Download or read book Temperature Shift of Void Swelling Observed in Annealed Fe 18Cr 10Ni Ti Stainless Steel Irradiated in the Reflector Region of BOR 60 written by VK. Shamardin and published by . This book was released on 2000 with total page 9 pages. Available in PDF, EPUB and Kindle. Book excerpt: Evidence has been provided to support the "temperature-shift" concept of swelling. The swelling of annealed X18H10T stainless steel in the BOR-60 fast reactor at PWR-relevant dpa rates was found to be higher than predicted by a swelling equation developed from data obtained at dpa rates almost an order of magnitude higher. These data were derived from identical hexagonal wrappers irradiated in both in-core and reflector positions. Based on these results, irradiation in PWRs would yield higher swelling to lower temperatures than would occur in core regions of fast reactors.
Download or read book Effects of Radiation on Materials written by Margaret L. Hamilton and published by ASTM International. This book was released on 2000 with total page 1266 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Effects of Radiation on Materials written by Martin L. Grossbeck and published by ASTM International. This book was released on 2004 with total page 767 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 Void Swelling at Low Displacement Rates in Annealed X18H10T Stainless Steel at 4 to 56 DPA and 280 332 Degrees Centigrade written by and published by . This book was released on 2001 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Various components of pressurized water power reactors (PWRs) and some proposed fusion devices such as ITER will operate at lower temperatures and displacement rates than are encountered in many test reactors such as EBR-II, FFTF and HFIR. The question arises if the presence and magnitude of void swelling can be predicted for such irradiation environments. Data on Russian steel can be used to address part of this question. In reactor applications where Western countries typically use annealed AISI 304 stainless steel, it is the Russian practice to use annealed X18H10T, a titanium-stabilized 18Cr-10Ni stainless steel analogous to AISI 321. Using a flow restrictor component from the low-flux breeder zone of the BN-350 reactor in Kazakhstan, it was possible to examine the behavior of void swelling at relatively low temperatures and low displacement rates after 12 years of irradiation. The temperature of this component ranged from 270-340 degrees centigrade with a peak dose rate of 1.6 x 10−7 power dpa/sec and a peak dose of 56 dpa. Careful sectioning of the component has yielded a large number of microscopy specimens over a ITER-relevant range of temperatures and displacement rates. Microstructural data are presented and show that void swelling at 10 to 50 dpa persists down to (almost equal to)306 degrees centigrade for dose rates on the order of 1 x 10−7 power dpa/sec.
Download or read book Modeling of Void Swelling in Irradiated Steels written by BB. Glasgow and published by . This book was released on 1985 with total page 15 pages. Available in PDF, EPUB and Kindle. Book excerpt: The theory of void swelling in irradiated steels has been critically re-examined and a model has been developed from first principles for such fundamental parameters as the bias factors for dislocations and voids, relaxation volumes for interstitials and vacancies, and migration and formation energies for interstitials and vacancies. Four types of growth mechanisms are modeled: self-interstitial emission, loop punching, thermal vacancy exchange, and bias driven growth. The net bias of the system is allowed to change continuously with the evolving microstructure. Combining these growth mechanisms has resulted in a swelling model capable of reproducing some important features of the observed swelling behavior in austenitic stainless steels. For austenitic stainless steels, the model predicts bias driven growth for temperatures between 300°C and 600°C. The long-term swelling rate in this temperature range is roughly independent of temperature and helium concentration and is between 0.7%/dpa (displacements per atom) and 1.3%/dpa. For ferrite phase, the model predicts bias driven growth for temperatures between 300°C and 500°C. However, for this class of steel, the swelling rate is at or below 0.3%/dpa. For bias driven growth, it has been shown that the interstitial and vacancy relaxation volumes ultimately determine the maximum possible swelling rate. It is for this reason that the model predicts a much lower swelling rate for ferrite phase than for austenitic stainless steels.
Download or read book Void Swelling and Microstructural Change in Neutron Irradiated Type 316 Stainless Steel written by M. Itoh and published by . This book was released on 1987 with total page 13 pages. Available in PDF, EPUB and Kindle. Book excerpt: Microstructural observations were made on cold-worked Type 316 stainless steel following fast neutron irradiation to a peak neutron fluence of 1.3 x 1027 n/m2 (E > 0.1 MeV) at temperatures from 400 to 620°C. The results show that nickel depletion in the matrix after neutron irradiation is related to the formation of phosphides and void swelling. Phosphorus in solution retards not only the recovery of dislocations but also the development of nickel- and silicon-rich phases, and this results in suppressing nickel depletion in the matrix and an extension of the transient regime in void swelling. Phosphides are formed during the transient regime at temperatures from 490 to 570°C. After phosphides are formed, retardation effects on both dislocation recovery and the formation of nickel- and silicon-rich phases are lost, and this leads to the onset of void swelling.
Download or read book Irradiation Effects on the Microstructure and Properties of Metals written by and published by ASTM International. This book was released on 1976 with total page 491 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book The Growth and Coalescence of Voids in Irradiated Metals written by Louis Kenneth Mansur and published by . This book was released on 1974 with total page 270 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book OBSERVATIONS OF VOID SWELLING IN 12R72HV AUSTENITIC STAINLESS STEEL AFTER46 MEV NI SUP 6 IRRADIATION TO 120 DPA IN THE HARWELL VEC written by R. M. Deakin and published by . This book was released on 1987 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Determination of the Lower Temperature Limit of Void Swelling of Stainless Steels at Relatively Low Displacement Rates written by and published by . This book was released on 2002 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: An issue of current interest to PWRs is the possibility that void swelling of austenitic near-core internal components may exert some deleterious effect on component functionality, particularly during extended operation to 60 years. A similar concern has also been raised for water-cooled fusion devices. One question of particular interest is the range of temperature over which void swelling can occur, since the internal components experience temperatures from ~290 to perhaps as high as 390 degrees C in some limited locations. This question was addressed using a flow restrictor component from the low-flux breeder zone of the BN-350 fast reactor in Kazakhstan. This component was constructed of annealed 12X18H10T, an alloy similar to AISI 321 which is used in Russian reactors for applications where AISI 304L would be used in comparable Western and Japanese reactors. Extensive sectioning to produce 114 separate specimens, followed by examination of the radiation-induced microstructure showed that void swelling in the range of temperatures and dpa rates of PWR interest occurs down to ~305 degrees C. At 330 degrees C the swelling reached ~1% at 20 dpa. Comparison of these data with other published results from Russian LWR reactors at
Download or read book IRRADIATION CREEP AND SWELLING OF ANNEALED TYPE 304L STAINLESS STEEL AT APPROXIMATELY 390 DEGREE C AND HIGH NEUTRON FLUENCE written by D.L. PORTER. and published by . This book was released on 1989 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Comparison of Irradiation Creep and Swelling of an Austenitic Alloy Irradiated in FFTF and PFR written by FA. Garner and published by . This book was released on 1999 with total page 12 pages. Available in PDF, EPUB and Kindle. Book excerpt: Comparative irradiation of identically constructed creep tubes in the Fast Flux Test Facility (FFTF) and the Prototypic Fast Reactor (PFR) shows that differences in irradiation conditions arising from both reactor operation and the design of the irradiation vehicle can have a significant impact on the void swelling and irradiation creep of austenitic stainless steels. In spite of these differences, the derived creep coefficients fall within the range of previously observed values for 316 SS.
Download or read book Effects of Second Phase Particles on Irradiation Swelling of Austenitic Alloys written by WK. Appleby and published by . This book was released on 1973 with total page 15 pages. Available in PDF, EPUB and Kindle. Book excerpt: Results are described from the investigation of swelling caused by fast-neutron-induced void formation in stainless steel fuel cladding and high-nickel alloys irradiated in fast flux (EBR-II) to fluences up to 7.3 x 1022 n/cm2, E > 0.1 MeV. Immersion density measurements of swelling gave results ranging from a volume increase of 10 percent in solution-annealed Type 347 stainless steel to a small densification in Inconel-625. Transmission electron microscopy investigations revealed that the extent of void formation in Types 316 and 347 stainless steels, Incoloy-800, and Hastelloy-X was related to the degree of intragranular precipitation (principally carbides), whereas the Ni3 Cb precipitation in Inconel-625 had apparently inhibited void formation altogether. It is concluded that precipitate particles can play a major role in determining the extent of void formation in austenitic alloys during neutron irradiation, and in the optimum morphology (a high concentration of closely spaced, small and possibly coherent particles) they offer promise of largely inhibiting void formation and swelling.
Download or read book Surface Reaction Limited Void Growth written by C-Y Li and published by . This book was released on 1975 with total page 18 pages. Available in PDF, EPUB and Kindle. Book excerpt: The surface reaction rate controlled void growth model is extended to include the effects of point defect recombination and thermal emission of vacancies from the void surface. Experimental data on solution annealed Type 316 stainless steel from both neutron irradiation and ion bombardment are examined using this model. It is concluded that void growth in this material is controlled by reaction rate at the void surface; also, the temperature dependence of void growth rate mainly reflects the temperature dependence of dislocation density. Approaches for swelling data extrapolation and translation are discussed.
Download or read book Liquid Metal Cooled Reactors written by International Atomic Energy Agency and published by . This book was released on 2007 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: Presents a survey of worldwide experience gained with fast breeder reactor design, development and operation. Coverage includes state of the art of liquid metal fast reactor development; lead-bismuth cooled (LBC) ship reactor operation experience and LBC fast power reactor development; and treatment and disposal of spent sodium.
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
