Download or read book Cracking Behavior and Microstructure of Austenitic Stainless Steels and Alloy 690 Irradiated in BOR 60 Reactor Phase I written by and published by . This book was released on 2010 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Cracking behavior of stainless steels specimens irradiated in the BOR-60 at about 320 C is studied. The primary objective of this research is to improve the mechanistic understanding of irradiation-assisted stress corrosion cracking (IASCC) of core internal components under conditions relevant to pressurized water reactors. The current report covers several baseline tests in air, a comparison study in high-dissolved-oxygen environment, and TEM characterization of irradiation defect structure. Slow strain rate tensile (SSRT) tests were conducted in air and in high-dissolved-oxygen (DO) water with selected 5- and 10-dpa specimens. The results in high-DO water were compared with those from earlier tests with identical materials irradiated in the Halden reactor to a similar dose. The SSRT tests produced similar results among different materials irradiated in the Halden and BOR-60 reactors. However, the post-irradiation strength for the BOR-60 specimens was consistently lower than that of the corresponding Halden specimens. The elongation of the BOR-60 specimens was also greater than that of their Halden specimens. Intergranular cracking in high-DO water was consistent for most of the tested materials in the Halden and BOR-60 irradiations. Nonetheless, the BOR-60 irradiation was somewhat less effective in stimulating IG fracture among the tested materials. Microstructural characterization was also carried out using transmission electron microscopy on selected BOR-60 specimens irradiated to ≈25 dpa. No voids were observed in irradiated austenitic stainless steels and cast stainless steels, while a few voids were found in base and grain-boundary-engineered Alloy 690. All the irradiated microstructures were dominated by a high density of Frank loops, which varied in mean size and density for different alloys.
Download or read book Irradiation Assisted Stress Corrosion Cracking of Austenitic Stainless Steels and Alloy 690 from Halden Phase II Irradiations written by Y. Chen and published by . This book was released on 2008 with total page 48 pages. Available in PDF, EPUB and Kindle. Book excerpt: This work is an ongoing effort at Argonne National Laboratory on the mechanistic study of irradiation-assisted stress corrosion cracking (IASCC) in the core internals of light water reactors.
Download or read book IRRADIATION ASSISTED STRESS CORROSION CRACKING OF MODEL AUSTENITIC STAINLESS STEELS IRRADIATED IN THE HALDEN REACTOR NUREG CR 5608 U S NUCLEAR REGULATORY Commission written by U.S. Nuclear Regulatory Commission and published by . This book was released on 1999* with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Irradiation Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments written by and published by . This book was released on 2010 with total page 115 pages. Available in PDF, EPUB and Kindle. Book excerpt: The internal components of light water reactors are exposed to high-energy neutron irradiation and high-temperature reactor coolant. The exposure to neutron irradiation increases the susceptibility of austenitic stainless steels (SSs) to stress corrosion cracking (SCC) because of the elevated corrosion potential of the reactor coolant and the introduction of new embrittlement mechanisms through radiation damage. Various nonsensitized SSs and nickel alloys have been found to be prone to intergranular cracking after extended neutron exposure. Such cracks have been seen in a number of internal components in boiling water reactors (BWRs). The elevated susceptibility to SCC in irradiated materials, commonly referred to as irradiation-assisted stress corrosion cracking (IASCC), is a complex phenomenon that involves simultaneous actions of irradiation, stress, and corrosion. In recent years, as nuclear power plants have aged and irradiation dose increased, IASCC has become an increasingly important issue. Post-irradiation crack growth rate and fracture toughness tests have been performed to provide data and technical support for the NRC to address various issues related to aging degradation of reactor-core internal structures and components. This report summarizes the results of the last group of tests on compact tension specimens from the Halden-II irradiation. The IASCC susceptibility of austenitic SSs and heat-affected-zone (HAZ) materials sectioned from submerged arc and shielded metal arc welds was evaluated by conducting crack growth rate and fracture toughness tests in a simulated BWR environment. The fracture and cracking behavior of HAZ materials, thermally sensitized SSs and grain-boundary engineered SSs was investigated at several doses (≤3 dpa). These latest results were combined with previous results from Halden-I and II irradiations to analyze the effects of neutron dose, water chemistry, alloy compositions, and welding and processing conditions on IASCC. The effect of neutron irradiation on the fracture toughness of austenitic SSs was also evaluated at dose levels relevant to BWR internals.
Download or read book Irradiation assisted Stress Corrosion Cracking of Model Austenitic Stainless Steels Irradiated in the Halden Reactor written by H. M. Chung and published by . This book was released on 1999 with total page 30 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Irradiation Behavior of Austenitic Stainless Steel Burnable Poisons written by R. G. Gray and published by . This book was released on 1962 with total page 32 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Stress corrosion Cracking written by Russell H. Jones and published by ASM International(OH). This book was released on 1992 with total page 466 pages. Available in PDF, EPUB and Kindle. Book excerpt: Details the many conditions under which stress-corrosion cracking (SCC) can occur, the parameters which control SCC, and the methodologies for mitigating and testing for SCC, plus information on mechanisms of SCC with experimental data on a variety of materials. Contains information about environmen
Download or read book Fracture Properties of Irradiated Alloys written by F. H. Huang and published by . This book was released on 1995 with total page 446 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Intergranular Stress Corrosion Cracking written by Hannu Hänninen and published by Nordic Council of Ministers. This book was released on 1989 with total page 92 pages. Available in PDF, EPUB and Kindle. Book excerpt: NKA
Download or read book IRRADIATION ASSISTED STRESS CORROSION CRACKING OF MODEL AUSTENITIC STAINLESS STEEL ALLOYS NUREG CR 6678 U S NUCLEAR REGULATORY COMMIS written by and published by . This book was released on 2001* with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Irradiation assisted Stress Corrosion Cracking of Model Austenitic Stainless Steel Alloys written by H. M. Chung and published by . This book was released on 2000 with total page 31 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Tensile Stress Corrosion Cracking of Type 304 Stainless Steel Irradiated to Very High Dose written by and published by . This book was released on 2001 with total page 14 pages. Available in PDF, EPUB and Kindle. Book excerpt: Certain safety-related core internal structural components of light water reactors, usually fabricated from Type 304 or 316 austenitic stainless steels (SSs), accumulate very high levels of irradiation damage (20--100 displacement per atom or dpa) by the end of life. The data bases and mechanistic understanding of, the degradation of such highly irradiated components, however, are not well established. A key question is the nature of irradiation-assisted intergranular cracking at very high dose, i.e., is it purely mechanical failure or is it stress-commotion cracking? In this work, hot-cell tests and microstructural characterization were performed on Type 304 SS from the hexagonal fuel can of the decommissioned EBR-11 reactor after irradiation to (approximately)50 dpa at (approximately)370 C. Slow-strain-rate tensile tests were conducted at 289 C in air and in water at several levels of electrochemical potential (ECP), and microstructural characteristics were analyzed by scanning and transmission electron microcopies. The material deformed significantly by twinning and exhibited surprisingly high ductility in air, but was susceptible to severe intergranular stress corrosion cracking (IGSCC) at high ECP. Low levels of dissolved O and ECP were effective in suppressing the susceptibility of the heavily irradiated material to IGSCC, indicating that the stress corrosion process associated with irradiation-induced grain-boundary Cr depletion, rather than purely mechanical separation of grain boundaries, plays the dominant role. However, although IGSCC was suppressed, the material was susceptible to dislocation channeling at low ECP, and this susceptibility led to poor work-hardening capability and low ductility.
Download or read book Effects of Water Chemistry on Intergranular Cracking of Irradiated Austenitic Stainless Steels written by and published by . This book was released on 1995 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Effect of Irradiation on the Stress Corrosion Cracking Behavior of Alloy X 750 and Alloy 625 written by and published by . This book was released on 1993 with total page 10 pages. Available in PDF, EPUB and Kindle. Book excerpt: In-reactor testing of bolt-loaded precracked and as-notched compact tension specimens was performed in 360°C water to determine effect of irradiation on SCC of Condition HTH and Condition BH Alloy X-750 and age-hardened Alloy 625. Variables were stress intensity factor (K{sub I}) level, fluence, grade of HTH material, prestraining and material chemistry. Effects of irradiation on high temperature SCC and the rapid cracking that occurs during cooldown below 150°C were characterized. Significant degradation in the in-reactor SCC resistance of HTH material was observed at initial K{sub I} levels above 30 MPa√m and fluences greater than 1019 n/cm2 (E > 1 MeV). A small degradation in SCC resistance of HTH material was observed at low fluences (1016 n/cm2). As-notched specimens displayed less degradation in SCC resistance than precracked specimens. Prestraining greatly improved in-flux and out-of-flux SCC resistance of HTH material, as little or no SCC was observed in precracked specimens prestrained 20 to 30%, whereas extensive cracking was observed in nonprestrained specimens. Condition HTH heats with low boron (10 ppM or less) had improved in-reactor SCC resistance compared to heats with high and intermediate boron (20 ppM). Age-hardened Alloy 625 exhibited superior in-reactor SCC behavior compared to HTH material as no crack extension occurred in any of the precracked Alloy 625 specimens tested at initial K{sub I} levels up to 80 MPa√m.
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 Radiation Hardening and Radiation induced Chromium Depletion Effects on Intergranular Stress Corrosion Cracking of Austenitic Stainless Steels written by and published by . This book was released on 1993 with total page 14 pages. Available in PDF, EPUB and Kindle. Book excerpt: Available data on neutron-irradiated materials have been analyzed and correlations developed between fluence, yield strength, grain boundary chromium concentration and cracking susceptibility in high-temperature water environments. Large heat-to-heat differences in critical fluence (0.2 to 2.5 n/cm[sup 2]) for IGSCC are documented. In many cases, this variability is consistent with yield strength differences among irradiated materials. IGSCC correlated better to yield strength than to fluence for most heats suggesting a possible role of the radiation-induced hardening (and microstructure) on cracking. However, isolatedheats reveal a wide range of yield strengths from 450 to 800 MPa necessary to promote IGSCC which cannot be understood by strength effects alone. Grain boundary Cr depletion explain differences in IGSCC susceptibility for irradiated stainless steels. Cr contents versus SCC shows that all materials showing IG cracking have some grain boundary depletion ([ge]2%). Grain boundary Cr concentrations for cracking (below [approximately]16 wt %) are in good agreement with similar SCC tests on unirradiated 304 SS with controlled depletion profiles. Heats that prompt variability in the yield strength correlation, are accounted for bydifferences in their interfacial Cr contents. Certain stainless steels are more resistant to cracking even though they have significant radiation-induced Cr depletion. It is proposed that Cr depletion is required for IASCC, but observed susceptibility is modified by other microchemical and microstructural components.
Download or read book Behaviour of Short Fatigue Cracks in Austenitic Stainless Steels written by H. Hurhmann and published by Bernan Assoc. This book was released on 1996 with total page 99 pages. Available in PDF, EPUB and Kindle. Book excerpt:
