Download or read book Fatigue Crack Growth Behavior of Different Stainless Steels in Pressurized Water Reactor Environments written by C. Amzallag and published by . This book was released on 1990 with total page 32 pages. Available in PDF, EPUB and Kindle. Book excerpt: An experimental program has been conducted in order to determine the fatigue crack growth rate (FCGR) curves of different stainless steels for nuclear pressure vessels and pipings in pressurized water reactor environments and to define reference fatigue crack growth rate curves for these materials. The parameters studied are: the steel structure, the load ratio (R), and the water chemistry. A slight, but clear effect of environment was observed for the five steels studied. The paper presents the FCGR curves and the fractographic examinations for the different testing conditions and stresses the differences between FCGR behavior of stainless steels and low-alloy steels.
Download or read book A Review of Fatigue Crack Growth of Pressure Vessel and Piping Steels in High temperature Pressurized Reactor grade Water written by W. H. Cullen and published by . This book was released on 1980 with total page 134 pages. Available in PDF, EPUB and Kindle. Book excerpt: Fatigue crack growth data sets, for pressure vessel and piping steels, in reactor-grade water environment have appeared in various reports and publications since about 1972. All of the results which have been published from 1972 through 1979 have been plotted and are presented in this report. Beginning with a discussion of the need for these data, and an explanation of the laboratory facilities which are required for this research, this report goes on to describe the overall trends which have evolved through consideration of the data sets and the conditions under which they were generated. A model for hydrogen assisted fatigue crack growth is described and applied to the pressurized water reactor type of environment. A complete listing of references is included in the report. (Author).
Download or read book Review of Environmental Effects on Fatigue Crack Growth of Austenitic Stainless Steels written by and published by . This book was released on 1994 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Fatigue and environmentally assisted cracking of piping, pressure vessel cladding, and core components in light water reactors are potential concerns to the nuclear industry and regulatory agencies. The degradation processes include intergranular stress corrosion cracking of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or stress corrosion cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Crack growth data for wrought and cast austenitic SSs in simulated BWR water, developed at Argonne National Laboratory under US Nuclear Regulatory Commission sponsorship over the past 10 years, have been compiled into a data base along with similar data obtained from the open literature. The data were analyzed to develop corrosion-fatigue curves for austenitic SSs in aqueous environments corresponding to normal BWR water chemistries, for BWRs that add hydrogen to the feedwater, and for pressurized water reactor primary-system-coolant chemistry. The corrosion-fatigue data and curves in water were compared with the air line in Section XI of the ASME Code.
Download or read book Mechanics of Fatigue Crack Closure written by Wolf Elber and published by ASTM International. This book was released on 1988 with total page 671 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Some Mechanistic Observations on the Crack Growth Characteristics of Pressure Vessel and Piping Steels in PWR Environment written by and published by . This book was released on 1979 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The fatigue crack growth behavior of A533B and A508 pressure vessel steel and AISI Types 304 and 316 steels used in reactor coolant piping have been studied in a pressurized water reactor environment at 288/sup 0/C (550/sup 0/F). The influence of stress ratio (P/sub min//P/sub max/), frequency, ramp times, specimen orientation and material microstructures were included in the study. While none of the materials showed evidence of static crack growth in the environment, the ferritic steels did show an enhanced fatigue crack growth rate at test frequencies of five cycles per minute and lower. Based on fractographic examinations the enhanced growth rate is not the result of environmentally induced intergranular or cleavage modes of crack propagation. Instead, striation spacing measurements were found to agree with the macroscopic crack growth rate, demonstrating a time dependent environmental interaction which introduces a frequency dependent enhancement of the mechanically developed striations. Crack growth experiments using hold times have confirmed the absence of any superimposed contribution of static crack growth components. Fatigue crack growth tests were conducted in an environment of Hydrogen Sulfide gas to establish the contribution of hydrogen embrittlement and will also be described.
Download or read book Fatigue Crack Growth Rates of A 508 2 Steel in Pressurized High Temperature Water written by W. H. Cullen and published by . This book was released on 1983 with total page 40 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Environmentally Assisted Cracking written by W. Barry Lisagor and published by ASTM International. This book was released on 1990 with total page 540 pages. Available in PDF, EPUB and Kindle. Book excerpt: Papers included topics of phenomena, basic mechanisms, modeling, test methodologies, materials performance, engineering applications and service experience and failures and reflects the current emphasis with regard to material/environment systems.
Download or read book Fatigue Crack Growth of A508 Steel in High temperature Pressurized Reactor Grade Water written by W. H. Cullen and published by . This book was released on 1979 with total page 64 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Fatigue Crack Growth Rates in Pressure Vessel and Piping Steels in LWR Environments written by W. H. Cullen and published by . This book was released on 1987 with total page 80 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Effect of Light Water Reactor Environments on Fatigue Crack Growth Rate in Reactor Pressure Vessel Steels written by WH. Cullen and published by . This book was released on 1982 with total page 22 pages. Available in PDF, EPUB and Kindle. Book excerpt: This paper presents an overview of the effect of light-water reactor environments on the fatigue crack growth rate in reactor pressure vessel steels. The effect of different variables known to influence the environmental assistance--stress intensity range, load ratio, frequency, temperature, water chemistry, and irradiation--is described. The fractographic phenomena and mechanisms of fatigue crack propagation associated with environmental influence is emphasized. An analytical approach for predicting the crack growth rates based on the hydrogen embrittlement model is described. Finally, a discussion of the effect of different transients occurring in the nuclear reactors on the crack growth rate predictions is presented.
Download or read book The Effect of Environment Chemistry and Microstructure on the Corrosion Fatigue Behavior of Austenitic Stainless Steels in High Temperature Water written by Lindsay Beth O'Brien and published by . This book was released on 2014 with total page 122 pages. Available in PDF, EPUB and Kindle. Book excerpt: The effect of sulfur on the corrosion fatigue crack growth of austenitic stainless steel was evaluated under Light Water Reactor (LWR) conditions of 288°C deaerated (less than 5ppb O2) water, to shed light on the accelerating effect of the LWR environment and to explore the effect of high sulfur content on the retardation of fatigue crack growth rates. Fatigue tests were performed using a trapezoidal loading pattern with rise times of 5.1, 51, 510, and 5100 seconds (fall time of 0.9, 9, 90, and 900 seconds), with Kmzx of 28.6 or 31.9 MPa[mathematical symbol]m and stress ratios (R, Pmin/Pmax) of 0.4 or 0.7. Two test materials were used to evaluate the effect of sulfur: (1) a low sulfur (
Download or read book Fatigue and Environmentally Assisted Cracking in Light Water Reactors written by and published by . This book was released on 1992 with total page 22 pages. Available in PDF, EPUB and Kindle. Book excerpt: Fatigue and stress corrosion cracking (SCC) for low-alloy steel used in piping and in steam generator and reactor pressure vessels have been investigated. Fatigue data were obtained on medium-sulfur-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor water, and in air. Analytical studies focused on the behavior of carbon steels in boiling water reactor (BWR) environments. Crack-growth rates of composite fracture-mechanics specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B steel were determined under small-amplitude cyclic loading in HP water with ≈300 pbb dissolved oxygen. Radiation-induced segregation and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence also have been investigated. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain-rate tensile tests were conducted on tubular specimens in air and in simulated BWR water at 289°C.
Download or read book Environmental Acceleration of Fatigue Crack Growth in Reactor Pressure Vessel Materials and Environments written by RH. Emanuelson and published by . This book was released on 1990 with total page 19 pages. Available in PDF, EPUB and Kindle. Book excerpt: This paper presents the results of experiments to determine which variables have the most effect on environmentally assisted cracking (EAC) of reactor pressure vessel steels. Four heats of SA508 and three heats of SA533 were studied in 288°C pressurized water reactor, light water reactor, and boiling water reactor environments. The sulfur content of these materials ranged from 0.004 to 0.025%. The most important variable was found to be the sulfur content and sulfur morphology of the material. A material with a bulk sulfur content greater than 0.010% may be susceptible to EAC. The loading frequency was found to be an important variable. Minor effects of EAC were attributed to the load ratio, ?K level, and oxygen content of the environment. An experimental method to introduce sulfur directly to the crack tip is described. A time-based method of data interpretation is discussed.
Download or read book Environmentally Assisted Cracking in Light Water Reactors written by and published by . This book was released on 2002 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from July 2000 to December 2000. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. The fatigue strain-vs.-life data are summarized for the effects of various material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs. Effects of the reactor coolant environment on the mechanism of fatigue crack initiation are discussed. Two methods for incorporating the effects of LWR coolant environments into the ASME Code fatigue evaluations are presented. Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to (almost equal to)0.9 x 1021 n · cm−2 (E> 1 MeV) in He at 289 C in the Halden reactor. The results were used to determine the influence of alloying and impurity elements on the susceptibility of these steels to IASCC. A fracture toughness J-R curve test was conducted on a commercial heat of Type 304 SS that was irradiated to (almost equal to)2.0 x 1021 n · cm−2 in the Halden reactor. The results were compared with the data obtained earlier on steels irradiated to 0.3 and 0.9 x 1021 n · cm−2 (E> 1 MeV) (0.45 and 1.35 dpa). Neutron irradiation at 288 C was found to decrease the fracture toughness of austenitic SSs. Tests were conducted on compact-tension specimens of Alloy 600 under cyclic loading to evaluate the enhancement of crack growth rates in LWR environments. Then, the existing fatigue crack growth data on Alloys 600 and 690 were analyzed to establish the effects of temperature, load ratio, frequency, and stress intensity range on crack growth rates in air.
Download or read book Fatigue Life and Crack Growth Predictions of Irradiated Stainless Steels written by Robert William Fuller and published by . This book was released on 2018 with total page 156 pages. Available in PDF, EPUB and Kindle. Book excerpt: One of prominent issues related to failures in nuclear power components is attributed to material degradation due the aggressive environment conditions, and mechanical stresses. For instance, reactor core support components, such as fuel claddings, are under prolonged exposure to an intense neutron field from the fission of fuel and operate at elevated temperature under fatigue loadings caused by start-up, shut-down, and unscheduled emergency shut-down. Additionally, exposure to high-fluence neutron radiation can lead to microscopic defects that result in material hardening and embrittlement, which significantly affects the physical and mechanical properties of the materials, resulting in further reduction in fatigue life of reactor structural components. The effects of fatigue damage on material deterioration can be further exacerbated by the presence of thermal loading, hold-time, and high-temperature water coolant environments. In this study, uniaxial fatigue models were used to predict fatigue behavior based only on simple monotonic properties including ultimate tensile strength and Brinell hardness. Two existing models, the Bäumel-Seeger uniform material law and the Roessle-Fatemi hardness method, were employed and extended to include the effects of test temperature, neutron irradiation fluence, irradiation-induced helium and irradiation-induced swellings on fatigue life of austenitic stainless steels. Furthermore, a methodology to estimate fatigue crack length using a strip-yield based model is presented. This methodology is also extended to address the effect of creep deformation in a presence of hold- times, and expanded to include the effects of irradiation and water environment. Reasonable fatigue life predictions and crack growth estimations are obtained for irradiated austenitic stainless steels types 304, 304L, and 316, when compared to the experimental data available in the literature. Lastly, a failure analysis methodology of a mixer unit shaft made of AISI 304 stainless steel is also presented using a conventional 14-step failure analysis approach. The primary mode of failure is identified to be intergranular stress cracking at the heat affected zones. A means of circumventing this type of failure in the future is presented.
Download or read book Fatigue and Corrosion in Metals written by Pietro Paolo Milella and published by Springer Nature. This book was released on with total page 965 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Energy Research Abstracts written by and published by . This book was released on 1988 with total page 312 pages. Available in PDF, EPUB and Kindle. Book excerpt:
