests

2008 ◽  
Author(s):  
Robert Gurdal ◽  
Steven X. Xu
Keyword(s):  
Finite Element ◽  
Environmental Effects ◽  
Nuclear Power ◽  
Power Plants ◽  
Strain Range ◽  
The United States ◽  
Fatigue Tests ◽  
Finite Element Analyses ◽  
Strain Measure ◽  
Component Design

Various strain measure formulas exist at this time for the calculation of the strain amplitude required for fatigue calculations, and various methodologies have been suggested in the years 2005 through 2008 to take into account — in general — the environmental effects on fatigue (EAF = Environmentally-Assisted Fatigue). The purpose of this technical paper is to compare these strain measure formulas and these EAF methodologies for the case of the thermal fatigue tests of a stainless steel stepped pipe for which results have been published in the Proceedings of the 2004 PVP Conference [1]. Thermal transient finite element analyses and cyclic elasto-plastic finite element analyses were performed to obtain the thermal gradients through the pipe thickness and the resulting strain ranges. These strain ranges are based on the various strain measure definitions presented at the 2001/2005 PVP Conferences (see Ref. [2] and [3]). These various strain measure definitions were compared. Using one of the stepped pipe inside surface locations and using one of the strain range values (out of the various strain measure definitions), the allowable number of design cycles has been calculated, based on the currently mandated methodologies for the environmental effects on fatigue (EAF). These methodologies are the EAF methodologies to be applied in the United States for future fatigue calculations, either for license renewal of the currently operating nuclear power plants or for the design of new plants. The fatigue results are compared and discussed for their implication in component design.

Evaluation of Low-Cycle Fatigue in Simulated PWR Environment

2006 ◽  
Vol 326-328 ◽  
pp. 1011-1014 ◽  
Author(s):  
Ill Seok Jeong ◽  
Sang Jai Kim ◽  
Taek Ho Song ◽  
Sung Yull Hong
Keyword(s):  
Fatigue Life ◽  
Nuclear Power ◽  
Power Plants ◽  
Low Cycle Fatigue ◽  
Water Environment ◽  
Fatigue Tests ◽  
Cycle Fatigue ◽  
Fatigue Design ◽  
Number Of Cycles ◽  
Measured Length

For developing fatigue design curve of cast stainless steel that is used in piping material of nuclear power plants, a low-cycle fatigue test rig was built. It is capable of performing tests in pressurized high temperature water environment of PWR. Cylindrical solid fatigue specimens of CF8M were used for the strain-controlled environmental fatigue tests. Fatigue life was measured in terms of the number of cycles with the variation of strain amplitude at 0.04%/s strain rates. The disparity between target length and measured length of specimens was corrected by using finite element method. The corrected test results showed similar fatigue life trend with other previous results.

Application of Ultrasonic Phased Array Techniques for Inspection of Stud Bolts in Nuclear Power Plants

2006 ◽  
Vol 110 ◽  
pp. 97-104 ◽  
Author(s):  
Sang Woo Choi ◽  
Joon Hyun Lee
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Phased Array ◽  
Nuclear Reactor ◽  
Fatigue Cracks ◽  
Ultrasonic Inspection ◽  
Phased Array Techniques ◽  
Ultrasonic Phased Array ◽  
Ultrasonic Images

The reactor vessel body and closure head are fastened with the stud bolt that is one of crucial parts for safety of the reactor vessels in nuclear power plants. It is reported that the stud bolt is often experienced by fatigue cracks initiated at threads. Stud bolts are inspected by the ultrasonic technique during the overhaul periodically for the prevention of failure which leads to radioactive leakage from the nuclear reactor. The conventional ultrasonic inspection for stud bolts was mainly conducted by reflected echo method based on shadow effect. However, in this technique, there were numerous spurious signals reflected from every oblique surfaces of the thread. In this study, ultrasonic phased array technique was applied to investigate detectability of flaws in stud bolts and characteristics of ultrasonic images corresponding to different scanning methods, that is, sector and linear scan. For this purpose, simplified stud bolt specimens with artificial defects of various depths were prepared.

