Phase Transformations in Neutron-irradiated Zircaloys PDF Download
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Languages : en
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Microstructural evolution in Zircaloy-2 and -4 spent-fuel cladding specimens after approx. 3 years of irradiation in commercial power reactors has been investigated by TEM and HVEM. Two kinds of precipitates induced by the fast-neutron irradiation in the reactors have been identified, i.e., Zr3O and cubic-ZrO2 particles approximately 2 to 10 nm in size. By means of a weak-beam dark-field ''2-1/2D-microscopy'' technique, the bulk nature of the precipitates and the surficial nature of artifact oxide and hydride phases could be discerned. The Zr(Fe/sub x/, Cr/sub 1-x/)2 and Zr2(Fe/sub x/, Ni/sub 1-x/) intermetallic precipitates normally present in the as-fabricated material virtually dissolved in the spent-fuel cladding specimens after a fast-neutron fluence of approx. 4 x 1021 ncm−2 in the power reactors. The observed radiation-induced phase transformations are compared with predictions based on the currently available understanding of the alloy characteristics. 29 refs.
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Languages : en
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Book Description
Microstructural evolution in Zircaloy-2 and -4 spent-fuel cladding specimens after approx. 3 years of irradiation in commercial power reactors has been investigated by TEM and HVEM. Two kinds of precipitates induced by the fast-neutron irradiation in the reactors have been identified, i.e., Zr3O and cubic-ZrO2 particles approximately 2 to 10 nm in size. By means of a weak-beam dark-field ''2-1/2D-microscopy'' technique, the bulk nature of the precipitates and the surficial nature of artifact oxide and hydride phases could be discerned. The Zr(Fe/sub x/, Cr/sub 1-x/)2 and Zr2(Fe/sub x/, Ni/sub 1-x/) intermetallic precipitates normally present in the as-fabricated material virtually dissolved in the spent-fuel cladding specimens after a fast-neutron fluence of approx. 4 x 1021 ncm−2 in the power reactors. The observed radiation-induced phase transformations are compared with predictions based on the currently available understanding of the alloy characteristics. 29 refs.
Author: HM. Chung
Publisher:
ISBN:
Category : Brittle failure
Languages : en
Pages : 24
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Book Description
Microstructural evolution in Zircaloy-2 and -4 spent-fuel cladding specimens after three years of irradiation in commercial power reactors has been investigated by transmission and high-voltage electron microscopies. Two kinds of precipitates induced by the fastneutron irradiation in the reactors have been identified, that is, Zr3O and cubic-ZrO2 particles approximately 2 to 10 nm in size. By means of specimen tilting and dark-field stereomicroscopy, the bulk nature of the cubic-ZrO2 particles has been shown. The Zr(Fex,Cr1-x)2 and Zr2(Fex,Ni1-x) intermetallic precipitates normally present in the as-fabricated material virtually dissolved in the spent-fuel cladding specimens after a fast-neutron fluence of ~4 x 1021 n/cm2 in the power reactors. The observed radiation-induced phase transformations are discussed in relation with the currently available understanding of the alloy characteristics.
Author: WJS Yang
Publisher:
ISBN:
Category : Amorphous transformation
Languages : en
Pages : 15
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Book Description
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.
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Category : Power resources
Languages : en
Pages : 782
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Author: George P. Sabol
Publisher: ASTM International
ISBN: 0803124066
Category : Nuclear fuel claddings
Languages : en
Pages : 907
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Author: R. W. Gilbert
Publisher:
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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: C. E. Coleman
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Category :
Languages : en
Pages : 0
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Category : Aeronautics
Languages : en
Pages : 1330
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Author: R. A. Holt
Publisher:
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Category :
Languages : en
Pages : 0
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