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 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 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 Scientific and Technical Aerospace Reports written by and published by . This book was released on 1994 with total page 312 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Structural Alloys for Nuclear Energy Applications written by Robert Odette and published by Newnes. This book was released on 2019-08-15 with total page 676 pages. Available in PDF, EPUB and Kindle. Book excerpt: High-performance alloys that can withstand operation in hazardous nuclear environments are critical to presentday in-service reactor support and maintenance and are foundational for reactor concepts of the future. With commercial nuclear energy vendors and operators facing the retirement of staff during the coming decades, much of the scholarly knowledge of nuclear materials pursuant to appropriate, impactful, and safe usage is at risk. Led by the multi-award winning editorial team of G. Robert Odette (UCSB) and Steven J. Zinkle (UTK/ORNL) and with contributions from leaders of each alloy discipline, Structural Alloys for Nuclear Energy Applications aids the next generation of researchers and industry staff developing and maintaining steels, nickel-base alloys, zirconium alloys, and other structural alloys in nuclear energy applications. This authoritative reference is a critical acquisition for institutions and individuals seeking state-of-the-art knowledge aided by the editors' unique personal insight from decades of frontline research, engineering and management. - Focuses on in-service irradiation, thermal, mechanical, and chemical performance capabilities. - Covers the use of steels and other structural alloys in current fission technology, leading edge Generation-IV fission reactors, and future fusion power reactors. - Provides a critical and comprehensive review of the state-of-the-art experimental knowledge base of reactor materials, for applications ranging from engineering safety and lifetime assessments to supporting the development of advanced computational models.

Download or read book Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors written by John H. Jackson and published by Springer. This book was released on 2018-12-20 with total page 2460 pages. Available in PDF, EPUB and Kindle. Book excerpt: This two-volume set represents a collection of papers presented at the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors. The purpose of this conference series is to foster an exchange of ideas about problems and their remedies in water-cooled nuclear power plants of today and the future. Contributions cover problems facing nickel-based alloys, stainless steels, pressure vessel and piping steels, zirconium alloys, and other alloys in water environments of relevance. Components covered include pressure boundary components, reactor vessels and internals, steam generators, fuel cladding, irradiated components, fuel storage containers, and balance of plant components and systems.

Download or read book Proceedings of the Eighth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors written by and published by . This book was released on 1997 with total page 496 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Inhibition of Stress Corrosion Cracking of Alloy X 750 by Prestrain written by and published by . This book was released on 1997 with total page 25 pages. Available in PDF, EPUB and Kindle. Book excerpt: Tests of precracked and as-notched compact tension specimens were conducted in 3600C hydrogenated water to determine the effect of prestrain on the stress corrosion cracking (SCC) resistance of Alloy X-750 in the HTH, AH and HOA heat treated conditions. Prestraining is defined as the intentional application of an initial load (or strain) that is higher than the final test load. Prestrain was varied from 10% to 40% (i.e., the initial to final load ratios ranged from 1.1 to 1.4). Other variables included notch root radius, stress level and irradiation. Specimens were bolt-loaded to maintain essentially constant displacement conditions during the course of the test. The frequent heat up and cooldown cycles that were necessary for periodic inspections provided an opportunity to evaluate the effect of test variables on rapid low temperature crack propagation to which this alloy is subject. For Condition HTH, application of 20% to 40% prestrain either eliminates or significantly retards SCC initiation in as-notched specimens and the onset of crack growth in precracked specimens. In addition, this procedure reduces the propensity for low temperature crack growth during cooldown. Similar results were observed for precracked HOA specimens. Application of 20% prestrain also retards SCC in as-notched and precracked AH specimens, but the effects are not as great as in Condition HTH. Prestraining at the 10% level was found to produce an inconsistent benefit. In-reactor SCC testing shows that prestrain greatly improves the in-flux and out-of-flux SCC resistance of Condition HTH material. No SCC was observed in precracked specimens prestrained 30%, whereas extensive cracking was observed in their nonprestrain counterparts.

Download or read book Energy Research Abstracts written by and published by . This book was released on 1994-08 with total page 442 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Understanding and Mitigating Ageing in Nuclear Power Plants written by Philip G Tipping and published by Elsevier. This book was released on 2010-10-26 with total page 953 pages. Available in PDF, EPUB and Kindle. Book excerpt: Plant life management (PLiM) is a methodology focussed on the safety-first management of nuclear power plants over their entire lifetime. It incorporates and builds upon the usual periodic safety reviews and licence renewals as part of an overall framework designed to assist plant operators and regulators in assessing the operating conditions of a nuclear power plant, and establishing the technical and economic requirements for safe, long-term operation. Understanding and mitigating ageing in nuclear power plants critically reviews the fundamental ageing-degradation mechanisms of materials used in nuclear power plant structures, systems and components (SSC), along with their relevant analysis and mitigation paths, as well as reactor-type specific PLiM practices. Obsolescence and other less obvious ageing-related aspects in nuclear power plant operation are also examined in depth. Part one introduces the reader to the role of nuclear power in the global energy mix, and the importance and relevance of plant life management for the safety regulation and economics of nuclear power plants. Key ageing degradation mechanisms and their effects in nuclear power plant systems, structures and components are reviewed in part two, along with routes taken to characterise and analyse the ageing of materials and to mitigate or eliminate ageing degradation effects. Part three reviews analysis, monitoring and modelling techniques applicable to the study of nuclear power plant materials, as well as the application of advanced systems, structures and components in nuclear power plants. Finally, Part IV reviews the particular ageing degradation issues, plant designs, and application of plant life management (PLiM) practices in a range of commercial nuclear reactor types. With its distinguished international team of contributors, Understanding and mitigating ageing in nuclear power plants is a standard reference for all nuclear plant designers, operators, and nuclear safety and materials professionals and researchers. Introduces the reader to the role of nuclear power in the global energy mix Reviews the fundamental ageing-degradation mechanisms of materials used in nuclear power plant structures, systems and components (SSC) Examines topics including elimination of ageing effects, plant design, and the application of plant life management (PLiM) practices in a range of commercial nuclear reactor types

Download or read book Government Reports Annual Index written by and published by . This book was released on 1994 with total page 1108 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book INIS Atomindex written by and published by . This book was released on 1995 with total page 550 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Selected Proceedings from the 231st ECS Meeting written by Alkire and published by The Electrochemical Society. This book was released on 2017-08-04 with total page 2056 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Stress Corrosion Cracking and Crack Tip Characterization of Alloy X 750 in Light Water Reactor Environments written by Jonathan Paul Gibbs and published by . This book was released on 2011 with total page 290 pages. Available in PDF, EPUB and Kindle. Book excerpt: Stress corrosion cracking (SCC) susceptibility of Inconel Alloy X-750 in the HTH condition has been evaluated in high purity water at 93 and 288°C under Boiling Water Reactor Normal Water Chemistry (NWC) and Hydrogen Water Chemistry (HWC) conditions. SCC crack growth rates of approximately 1.1 x 10-7 mm/s (K=28 MPa[square root symbol]m) under NWC conditions and 1.4xI 0-8 mm/s (K=28 MPa[square root symbol]m) under HWC in high purity water at 288°C were observed. The environmental conditions were changed from NWC to HWC during constant K loading, and the crack growth rate immediately slowed down by approximately one order of magnitude. The alloy was also tested in HWC at 93°C. No SCC crack growth was observed at K= 35 MPa[square root symbol]m for the length of time tested at 93°C. The fracture mode transitioned from predominantly transgranular cracking under fatigue conditions to a mixture of intergranular, pseudo-intergranular, and a small amount of transgranular fracture in constant stress intensity SCC. Pseudo-intergranular cracking is when a crack propagates directly adjacent to the grain boundary carbides and not actually on the grain boundary. The SCC crack tips were characterized with scanning electron microscopy (SEM) and 3D Atom Probe Tomography (APT). The SEM analysis was focused on the fractographic analysis and crack-propagation mode. The crack was observed to propagate adjacent to grain boundary carbides (pseudo-intergranular) and along a boundary with high coherency where no carbides were present (intergranular). The small and localized areas of transgranular cracking were occasionally seen between two regions of intergranular cracking. The APT reconstructions of the crack tips and crack wall identified several key features contributing to the SCC process: 1) Preferential oxygen transport occurs in either a finger-like or crystallographic morphology extending from the crack tip region. These regions are enriched in both oxygen and oxide with the oxide being a chromium-nickel spinel. 2) The matrix ahead of each finger-like "tunnel" is enriched in oxygen and predominantly chromium oxide. This indicates that oxygen is diffusing ahead of the crack tip into the bulk material. 3) The oxygen that penetrates directly into the base material from the crack walls in an ordered manner suggests that it is controlled by crystallographic features. 4) The main SCC crack tip is full of predominantly oxide phase and, to a lesser extent, metal atoms. The very crack tip forms a spinel of chromium and nickel oxides. Iron oxide begins to contribute to the oxide spinel approximately 25-30 nm from the actual tip. 5) The [gamma] precipitates that are directly adjacent to each crack tip and crack wall were deficient in aluminum content. The aluminum content in the bulk [gamma] was approximately 6.6 at% and the near-crack [gamma] aluminum content ranged from 2.5-3.5 at%. The range of affected [gamma] was approximately 100 nm wide.

Download or read book Effects of Neutron Irradiation on Deformation Behavior of Nickel base Fastener Alloys written by and published by . This book was released on 1999 with total page 10 pages. Available in PDF, EPUB and Kindle. Book excerpt: This paper presents the effects of neutron irradiation on the fracture behavior and deformation microstructure of high-strength nickel-base alloy fastener materials, Alloy X-750 and Alloy 625. Alloy X-750 in the HTH condition, and Alloy 625 in the direct aged condition were irradiated to a fluence of 2.4x102° n/cm2 at 264 C in the Advanced Test Reactor. Deformation structures at low strains were examined. It was previously shown that Alloy X-750 undergoes hardening, a significant degradation in ductility and an increase in intergranular fracture. In contrast, Alloy 625 had shown softening with a concomitant increase in ductility and transgranular failure after irradiation. The deformation microstructures of the two alloys were also different. Alloy X-750 deformed by a planar slip mechanism with fine microcracks forming at the intersections of slip bands with grain boundaries. Alloy 625 showed much more homogeneous deformation with fine, closely spaced slip bands and an absence of microcracks. The mechanism(s) of irradiation assisted stress corrosion cracking (IASCC) are discussed.

Download or read book Chemical Abstracts written by and published by . This book was released on 2002 with total page 2566 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Government Reports Announcements Index written by and published by . This book was released on 1995 with total page 782 pages. Available in PDF, EPUB and Kindle. Book excerpt: