Download or read book Studies on the Mechanism for Irradiation Assisted Stress Corrosion Cracking written by Kjell Pettersson and published by . This book was released on 1994 with total page 15 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Irradiation assisted Stress Corrosion Cracking of Fusion Reactor Material written by and published by . This book was released on 1990 with total page 22 pages. Available in PDF, EPUB and Kindle. Book excerpt: Irradiation-assisted stress-corrosion cracking (IASCC) is a phenomenon produced by radiation-induced alterations in the material and environment. These alternations include radiation-induced segregation and depletion of specific elements at grain boundaries, radiation creep and hardening and radiolytic effects induced in the aqueous environment. This phenomenon has been clearly identified as an active crack growth mechanism for in-core components in fission reactor must be considered as a potential crack growth mechanism for water-cooled fusion reactors such as ITER or power reactors. The potential for IASCC phenomenon occurring in ITER structural materials is being evaluated by modeling and experiment. Results from modeling calculations for impurity segregation at ITER-relevant temperatures have been completed and suggest that this phenomenon is not likely to induce IASCC during the ITER design life. If a fusion power reactor is water cooled, IASCC is a definite concern for austenitic stainless steels. It has been clearly demonstrated with modeling and experimental measurements that Cr depletion occurs within about 1 dpa. Phosphorus and Si grain boundary segregation can also occur at this same dose and temperature but their effect on IASCC appears to be secondary to Cr depletion. Also, irradiation creep-induced crack tip strain appears to be a secondary effect. However, there are a number of unexplained observations in the literature on IASCC which may be caused by radiation damage effects other than Cr depletion or impurity segregation.
Download or read book Localized Deformation as a Primary Cause of Irradiation Assisted Stress Corrosion Cracking written by and published by . This book was released on 2009 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The objective of this project is to determine whether deformation mode is a primary factor in the mechanism of irradiation assisted intergranular stress corrosion cracking of austenitic alloys in light watert reactor core components. Deformation mode will be controlled by both the stacking fault energy of the alloy and the degree of irradiation. In order to establish that localized deformation is a major factor in IASCC, the stacking fault energies of the alloys selected for study must be measured. Second, it is completely unknown how dose and SFE trade-off in terms of promoting localized deformation. Finally, it must be established that it is the localized deformation, and not some other factor that drives IASCC.
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 0 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 Y. Chen and published by . This book was released on 2010 with total page 89 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Design of an Experimental Facility for Irradiation Assisted Stress Corrosion Cracking Studies written by Julio Andres Vergara and published by . This book was released on 1992 with total page 316 pages. Available in PDF, EPUB and Kindle. Book excerpt:
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 Assessment of Initial Test Conditions for Experiments to Assess Irradiation Assisted Stress Corrosion Cracking Mechanisms written by and published by . This book was released on 2011 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Irradiation-assisted stress corrosion cracking is a key materials degradation issue in today s nuclear power reactor fleet and affects critical structural components within the reactor core. The effects of increased exposure to irradiation, stress, and/or coolant can substantially increase susceptibility to stress-corrosion cracking of austenitic steels in high-temperature water environments. . Despite 30 years of experience, the underlying mechanisms of IASCC are unknown. Extended service conditions will increase the exposure to irradiation, stress, and corrosive environment for all core internal components. The objective of this effort within the Light Water Reactor Sustainability program is to evaluate the response and mechanisms of IASCC in austenitic stainless steels with single variable experiments. A series of high-value irradiated specimens has been acquired from the past international research programs, providing a valuable opportunity to examine the mechanisms of IASCC. This batch of irradiated specimens has been received and inventoried. In addition, visual examination and sample cleaning has been completed. Microhardness testing has been performed on these specimens. All samples show evidence of hardening, as expected, although the degree of hardening has saturated and no trend with dose is observed. Further, the change in hardening can be converted to changes in mechanical properties. The calculated yield stress is consistent with previous data from light water reactor conditions. In addition, some evidence of changes in deformation mode was identified via examination of the microhardness indents. This analysis may provide further insights into the deformation mode under larger scale tests. Finally, swelling analysis was performed using immersion density methods. Most alloys showed some evidence of swelling, consistent with the expected trends for this class of alloy. The Hf-doped alloy showed densification rather than swelling. This observation may be related to the formation of second-phases under irradiation, although further examination is required.
Download or read book Irradiation Programs and Test Plans to Assess High Fluence Irradiation Assisted Stress Corrosion Cracking Susceptibility written by and published by . This book was released on 2015 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: . Irradiation assisted stress corrosion cracking (IASCC) is a known issue in current reactors. In a 60 year lifetime, reactor core internals may experience fluence levels up to 15 dpa for boiling water reactors (BWR) and 100+ dpa for pressurized water reactors (PWR). To support a safe operation of our fleet of reactors and maintain their economic viability it is important to be able to predict any evolution of material behaviors as reactors age and therefore fluence accumulated by reactor core component increases. For PWR reactors, the difficulty to predict high fluence behavior comes from the fact that there is not a consensus of the mechanism of IASCC and that little data is available. It is however possible to use the current state of knowledge on the evolution of irradiated microstructure and on the processes that influences IASCC to emit hypotheses. This report identifies several potential changes in microstructure and proposes to identify their potential impact of IASCC. The susceptibility of a component to high fluence IASCC is considered to not only depends on the intrinsic IASCC susceptibility of the component due to radiation effects on the material but to also be related to the evolution of the loading history of the material and interaction with the environment as total fluence increases. Single variation type experiments are proposed to be performed with materials that are representative of PWR condition and with materials irradiated in other conditions. To address the lack of IASCC propagation and initiation data generated with material irradiated in PWR condition, it is proposed to investigate the effect of spectrum and flux rate on the evolution of microstructure. A long term irradiation, aimed to generate a well-controlled irradiation history on a set on selected materials is also proposed for consideration. For BWR, the study of available data permitted to identify an area of concern for long term performance of component. The efficiency of hydrogen water chemistry mitigation technology may decrease as fluence increases for high-stress intensity factors. This report describes a program plan to determine the efficiency of hydrogen water chemistry as a function of the stress intensity factor applied and fluence. The use of existing, available, materials and the generation of additional materials via irradiation in a research reactor are considered.
Download or read book Irradiation Assisted Stress Corrosion Cracking of HTH Alloy X 750 and Alloy 625 written by and published by . This book was released on 1994 with total page 32 pages. Available in PDF, EPUB and Kindle. Book excerpt: In-reactor testing of bolt-loaded precracked compact tension specimens was performed in 360°C water to determine effect of irradiation on the SCC behavior of HTH Alloy X-750 and direct aged Alloy 625. Out-of-flux and autoclave control specimens provided baseline data. Primary test variables were stress intensity factor, fluence, chemistry, processing history, prestrain. Results for the first series of experiments were presented at a previous conference. Data from two more recent experiments are compared with previous results; they confirm that high irradiation levels significantly reduce SCC resistance in HTH Alloy X-750. Heat-to-heat differences in IASCC were related to differences in boron content, with low boron heats showing improved SCC resistance. The in-reactor SCC performance of Alloy 625 was superior to that for Alloy X-750, as no cracking was observed in any Alloy 625 specimens even though they were tested at very high K1 and fluence levels. A preliminary SCC usage model developed for Alloy X-750 indicates that in-reactor creep processes, which relax stresses but also increase crack tip strain rates, and radiolysis effects accelerate SCC. Hence, in-reactor SCC damage under high flux conditions may be more severe than that associated with postirradiation tests. In addition, preliminary mechanism studies were performed to determine the cause of IASCC In Alloy X-750.
Download or read book Irradiation assisted Stress Corrosion Cracking Considerations at Temperatures Below 288 C written by and published by . This book was released on 1995 with total page 15 pages. Available in PDF, EPUB and Kindle. Book excerpt: Irradiation-assisted stress corrosion cracking (IASCC) occurs above a critical neutron fluence in light-water reactor (LWR) water environments at 288 C, but very little information exists to indicate susceptibility as temperatures are reduced. Potential low-temperature behavior is assessed based on the temperature dependencies of intergranular (IG) SCC in the absence of irradiation, radiation-induced segregation (RIS) at grain boundaries and micromechanical deformation mechanisms. IGSCC of sensitized SS in the absence of irradiation exhibits high growth rates at temperatures down to 200 C under conditions of anodic dissolution control, while analysis of hydrogen-induced cracking suggests a peak crack growth rate near 100 C. Hence from environmental considerations, IASCC susceptibility appears to remain likely as water temperatures are decreased. Irradiation experiments and model predictions indicate that RIS also persists to low temperatures. Chromium depletion may be significant at temperatures below 100C for irradiation doses greater than 10 displacements per atom (dpa). Macromechanical effects of irradiation on strength and ductility are not strongly dependent on temperature below 288 C. However, temperature does significantly affect radiation effects on SS microstructure and micromechanical deformation mechanisms. The critical conditions for material susceptibility to IASCC at low temperatures may be controlled by radiation-induced grain boundary microchemistry, strain localization due to irradiation microstructure and irradiation creep processes. 39 refs.
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 assisted Stress Corrosion Cracking of Materials from Commercial BWRs written by and published by . This book was released on 1993 with total page 18 pages. Available in PDF, EPUB and Kindle. Book excerpt: Constant-extension-rate tensile tests and grain-boundary analysis by Auger electron spectroscopy which were conducted on high- and commercial-purity (HP and CP) Type 304 stainless steel (SS) specimens from irradiated boiling-water reactor (BWR) components to determine susceptibility to irradiation-assisted stress corrosion cracking (IASCC) and to identify the mechanisms of intergranular failure. The susceptibility of HP neutron absorber tubes to intergranular stress corrosion cracking (IGSCC) was higher than that of CP absorber tubes or CP control blade sheath. Contrary to previous beliefs, susceptibility to intergranular fracture could not be correlated with radiation-induced segregation of impurities such as Si, P, C, N, or S, but a correlation was obtained with grain-boundary Cr concentration, indicating a role for Cr depletion that promotes IASCC. Detailed analysis of grain-boundary chemistry was conducted on neutron absorber tubes that were fabricated from two similar heats of HP Type 304 SS of virtually identical bulk chemical composition but exhibiting a significant difference in susceptibility to IGSCC for similar fluence. Grain-boundary concentrations of Cr, Ni, Si, P, S, and C in the crack-resistant and susceptible HP heats were virtually identical. However, grain boundaries of the cracking-resistant material contained less N and more B and Li (transmutation product from B) than those of the crack-susceptible material, indicating beneficial effects of low N and high B contents.
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 Nuclear Corrosion written by Stefan Ritter and published by Woodhead Publishing. This book was released on 2020-08-29 with total page 496 pages. Available in PDF, EPUB and Kindle. Book excerpt: Nuclear Corrosion: Research, Progress and Challenges, part of the "Green Book” series of the EFC, builds upon the foundations of the very first book published in this series in 1989 ("Number 1 - Corrosion in the Nuclear Industry”). This newest volume provides an overview on state-of-the-art research in some of the most important areas of nuclear corrosion. Chapters covered include aging phenomena in light water reactors, reprocessing plants, nuclear waste disposal, and supercritical water and liquid metal systems. This book will be a vital resource for both researchers and engineers working within the nuclear field in both academic and industrial environments. Discusses industry related aspects of materials in nuclear power generation and how these materials react with the environment Provides comprehensive coverage of the topic as written by noted experts in the field Includes coverage of nuclear waste corrosion
Download or read book Irradiation Induced Microchemical Effects and IASCC Initiation written by EP. Simonen and published by . This book was released on 1992 with total page 12 pages. Available in PDF, EPUB and Kindle. Book excerpt: Predicted irradiation-induced enrichment of silicon and depletion of chromium at grain boundaries were consistent with measurements of ion-irradiated stainless steel grain boundary microchemistries. The calibrated solute drag mechanism for silicon enrichment and the inverse-Kirkendall mechanism for chromium depletion were then used to predict behaviors for neutron irradiation conditions relevant to irradiation-assisted stress corrosion cracking. A simple crack chemistry model illustrated how irradiation-induced chromium depletion might affect intergranular corrosion as a function of dose.
Download or read book Effects of Irradiation on Intergranular Stress Corrosion Cracking of Type 304 Stainless Steel written by and published by . This book was released on 1989 with total page 11 pages. Available in PDF, EPUB and Kindle. Book excerpt:
