In-Reactor Deformation of Zirconium Alloy Components PDF Download
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Author: R. A. Holt
Publisher:
ISBN:
Category : Anisotropy
Languages : en
Pages : 16
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Book Description
This paper briefly reviews work by the author identifying and describing in-reactor deformation mechanisms of materials and structures used in nuclear reactors, in particular, Zircaloy-2, Zircaloy-4, and Zr-2.5Nb, and the CANDU fuel channel (comprising Zr alloy pressure tubes, calandria tubes, and spacers). The discussion is set in the context of contemporary findings of other workers in the international community. The following themes are highlighted: The contributions of creep and growth to deformation; c-component dislocations and the fluence dependence of irradiation growth; anisotropy of irradiation growth; deformation equations and pressure tube-to-calandria tube contact in CANDU reactors; low temperature flux (damage rate) dependence of deformation rates. The first developments were reported in 1976 at the third conference in this series and there are ongoing developments in all areas. The linear low temperature flux dependence of creep and growth rates is yet to be satisfactorily explained.
Author: R. A. Holt
Publisher:
ISBN:
Category : Anisotropy
Languages : en
Pages : 16
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Book Description
This paper briefly reviews work by the author identifying and describing in-reactor deformation mechanisms of materials and structures used in nuclear reactors, in particular, Zircaloy-2, Zircaloy-4, and Zr-2.5Nb, and the CANDU fuel channel (comprising Zr alloy pressure tubes, calandria tubes, and spacers). The discussion is set in the context of contemporary findings of other workers in the international community. The following themes are highlighted: The contributions of creep and growth to deformation; c-component dislocations and the fluence dependence of irradiation growth; anisotropy of irradiation growth; deformation equations and pressure tube-to-calandria tube contact in CANDU reactors; low temperature flux (damage rate) dependence of deformation rates. The first developments were reported in 1976 at the third conference in this series and there are ongoing developments in all areas. The linear low temperature flux dependence of creep and growth rates is yet to be satisfactorily explained.
Author:
Publisher: ASTM International
ISBN:
Category :
Languages : en
Pages : 308
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Book Description
Author: E. F. Ibrahim
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
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Book Description
Author: D. G. Franklin
Publisher: ASTM International
ISBN: 9780803102590
Category : Science
Languages : en
Pages : 322
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Book Description
Author: Gerry D. Moan
Publisher: ASTM International
ISBN: 0803128959
Category : Nuclear fuel claddings
Languages : en
Pages : 891
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Book Description
Annotation The 41 papers of this proceedings volume were first presented at the 13th symposium on Zirconium in the Nuclear Industry held in Annecy, France in June of 2001. Many of the papers are devoted to material related issues, corrosion and hydriding behavior, in-reactor studies, and the behavior and properties of Zr alloys used in storing spent fuel. Some papers report on studies of second phase particles, irradiation creep and growth, and material performance during loss of coolant and reactivity initiated accidents. Annotation copyrighted by Book News, Inc., Portland, OR.
Author: George P. Sabol
Publisher: ASTM International
ISBN: 0803124066
Category : Nuclear fuel claddings
Languages : en
Pages : 907
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Book Description
Author: Craig M. Eucken
Publisher: ASTM International
ISBN: 080311463X
Category : Nuclear fuel claddings
Languages : en
Pages : 794
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Book Description
The proceedings of the Ninth International Symposium on [title], held in Kobe, Japan, November 1990, address current trends in the development, performance, and fabrication of zirconium alloys for nuclear power reactors. the bulk of the most recent work on zirconium alloy behavior has concerned corr
Author: Manfred P. Puls
Publisher: Springer Science & Business Media
ISBN: 1447141954
Category : Science
Languages : en
Pages : 475
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Book Description
By drawing together the current theoretical and experimental understanding of the phenomena of delayed hydride cracking (DHC) in zirconium alloys, The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components: Delayed Hydride Cracking provides a detailed explanation focusing on the properties of hydrogen and hydrides in these alloys. Whilst the emphasis lies on zirconium alloys, the combination of both the empirical and mechanistic approaches creates a solid understanding that can also be applied to other hydride forming metals. This up-to-date reference focuses on documented research surrounding DHC, including current methodologies for design and assessment of the results of periodic in-service inspections of pressure tubes in nuclear reactors. Emphasis is placed on showing how our understanding of DHC is supported by progress in general understanding of such broad fields as the study of hysteresis associated with first order phase transformations, phase relationships in coherent crystalline metallic solids, the physics of point and line defects, diffusion of substitutional and interstitial atoms in crystalline solids, and continuum fracture and solid mechanics. Furthermore, an account of current methodologies is given illustrating how such understanding of hydrogen, hydrides and DHC in zirconium alloys underpins these methodologies for assessments of real life cases in the Canadian nuclear industry. The all-encompassing approach makes The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Component: Delayed Hydride Cracking an ideal reference source for students, researchers and industry professionals alike.
Author: Erich Tenckhoff
Publisher: ASTM International
ISBN: 080310958X
Category : Anisotropy
Languages : en
Pages : 85
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Book Description
Author: CE. Ells
Publisher:
ISBN:
Category : Fracture (materials)
Languages : en
Pages : 17
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Book Description
This review summarizes the flow and fracture behavior of irradiated zirconium-niobium alloys. It is confined to the alloys currently used in power reactors, namely Zr-1.ONb, Zr-2.5Nb, and Zr-2.5Nb-0.5Cu, and of these, most of the literature deals with the Zr-2.5Nb alloy. Both before and after irradiation, the Zr-2.5Nb alloy has a good combination of strength and ductility; these properties have enabled reactor designers to replace Zircaloy with Zr-2.5Nb in applications where the superior in-reactor creep resistance of the latter alloy is important. The good ductility of the irradiated Zr-2.5Nb alloy, however, is dependent on the alloy remaining below the beta transus in the final stages of heat-treatment/fabrication before irradiation. Although there is much information on the deformation of the irradiated material relevant to reactor design, the basic understanding of the deformation is in an early stage of development.
