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Author: R. A. Holt
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Languages : en
Pages : 0
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Author: R. A. Holt
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Category :
Languages : en
Pages : 0
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Author: Ram B. Roy
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Category :
Languages : en
Pages : 18
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Author: VN. Shishov
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Category : Composition
Languages : en
Pages : 20
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Studied were evolution of dislocation structure, phase, and element composition of binary alloys Zr-1Nb and Zr-2.5Nb and multicomponent alloys Zr-1Nb-1.2Sn-0.4Fe and Zr-1.2Sn-0.4Fe under neutron irradiation. The investigations were carried out using cladding and pressure tubes before and after irradiation to a fluence of ~1026 n/m2 (E >= 0.1 MeV) in experimental and commercial reactors at 300 to 350°C using TEM, EDX, and XRD. In most cases, irradiation-induced defects are in the form of dislocation loops with Burgers vector 1/3 ?1120?. The density of dislocations with a ?c? component is less than 2 x 1014 m-2. A higher fluence or the presence of strain results in the ordering of the dislocation structure of ?c? component and ?a?-type dislocation loops. Before irradiation, the multicomponent alloys contain fine precipitates of Zr-Nb-Fe composition, and the matrix is depleted in Fe. Under irradiation, recrystallization proceeds intensively (as distinct from Zr-Nb alloys), changes take place in size, distribution, and composition of precipitates (with a relative decrease of Fe content compared to Nb), and the Fecontent of ?-Zr matrix is increased. None of the materials studied showed any significant evidence of secondary phase particle amorphization. The density of dislocations with ?a? and ?c? components and irradiation-induced defects, their mean size, the extent of ordering, and the planes of their occurrence were determined. A comparison was made between irradiation-induced evolutions of microstructures of the different alloys.
Author: R. W. Gilbert
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Category :
Languages : en
Pages : 0
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Author: R. W. Gilbert
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Category :
Languages : en
Pages : 0
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Author: RB. Adamson
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Category : Crack initiation
Languages : en
Pages : 17
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The onset of plastic instability in irradiated materials is known to be governed by deformation processes that occur on a microscopic scale. In order to identify the nature of the controlling mechanisms giving rise to plastic instability, details of the deformation processes were examined for Zircaloy-2 irradiated to fluences of 1.0 to 3.0 x 1021 neutrons (n)/cm2 (E >1 MeV) and tested in inert and 0.3-torr iodine atmospheres at 322 ~ 335°C. For irradiated materials tested in air, it was shown that plastic strain at the maximum load decreased by more than an order of magnitude relative to unirradiated material, especially under plane-stress conditions. This was due largely to concentrated shear deformation resulting from dislocation channel development. When the similar materials were tested in iodine, the load-carrying capacity was reduced further. Based on the evidence obtained by metallographic examination, a model has been proposed describing deformation behavior of irradiated plane-stress specimens by assuming consecutive formation and strain hardening of new deformation zones containing channels. Crack nucleation under an iodine atmosphere is believed to occur when the local strain at the channel zone becomes sufficiently large to break oxide films and cause intergranular cracking.
![Damage Stucture in Zirconium Alloys Irradiated at 573 to 923°K. [Neutron Fluence 1 X 1025 N. M−2]. PDF](https://books.google.com/books/content/images/frontcover/oEZsAQAACAAJ?fife=w80-h100)
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Languages : en
Pages :
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The microstructures of annealed zirconium, Zircaloy-2 and Zr-2.5 wt % Nb alloy and of Zr-2.5 Nb containing .cap alpha.' were studied after neutron irradiation to fluences approximately equal to 1 x 1025 n x m−2 in the temperature range 573 to 923°K. The principal form of damage was dislocation loops which increased in size and decreased in density with increasing temperature and which did not exist above 773°K. The Burgers vector of the loops was consistent with a/3 1120. Half or more of the loops were of vacancy type. No dislocation networks or voids were seen. It is argued that the bias of loops for self-interstitial atoms in .cap alpha.-zirconium is very weak, permitting competitive vacancy and interstitial loops, preventing growth of loops into gross dislocation structure, and depressing the vacancy super-saturation so that voids cannot arise.
Author: WJS Yang
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Category : Amorphous transformation
Languages : en
Pages : 15
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Zircaloy-4, a zirconium base alloy used extensively as cladding and core structural material in water cooled nuclear reactors, was examined by transmission electron microscopy after neutron irradiation and postirradiation annealing. Phase instabilities found during irradiation include the amorphous transformation and the dissolution of intermetallic precipitate Zr(Fe,Cr)2 in the ?-recrystallized matrix and the dissolution of the metastable precipitate Zr4(Fe,Cr) in the ?-quenched matrix. The alloy is driven toward a single phase solid solution during the irradiation. The presence of fast diffusion iron species in the matrix due to the precipitate dissolution may have caused the irradiation growth breakaway phenomenon. The microstructural evolution during irradiation consists of ̄c dislocation development and grain boundary migration. The presence of ̄c dislocations indicates permanent strain in the matrix. The postirradiation annealing at 833 K does not anneal out the ̄c dislocations. The ̄c dislocation is postulated to have developed due to the intergranular constraints under the continuous growth in the breakaway region.
Author: W. B. Rotsey
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Category : Cold working
Languages : en
Pages : 16
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The strain-aging behavior of annealed Zircaloy-2 in vacuum was studied as a function of deformation temperature, integrated fast-neutron dose, and degree of cold work. Strain aging occurred in the temperature range 200 to 500 C. Transmission electron microscopy showed that in the temperature range 300 to 450 C, dislocations were arranged in a cell structure, and that at 500 C two-dimensional hexagonal dislocation nets were present. Between 200 and 400 C, the strain aging was found to be due to the presence of oxygen; however, the origin of the strain aging between 400 and 500 C was not identified.
Author: R. W. Gilbert
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Category :
Languages : en
Pages : 0
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