Download or read book Current Understanding of the Irradiation Assisted Stress Corrosion Cracking Potential for PWRs written by H. Xu and published by . This book was released on 1998 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:

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 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 Understanding Susceptibility of In core Components to Irradiation assisted Stress Corrosion Cracking written by and published by . This book was released on 1991 with total page 9 pages. Available in PDF, EPUB and Kindle. Book excerpt: As nuclear plants age and accumulated fluences of core structural components increase, susceptibility of the components to irradiation-assisted stress corrosion cracking (IASCC) is also expected to increase. Irradiation-induced sensitization, commonly associated with an IASCC failure, was investigated in this study to provide a better understanding of long-term structural integrity of safety-significant in-core components. Irradiation-induced sensitization of high- and commercial-purity Type 304 stainless steels irradiated in BWRs was analyzed. 7 refs., 8 figs.

Download or read book Understanding Susceptibility of In core Components to Irradiation assisted Stress Corrosion Cracking written by and published by . This book was released on 1991 with total page 9 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 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 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 Fundamentals of Radiation Materials Science written by GARY S. WAS and published by Springer. This book was released on 2016-07-08 with total page 1014 pages. Available in PDF, EPUB and Kindle. Book excerpt: The revised second edition of this established text offers readers a significantly expanded introduction to the effects of radiation on metals and alloys. It describes the various processes that occur when energetic particles strike a solid, inducing changes to the physical and mechanical properties of the material. Specifically it covers particle interaction with the metals and alloys used in nuclear reactor cores and hence subject to intense radiation fields. It describes the basics of particle-atom interaction for a range of particle types, the amount and spatial extent of the resulting radiation damage, the physical effects of irradiation and the changes in mechanical behavior of irradiated metals and alloys. Updated throughout, some major enhancements for the new edition include improved treatment of low- and intermediate-energy elastic collisions and stopping power, expanded sections on molecular dynamics and kinetic Monte Carlo methodologies describing collision cascade evolution, new treatment of the multi-frequency model of diffusion, numerous examples of RIS in austenitic and ferritic-martensitic alloys, expanded treatment of in-cascade defect clustering, cluster evolution, and cluster mobility, new discussion of void behavior near grain boundaries, a new section on ion beam assisted deposition, and reorganization of hardening, creep and fracture of irradiated materials (Chaps 12-14) to provide a smoother and more integrated transition between the topics. The book also contains two new chapters. Chapter 15 focuses on the fundamentals of corrosion and stress corrosion cracking, covering forms of corrosion, corrosion thermodynamics, corrosion kinetics, polarization theory, passivity, crevice corrosion, and stress corrosion cracking. Chapter 16 extends this treatment and considers the effects of irradiation on corrosion and environmentally assisted corrosion, including the effects of irradiation on water chemistry and the mechanisms of irradiation-induced stress corrosion cracking. The book maintains the previous style, concepts are developed systematically and quantitatively, supported by worked examples, references for further reading and end-of-chapter problem sets. Aimed primarily at students of materials sciences and nuclear engineering, the book will also provide a valuable resource for academic and industrial research professionals. Reviews of the first edition: "...nomenclature, problems and separate bibliography at the end of each chapter allow to the reader to reach a straightforward understanding of the subject, part by part. ... this book is very pleasant to read, well documented and can be seen as a very good introduction to the effects of irradiation on matter, or as a good references compilation for experimented readers." - Pauly Nicolas, Physicalia Magazine, Vol. 30 (1), 2008 “The text provides enough fundamental material to explain the science and theory behind radiation effects in solids, but is also written at a high enough level to be useful for professional scientists. Its organization suits a graduate level materials or nuclear science course... the text was written by a noted expert and active researcher in the field of radiation effects in metals, the selection and organization of the material is excellent... may well become a necessary reference for graduate students and researchers in radiation materials science.” - L.M. Dougherty, 07/11/2008, JOM, the Member Journal of The Minerals, Metals and Materials Society.

Download or read book Irradiation assisted Stress Corrosion Cracking Considerations at Temperatures Below 288degreeC 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 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 Stress Corrosion Cracking in Light Water Reactors written by International Atomic Energy Agency and published by . This book was released on 2011 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: Provides general descriptions of degradation mechanisms of different types of stress corrosion cracking (SCC) which are concerned with systems, structures and components in PWRs and BWRs. This publication includes examples of good practices in preventing, mitigating and repairing SCC damage and summarizes research and development programmes.

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 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 Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems written by Steve Bruemmer and published by John Wiley & Sons. This book was released on 2013-10-18 with total page 1776 pages. Available in PDF, EPUB and Kindle. Book excerpt: This collection presents an exchange of ideas among scientists and engineers about the economic and safety concerns surrounding environmentally induced materials problems which lead to nuclear power plant outages. Scientists and engineers concerned with the environmental degradation processes (corrosion, mechanical, and radiation effects) present their latest results on such topics as life extension/relicensing and materials problems associated with spent fuel storage and radioactive waste disposal. This collection will be of interest to utility engineers, reactor vendor engineers, plant architect engineers, researchers concerned with materials degradation, and consultants involved in design, construction, and operation of water reactors.

Download or read book Stress Corrosion Cracking written by V S Raja and published by Elsevier. This book was released on 2011-09-22 with total page 817 pages. Available in PDF, EPUB and Kindle. Book excerpt: The problem of stress corrosion cracking (SCC), which causes sudden failure of metals and other materials subjected to stress in corrosive environment(s), has a significant impact on a number of sectors including the oil and gas industries and nuclear power production. Stress corrosion cracking reviews the fundamentals of the phenomenon as well as examining stress corrosion behaviour in specific materials and particular industries.The book is divided into four parts. Part one covers the mechanisms of SCC and hydrogen embrittlement, while the focus of part two is on methods of testing for SCC in metals. Chapters in part three each review the phenomenon with reference to a specific material, with a variety of metals, alloys and composites discussed, including steels, titanium alloys and polymer composites. In part four, the effect of SCC in various industries is examined, with chapters covering subjects such as aerospace engineering, nuclear reactors, utilities and pipelines.With its distinguished editors and international team of contributors, Stress corrosion cracking is an essential reference for engineers and designers working with metals, alloys and polymers, and will be an invaluable tool for any industries in which metallic components are exposed to tension, corrosive environments at ambient and high temperatures. - Examines the mechanisms of stress corrosion cracking (SCC) presenting recognising testing methods and materials resistant to SCC - Assesses the effect of SCC on particular metals featuring steel, stainless steel, nickel-based alloys, magnesium alloys, copper-based alloys and welds in steels - Reviews the monitoring and management of SCC and the affect of SCC in different industries such as petrochemical and aerospace

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.