Download or read book Environmentally Assisted Cracking of Light water Reactor Materials written by and published by . This book was released on 1996 with total page 7 pages. Available in PDF, EPUB and Kindle. Book excerpt: Environmentally assisted cracking (EAC) of lightwater reactor (LWR) materials has affected nuclear reactors from the very introduction of the technology. Corrosion problems have afflicted steam generators from the very introduction of pressurized water reactor (PWR) technology. Shippingport, the first commercial PWR operated in the United States, developed leaking cracks in two Type 304 stainless steel (SS) steam generator tubes as early as 1957, after only 150 h of operation. Stress corrosion cracks were observed in the heat-affected zones of welds in austenitic SS piping and associated components in boiling-water reactors (BRWs) as early as 1965. The degradation of steam generator tubing in PWRs and the stress corrosion cracking (SCC) of austenitic SS piping in BWRs have been the most visible and most expensive examples of EAC in LWRs, and the repair and replacement of steam generators and recirculation piping has cost hundreds of millions of dollars. However, other problems associated with the effects of the environment on reactor structures and components am important concerns in operating plants and for extended reactor lifetimes. Cast duplex austenitic-ferritic SSs are used extensively in the nuclear industry to fabricate pump casings and valve bodies for LWRs and primary coolant piping in many PWRs. Embrittlement of the ferrite phase in cast duplex SS may occur after 10 to 20 years at reactor operating temperatures, which could influence the mechanical response and integrity of pressure boundary components during high strain-rate loading (e.g., seismic events). The problem is of most concern in PWRs where slightly higher temperatures are typical and cast SS piping is widely used.

Download or read book Environmentally Assisted Cracking in Light Water Reactors written by W. J. Shack and published by . This book was released on 1983 with total page 108 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Environmentally Assisted Cracking in Light Water Reactors Annual Report January December 2005 written by and published by . This book was released on 2007 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: This report summarizes work performed from January to December 2005 by Argonne National Laboratory on fatigue and environmentally assisted cracking in light water reactors (LWRs). Existing statistical models for estimating the fatigue life of carbon and low-alloy steels and austenitic stainless steels (SSs) as a function of material, loading, and environmental conditions were updated. Also, the ASME Code fatigue adjustment factors of 2 on stress and 20 on life were critically reviewed to assess the possible conservatism in the current choice of the margins. An approach, based on an environmental fatigue correction factor, for incorporating the effects of LWR environments into ASME Section III fatigue evaluations is discussed. The susceptibility of austenitic stainless steels and their welds to irradiation-assisted stress corrosion cracking (IASCC) is being evaluated as a function of the fluence level, water chemistry, material chemistry, and fabrication history. For this task, crack growth rate (CGR) tests and slow strain rate tensile (SSRT) tests are being conducted on various austenitic SSs irradiated in the Halden boiling water reactor. The SSRT tests are currently focused on investigating the effects of the grain boundary engineering process on the IASCC of the austenitic SSs. The CGR tests were conducted on Type 316 SSs irradiated to 0.45-3.0 dpa, and on sensitized Type 304 SS and SS weld heat-affected-zone material irradiated to 2.16 dpa. The CGR tests on materials irradiated to 2.16 dpa were followed by a fracture toughness test in a water environment. The effects of material composition, irradiation, and water chemistry on growth rates are discussed. The susceptibility of austenitic SS core internals to IASCC and void swelling is also being evaluated for pressurized water reactors. Both SSRT tests and microstructural examinations are being conducted on specimens irradiated in the BOR-60 reactor in Russia to doses up to 20 dpa. Crack growth rate data, obtained in the pressurized water reactor environment, are presented on Ni-alloy welds prepared in the laboratory or obtained from the nozzle-to-pipe weld of the V.C. Summer reactor. The experimental CGRs under cyclic and constant load are compared with the existing CGR data for Ni-alloy welds to determine the relative susceptibility of these materials to environmentally enhanced cracking under a variety of loading conditions.

Download or read book Environmentally Assisted Cracking in Light Water Reactors written by and published by . This book was released on 1982 with total page 222 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Environmentally assisted Cracking in Austenitic Light Water Reactor Structural Materials written by Hans-Peter Seifert and published by . This book was released on 2012 with total page 139 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Environmentally Assisted Cracking in Light Water Reactors written by W. J. Shack and published by . This book was released on 1984 with total page 148 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Environmentally Assisted Cracking in Light Water Reactors Semiannual Report October 1993 March 1994 written by and published by . This book was released on 1995 with total page 51 pages. Available in PDF, EPUB and Kindle. Book excerpt: This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated pressurized water reactor water are valid for high-sulfur heats that show environmentally enhanced fatigue crack growth rates. Additional crack growth data were obtained on fracture-mechanics specimens of austenitic SSs to investigate threshold stress intensity factors for EAC in high-purity oxygenated water at 289°C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating boiling water reactors were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements, which are not specified in the ASTM specifications, may contribute to IASCC of solution-annealed materials.

Download or read book Environmentally Assisted Cracking in Light Water Reactors Semiannual Report April 1994 September 1994 written by and published by . This book was released on 1995 with total page 53 pages. Available in PDF, EPUB and Kindle. Book excerpt: This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors from April to September 1994. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in piping and reactor pressure vessels, (b) EAC of austenitic stainless steels (SSs) and Alloy 600, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests have been conducted on A106-Gr B and A533-Gr B steels in oxygenated water to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack growth data were obtained on fracture-mechanics specimens of SSs and Alloy 600 to investigate EAC in simulated boiling water reactor (BWR) and pressurized water reactor environments at 289°C. The data were compared with predictions from crack growth correlations developed at ANL for SSs in water and from rates in air from Section XI of the ASME Code. Microchemical changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

Download or read book Environmentally Assisted Cracking of LWR Materials written by and published by . This book was released on 1995 with total page 16 pages. Available in PDF, EPUB and Kindle. Book excerpt: Research on environmentally assisted cracking (EAC) of light water reactor materials has focused on (a) fatigue initiation in pressure vessel and piping steels, (b) crack growth in cast duplex and austenitic stainless steels (SSs), (c) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs, and (d) EAC in high- nickel alloys. The effect of strain rate during different portions of the loading cycle on fatigue life of carbon and low-alloy steels in 289°C water was determined. Crack growth studies on wrought and cast SSs have been completed. The effect of dissolved-oxygen concentration in high-purity water on IASCC of irradiated Type 304 SS was investigated and trace elements in the steel that increase susceptibility to intergranular cracking were identified. Preliminary results were obtained on crack growth rates of high-nickel alloys in water that contains a wide range of dissolved oxygen and hydrogen concentrations at 289 and 320°C. The program on Environmentally Assisted Cracking of Light Water Reactor Materials is currently focused on four tasks: fatigue initiation in pressure vessel and piping steels, fatigue and environmentally assisted crack growth in cast duplex and austenitic SS, irradiation-assisted stress corrosion cracking of austenitic SSs, and environmentally assisted crack growth in high-nickel alloys. Measurements of corrosion-fatigue crack growth rates (CGRs) of wrought and cast stainless steels has been essentially completed. Recent progress in these areas is outlined in the following sections.

Download or read book Environmentally assisted Cracking in Austenitic Light Water Reactor Structural Materials written by Hans-Peter Seifert and published by . This book was released on 2009 with total page 105 pages. Available in PDF, EPUB and Kindle. Book excerpt:

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 Environmentally assisted Cracking in Austenitic Light Water Reactor Structural Materials written by H. P. Seifert and published by . This book was released on 2012 with total page 139 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Environmentally Assisted Cracking in Light Water Reactors Annual Report January December 2001 written by R. W. Clark and published by . This book was released on 2003 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 January to December 2001. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs, and (c) EAC of Alloy 600. The effects of key material and loading variables, such as strain amplitude, strain rate, temperature, dissolved oxygen (DO) level in water, and material heat treatment, on the fatigue lives of wrought and cast austenitic SSs in air and LWR environments have been evaluated. The mechanism of fatigue crack initiation in austenitic SSs in LWR environments has also been examined. The results indicate that the presence of a surface oxide film or difference in the characteristics of the oxide film has no effect on fatigue crack initiation in austenitic SSs in LWR environments. Slow-strain-rate tensile tests and post-test fractographic analyses were conducted on several model SS alloys irradiated to {approx}2 x 10{sup 21} n {center_dot} cm{sup -2} (E> 1 MeV) ({approx}3 dpa) 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. Corrosion fatigue tests were conducted on nonirradiated austenitic SSs in high-purity water at 289 C to establish the test procedure and conditions that will be used for the tests on irradiated materials. A comprehensive irradiation experiment was initiated to obtain many tensile and disk specimens irradiated under simulated pressurized water reactor conditions at {approx}325 C to 5, 10, 20, and 40 dpa. Crack growth tests were completed on 30% cold-worked Alloy 600 in high-purity water under various environmental and loading conditions. The results are compared with data obtained earlier on several heats of Alloy 600 tested in high-DO water under several heat treatment conditions.

Download or read book Environmentally Assisted Cracking in Light Water Reactors written by W. J. Shack and published by . This book was released on 1986 with total page 4 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Environmentally Assisted Cracking in Light Water Reactors written by and published by . This book was released on 1996 with total page 87 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 April 1995 to December 1995. Topics that have been investigated include fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, EAC of Alloy 600 and 690, and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in simulated LWR environments. Effects of fluoride-ion contamination on susceptibility to intergranular cracking of high- and commercial- purity Type 304 SS specimens from control-tensile tests at 288 degrees Centigrade. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

Download or read book Environmental assisted Cracking in Austenitic Light Water Reactor Structural Materials written by Hans-Peter Seifert and published by . This book was released on 2009 with total page 105 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Environmentally Assisted Cracking of Structural Materials for Light Water Reactors written by Norio Nagata and published by . This book was released on 1994 with total page 31 pages. Available in PDF, EPUB and Kindle. Book excerpt: