An Experimental Facility for Oxidation and Biaxial Deformation Studies of Zircaloy-4 Fuel Cladding Samples PDF Download
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Category :
Languages : en
Pages : 14
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Book Description
Description of the major components and performance characteristics of theFuel Cladding Test Facility developed by Ontario Hydro Research Division.This facility willgenerate materials data on the oxidation and deformation behaviour of CANDUreactor Zircaloy-4 fuel cladding for temperatures in the range 673-1173K.Such data are needed in the analysis of some post-shutdown accident scenariosand fuel handling incidents.The test facility consists of: tube furnaces forspecimen heating under closed-loop control; apressurizing circuit for deformation under biaxial conditions; amicroprocessor-based data logger for data acquisition; and an elapsed timeindicator for time to failure monitoring.
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 14
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Book Description
Description of the major components and performance characteristics of theFuel Cladding Test Facility developed by Ontario Hydro Research Division.This facility willgenerate materials data on the oxidation and deformation behaviour of CANDUreactor Zircaloy-4 fuel cladding for temperatures in the range 673-1173K.Such data are needed in the analysis of some post-shutdown accident scenariosand fuel handling incidents.The test facility consists of: tube furnaces forspecimen heating under closed-loop control; apressurizing circuit for deformation under biaxial conditions; amicroprocessor-based data logger for data acquisition; and an elapsed timeindicator for time to failure monitoring.
Author: E. T. C. Ho
Publisher:
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Category :
Languages : en
Pages : 0
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Author: E. T. C. Ho
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Category :
Languages : en
Pages : 0
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Category : Canada
Languages : en
Pages : 846
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Book Description
"An index and document delivery service for Canadian report literature".
Author: Robert S. Daum
Publisher:
ISBN:
Category : Cladding
Languages : en
Pages : 22
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Book Description
Sufficient mechanical ductility of high-burnup Zircaloy-4 fuel cladding is important to prevent large-opening ruptures and significant fuel dispersal during postulated in-reactor and spent-fuel processing accidents. The effect of irradiation, oxidation, and hydriding at high fuel burnup may degrade cladding ductility to the extent that such large ruptures are possible under severe loadings. To understand this susceptibility to failure, this study focused on mechanical testing coupled with detailed finite-element modeling and analyses. Under ring-compression-type loading at room temperature, tensile cracks form within the corrosion-induced oxide layer under elastic loading. The oxide crack then propagates into the cladding wall under additional loading with little to no measurable plastic strain, as confirmed by both experiment and analyses of plastic hoop strain in the ring. For cladding with the oxide removed prior to testing at ≤1 %/s, cracking of the underlying hydride rim comprised of circumferentially oriented hydrides occurs at low plastic hoop strain (≤3 %), whereas the finite-element analysis suggests that the base alloy with a relatively small amount of hydrides appears to fail at higher strain (>8 %). At even higher strain rates (≈400 %/s), cracking within the hydride rim occurs at near-zero ductility, but the base alloy continues to remain highly ductile. These room-temperature results indicate that the hydride rim is sensitive to strain rate, whereas the base alloy is relatively not. With the precipitation of ≈100 % radially oriented hydrides, the cladding exhibits near-zero ductility at room temperature and ≈0.1 %/s. This study suggests that the ring-compression test coupled with finite-element modeling and analysis may be used to estimate crack-initiation strains in irradiated cladding materials with susceptible microstructures and under various deformation rates.
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Category : Nuclear energy
Languages : en
Pages : 774
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Category : Government publications
Languages : en
Pages : 1118
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Category : Power resources
Languages : en
Pages : 1468
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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: J. Desquines
Publisher:
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Category : Embrittlement
Languages : en
Pages : 22
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Book Description
The PROMETRA material testing program is a support program related to the study of high burnup fuel rod behavior under Reactivity Initiated Accidents (RIA) and to the interpretation of the CABRI REP-Na RIA test results. Hoop and axial tensile tests have been performed on fresh and irradiated Zircaloy-4 cladding alloy first at CEA Grenoble hot labs and now at CEA Saclay in order to assess the cladding mechanical behavior during RIA transients. Efforts have been continuously carried out in order to improve the prototipicallity of the tests for RIA studies involving new specimens and new testing techniques. The corrosion level of irradiated specimens reached up to 130 ?m of oxide layer thickness. The influence of in-pile oxide layer spallation has also been addressed. High strain-rate material properties of irradiated Zircaloy-4 and the consequences of hydride embrittlement can be derived from the PROMETRA program.