Investigation of Effect of Pre-Strain on Very High-Cycle Fatigue Strength of Austenitic Stainless Steels

2008 ◽  
Author(s):  
Takeshi Ogawa ◽  
Motoki Nakane ◽  
Kiyotaka Masaki ◽  
Shota Hashimoto ◽  
Yasuo Ochi ◽  
...  
Keyword(s):  
Fatigue Strength ◽  
Nuclear Power ◽  
Power Plants ◽  
Austenitic Stainless Steels ◽  
Fatigue Tests ◽  
Structural Components ◽  
Very High Cycle Fatigue ◽  
Tension And Compression ◽  
Strained Materials ◽  
Very High

The austenitic stainless steels have excellent mechanical and chemical characteristics and these materials are widely used for the main structural components in the nuclear power plants. A part of structural components using these materials is considered to have strain-history by machining, welding and etc in the process of manufacturing and these parts would be hardened because these materials have a remarkable work-hardening property. On the other hand, conventional studies for the fatigue strength used to be investigated by the results of fatigue tests applying normal specimens without the effect of hardening by pre-strain. This paper describes the effect of large pre-strain on very high cycle fatigue strength of the materials in consideration for the evaluation of strength of actual structures in the nuclear power plants. In order to achieve this purpose, the fatigue tests were carried out with strain hardened specimens. The material served in this study was type SUS316NG. Up to ±20% pre-strain was introduced to the round bar shaped materials by tension and compression load test, and the materials were mechanically machined to the hourglass shaped smooth specimens. On the other hand, the pre-strain of some specimens were introduced after machining so as to study the influence of roughness of the surface of the specimens for the fatigue property. Fatigue tests were conducted by ultrasonic and rotating-bending fatigue test machines and conditions were decided by preliminary examinations to control temperature elevation of the specimen during the fatigue test. The S-N curves obtained from fatigue tests show that increase in magnitude of the pre-strain cause increase in the fatigue strength of the material and this relationship is independent of type of the pre-strains of tension and compression. Though all specimens were fractured by the surface initiated fatigue crack, only one specimen was fractured by the internal crack and so-called “fish-eye” was observed on the fracture surface. However, the internal fracture of the SUS316NG does not cause sudden drop of the fatigue strength. Also, the Vickers hardness tests were carried out to discuss the relationship between fatigue strength and hardness of the pre-strained materials. It is found that the increase in fatigue limit of the pre-strained materials strongly depend on the hardness derived from the indentation size equals to the scale of stage I fatigue crack.

Stress Corrosion Cracking and Corrosion Fatigue Analysis of Nuclear Steam Turbine Rotor

2014 ◽  
Author(s):  
Gang Chen ◽  
Puning Jiang ◽  
Xingzhu Ye ◽  
Junhui Zhang ◽  
Yifeng Hu ◽  
...  
Keyword(s):  
Stress Corrosion ◽  
Steam Turbine ◽  
Corrosion Fatigue ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Fatigue Cracking ◽  
Corrosion Cracking ◽  
Major Influence ◽  
Steam Turbines

Although stress corrosion cracking (SCC) and corrosion fatigue cracking can occur in many locations of nuclear steam turbines, most of them initiate at low pressure disc rim, rotor groove and keyway of the shrunk-on disc. For nuclear steam turbine components, long life endurance and high availability are very important factors in the operation. Usually nuclear power plants operating more than sixty years are susceptible to this failure mechanism. If SCC or corrosion fatigue happens, especially in rotor groove or keyway, it has a major influence on nuclear steam turbine life. In this paper, established methods for the SCC and corrosion fatigue-controlled life prediction of steam turbine components were applied to evaluating a new shrunk-on disc that had suffered local keyway surface damage during manufacture and loss of residual compressive stress.

Improving Ultrasonic Examination Procedures for Detection of Thermal Fatigue

2017 ◽  
Author(s):  
Jack Spanner
Keyword(s):  
Thermal Fatigue ◽  
Nuclear Power ◽  
Power Plants ◽  
Fatigue Cracks ◽  
Computer Based Training ◽  
Ultrasonic Examination ◽  
Nuclear Plants ◽  
The Usa ◽  
Computer Based ◽  
Thermal Fatigue Cracks

This paper describes improvements to the ultrasonic procedures to be used for the detection of thermal fatigue in nuclear power plants in accordance with the requirements of the Electric Power Research Institute (EPRI) Material Reliability Program (MRP) inspection and evaluation guidelines. These examinations have been performed at nuclear plants in the USA since the 1980s with very few detections of degradation. However, since 2013 there have been ten instances of thermal fatigue cracks. The MRP formed a thermal fatigue focus group to analyze these leaks and flaws related to thermal fatigue inspection programs. Then the group developed recommendations to address these recent operational experiences. The MRP has been developing improvements to the ultrasonic examination process and this paper will share these. A computer based training program for the ultrasonic personnel has been developed that will be described. And finally, the MRP has fabricated a variety of thermal fatigue mockups that are loaned to member utilities prior to an outage so the ultrasonic personnel can practice detecting thermal fatigue just prior to the examinations. Implementation of these mockups will also be described.

A STUDY ON THE FRETTING FATIGUE LIFE OF ZIRCALOY ALLOYS

2008 ◽  
Vol 22 (11) ◽  
pp. 851-856
Author(s):  
JAE-DO KWON ◽  
DAE-KYU PARK ◽  
SEUNG-WAN WOO ◽  
YOUNG-SUCK CHAI
Keyword(s):  
Fatigue Life ◽  
Fatigue Strength ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Fatigue Behavior ◽  
Fretting Fatigue ◽  
Mechanical Tests ◽  
Fatigue Tests ◽  
Cyclic Damage

Studies on the strength and fatigue life of machines and structures have been conducted in accordance with the development of modern industries. In particular, fine and repetitive cyclic damage occurring in contact regions has been known to have an impact on fretting fatigue fractures. The main component of zircaloy alloy is Zr , and it possesses good mechanical characteristics at high temperatures. This alloy is used in the fuel rod material of nuclear power plants because of its excellent resistance. In this paper, the effect of the fretting damage on the fatigue behavior of the zircaloy alloy is studied. Further, various types of mechanical tests such as tension and plain fatigue tests are performed. Fretting fatigue tests are performed with a flat-flat contact configuration using a bridge-type contact pad and plate-type specimen. Through these experiments, it is found that the fretting fatigue strength decreases by about 80% as compared to the plain fatigue strength. Oblique cracks are observed in the initial stage of the fretting fatigue, in which damaged areas are found. These results can be used as the basic data for the structural integrity evaluation of corrosion-resisting alloys considering the fretting damages.

Initiation and Growth of Small Fatigue Cracks of Steels Used for Nuclear Power Plants Under Low Cycle Regime

2014 ◽  
Author(s):  
Shota Hasunuma ◽  
Takeshi Ogawa
Keyword(s):  
Fatigue Crack ◽  
Fracture Mechanics ◽  
Crack Initiation ◽  
Crack Growth ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Low Cycle Fatigue ◽  
Fatigue Cracks ◽  
Cycle Fatigue

Low cycle fatigue tests were conducted for carbon steel, STS410, low alloy steel, SFVQ1A, and austenitic stainless steel, SUS316NG, which were used for nuclear power plants, in order to investigate the mechanism of fatigue damage when the plants were subjected to huge seismic loads. In these tests, the surface behavior of fatigue crack initiation and growth was observed in detail using cellulose acetate replicas, while the interior behavior was detected in terms of fracture surface morphology developed by multiple two-step strain amplitude variations with periodical surface removals. Fatigue crack growth rates were evaluated by elasto-plastic fracture mechanics approach. For SFVQ1A and SUS316NG, the fracture mechanics approach is available in order to predict the crack growth life from the metallurgical crack initiation size to the final crack length of the specimens. For STS410, numerous small cracks initiated, grew and coalesced each other on the specimen surface under low cycle fatigue regime.

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