Download or read book Computer Simulation of Stress Corrosion Cracking Behavior of Reactor Pressure Vessel RPV Steels in Light Water Reactor LWR Environments in Slow Strain Rate Tests SSRT written by S. Moriya and published by . This book was released on 1992 with total page 23 pages. Available in PDF, EPUB and Kindle. Book excerpt: Based upon a slip dissolution model, crack initiation and growth behavior of stress corrosion cracking of reactor pressure vessel steels, in pressurized high temperature water, was simulated by FEM (finite element method), with a combination of nodal force release techniques, simulating the crack growth by dissolution.
Download or read book Computer Modeling in Corrosion written by Raymond S. Munn and published by ASTM International. This book was released on 1992 with total page 296 pages. Available in PDF, EPUB and Kindle. Book excerpt:
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 Fracture Mechanics Applications written by Hardayal S. Mehta and published by . This book was released on 1994 with total page 180 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Environmentally Assisted Cracking in Light Water Reactors Semiannual Report July 1998 December 1998 written by and published by . This book was released on 1999 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 from July 1998 to December 1998. 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 of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. Fatigue tests have been conducted to determine the crack initiation and crack growth characteristics of austenitic SSs in LWR environments. Procedures are presented for incorporating the effects of reactor coolant environments on the fatigue life of pressure vessel and piping steels. Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to H".3 and 0.9 x 1021 n · cm−2 (E> 1 MeV) in helium at 289 C in the Halden reactor. The results have been used to determine the influence of alloying and impurity elements on the susceptibility of these steels to irradiation-assisted stress corrosion cracking. Fracture toughness J-R curve tests were also conducted on two heats of Type 304 SS that were irradiated to H".3 x 1021 n · cm−2 in the Halden reactor. Crack-growth-rate tests have been conducted on compact-tension specimens of Alloys 600 and 690 under constant load to evaluate the resistance of these alloys to stress corrosion cracking in LWR environments.
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 Environmentally Assisted Cracking in Light Water Reactors 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 by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2002. Topics that have been investigated include: (a) environmental effects on fatigue crack initiation in carbon and low-alloy steels and austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs in BWRs, (c) evaluation of causes and mechanisms of irradiation-assisted cracking of austenitic SS in PWRs, and (d) cracking in Ni-alloys and welds. A critical review of the ASME Code fatigue design margins and an assessment of the conservation in the current choice of design margins are presented. The existing fatigue {var_epsilon}-N data have been evaluated to define the effects of key material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs. Experimental data are presented on the effects of surface roughness on fatigue crack initiation in these materials in air and LWR environments. Crack growth tests were performed in BWR environments on SSs irradiated to 0.9 and 2.0 x 1021 n x cm−2. The crack growth rates (CGRs) of the irradiated steels are a factor of (almost equal to)5 higher than the disposition curve proposed in NUREG-0313 for thermally sensitized materials. The CGRs decreased by an order of magnitude in low-dissolved oxygen (DO) environments. Slow-strain-rate tensile (SSRT) tests were conducted in high-purity 289 C water on steels irradiated to (almost equal to)3 dpa. The bulk S content correlated well with the susceptibility to intergranular SCC in 289 C water. The IASCC susceptibility of SSs that contain>0.003 wt. % S increased drastically. bend tests in inert environments at 23 C were conducted on broken pieces of SSRT specimens and on unirradiated specimens of the same materials after hydrogen charging. The results of the tests and a review of other data in the literature indicate that IASCC in 289 C water is dominated by a crack-tip grain-boundary process that involves S. An initial IASCC model has been proposed. A crack growth test was completed on mill annealed Alloy 600 in high-purity water at 289 C and 320 C under various environmental and loading conditions. The results from this test are compared with data obtained earlier on several other heats of Alloy 600.
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 Irradiation assisted Stress Corrosion Cracking Behavior of Austenitic Stainless Steels Applicable to LWR Core Internals written by H. M. Chung and published by . This book was released on 2006 with total page 91 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 Radiation Embrittlement of Nuclear Reactor Pressure Vessel Steels written by Lendell E. Steele and published by ASTM International. This book was released on 1986 with total page 303 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 Slow Strain Rate Stress Corrosion Testing of RPV Steels in High Temperature Water written by J. Congleton and published by . This book was released on 1985 with total page 16 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book The Effect of Cold Rolling on the Susceptibility of Austenitic Stainless Steel to Stress Corrosion Cracking in Primary Circuit Pressurised Water Reactor Environment written by David Marc Wright and published by . This book was released on 2012 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The stress corrosion cracking (SCC) of components which are fabricated from austenitic stainless steel has been observed in the primary circuit of pressurised water reactors (PWR). In recent years it has become an increasing concern that cold work can induce susceptibility to SCC in these materials, even when exposed to good-quality flowing coolant. Laboratory studies which were launched in response to this observation have confirmed that SCC susceptibility is enhanced by cold work. The intention of this study is therefore to investigate the link between the effects of cold work on the material and the susceptibility to SCC. The investigation has been conducted on a grade 304 austenitic stainless steel. Characterisation of the microstructure and mechanical properties has been carried out in the annealed condition, and following cold rolling to a reduction in thickness of 20 %. The cold rolled material has then been subjected to SCC tests in simulated PWR primary circuit coolant. Two types of test were utilised: slow strain rate tests (SSRTs) were carried out in order to investigate the initiation of cracks from a smooth surface and constant load tests using pre-cracked specimens were used to investigate the crack propagation behaviour. In both types of test the SCC produced was predominantly intergranular. The SSRTs revealed that the most susceptible grain boundaries separated grains which had dissimilar deformation microstructures (one grain deformed heavily by planar bands, the other more homogenously). It was also observed that initiation could occur on a grain boundary which is adjacent to an annealing twin. In both microstructural configurations the susceptibility is likely to be due to the deformation incompatibility across the failed boundary, possible indicating that shear at the boundary is important for the initiation of cracking. The crack propagation behaviour of the rolled material was particularly anisotropic; regardless of the loading direction (specimens were manufactured to allow loading along the rolling, transverse and normal plate directions) cracking was observed to occur parallel to the rolling-transverse plane. The origin of this behaviour was explored in terms of preferential alignment of the deformation microstructure and the anisotropic mechanical properties of the rolled plate. Limited transgranular cracking was also observed, which occurred along oxidised deformation bands. The results overall indicate that heterogeneous deformation between different regions of the material, and preferential alignment of the deformation microstructure are important with respect to the SCC susceptibility of the rolled material.
Download or read book Crack Arrest Behavior of Reactor Pressure Vessel Steels at High Temperatures written by and published by . This book was released on 1988 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The Heavy-Section Steel Technology Program at the Oak Ridge National Laboratory under the sponsorship of the US Nuclear Regulatory Commission is conducting experimental and analytical studies to improve the understanding of conditions that govern the initiation, rapid propagation, arrest and ductile tearing of cracks in reactor pressure vessel (RPV) steels. In support of this objective, large-scale wide-plate experiments are performed to generate crack-arrest toughness data for RPV steels at temperatures approaching and above the onset of Charpy upper-shelf behavior. Analytical studies are addressing the role of dynamics and nonlinear rate-dependent (i.e., viscoplastic) effects in the interpretation of crack run-arrest events in these ductile materials. A summary of the wide-plate tests performed to date is presented, including details of test procedures, test data, and results of analyses performed to date. The importance of incorporating viscoplastic effects into dynamic analysis of crack run-arrest events in these strain-rate sensitive steels is examined through applications of selected proposed viscoplastic constitutive equations and fracture parameters to the interpretation of data from the wide-plate tests. The crack-arrest data are compared with those from small ASTM-type specimens and other large structural tests.
Download or read book Stress Corrosion Cracking of Candidate Structural Materials Under Simulated First wall written by and published by . This book was released on 1990 with total page 31 pages. Available in PDF, EPUB and Kindle. Book excerpt: Stress corrosion cracking (SCC) susceptibility of Types 316NG, 316, and 304 stainless steels (SS) was investigated in slow-strain-rate tests (SSRTs) in oxygenated water that simulates important parameters anticipated in first-wall/blanket systems. The water chemistry was based on a computer code which yielded the nominal concentrations of radiolytic species produced in an aqueous environment under ITER-type conditions. Actual SSRTs were performed in a less benign, more oxidizing reference environment at temperatures from 52 to 150°C. Predominantly ductile fracture was observed in Type 316NG and nonsensitized Types 316 SS and 304 SS SSRT specimens strained to failure in a reference ITER water chemistry. The failure behavior of Type 304 SS specimens heat-treated to yield sensitization values of 2, 3, or 20 Coulomb (C)/cm2 by the electrochemical potentiokinetic reactivation (EPR) technique, demonstrated that the degree of sensitization had a dramatic effect on intergranular stress corrosion cracking (IGSCC) susceptibility. Ranking for resistance to SCC in simulated ITER water by electron microscopy and SSRT parameters, i.e., failure time, ultimate strength, total elongation and stress ratio is 304 SS (EPR = 20
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.
