Mechanical Property Testing of Irradiated Zircaloy Cladding Under Reactor Transient Conditions PDF Download
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Category :
Languages : en
Pages : 16
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Book Description
Specimen geometries have been developed to determine the mechanical properties of irradiated Zircaloy cladding subjected to the mechanical conditions and temperatures associated with reactivity-initiated accidents (RIA) and loss-of-coolant accidents (LOCA). Miniature ring-stretch specimens were designed to induce both uniaxial and plane-strain states of stress in the transverse (hoop) direction of the cladding. Also, longitudinal tube specimens were also designed to determine the constitutive properties in the axial direction. Finite-element analysis (FEA) and experimental parameters and results were closely coupled to optimize an accurate determination of the stress-strain response and to induce fracture behavior representative of accident conditions. To determine the constitutive properties, a procedure was utilized to transform measured values of load and displacement to a stress-strain response under complex loading states. Additionally, methods have been developed to measure true plastic strains in the gauge section and the initiation of failure using real-time data analysis software. Strain rates and heating conditions have been selected based on their relevance to the mechanical response and temperatures of the cladding during the accidents.
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 16
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Book Description
Specimen geometries have been developed to determine the mechanical properties of irradiated Zircaloy cladding subjected to the mechanical conditions and temperatures associated with reactivity-initiated accidents (RIA) and loss-of-coolant accidents (LOCA). Miniature ring-stretch specimens were designed to induce both uniaxial and plane-strain states of stress in the transverse (hoop) direction of the cladding. Also, longitudinal tube specimens were also designed to determine the constitutive properties in the axial direction. Finite-element analysis (FEA) and experimental parameters and results were closely coupled to optimize an accurate determination of the stress-strain response and to induce fracture behavior representative of accident conditions. To determine the constitutive properties, a procedure was utilized to transform measured values of load and displacement to a stress-strain response under complex loading states. Additionally, methods have been developed to measure true plastic strains in the gauge section and the initiation of failure using real-time data analysis software. Strain rates and heating conditions have been selected based on their relevance to the mechanical response and temperatures of the cladding during the accidents.
Author: H. Tsai
Publisher:
ISBN:
Category : Axial specimens
Languages : en
Pages : 16
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Book Description
Specimen geometries have been developed to determine the mechanical properties of irradiated Zircaloy cladding subjected to the mechanical conditions and temperatures associated with reactivity-initiated accidents (RIA) and loss-of-coolant accidents (LOCA). Miniature ring-stretch specimens were designed to induce both uniaxial and plane-strain states of stress in the transverse (hoop) direction of the cladding. Also, longitudinal tube specimens were also designed to determine the constitutive properties in the axial direction. Finite-element analysis (FEA) and experimental parameters and results were closely coupled to optimize an accurate determination of the stress-strain response and to induce fracture behavior representative of accident conditions. To determine the constitutive properties, a procedure was utilized to transform measured values of load and displacement to a stress-strain response under complex loading states. Additionally, methods have been developed to measure true plastic strains in the gauge section and the initiation of failure using real-time data analysis software. Strain rates and heating conditions have been selected based on their relevance to the mechanical response and temperatures of the cladding during the accidents.
Author: LM. Lowry
Publisher:
ISBN:
Category : Burst pressure
Languages : en
Pages : 16
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Book Description
The Nuclear Regulatory Commission supported an experimental program to study the mechanical properties of both unirradiated and irradiated fuel rod cladding. The program was designed to produce a mechanical property data base for use in developing modeling codes which could then be used to predict the performance of Zircaloy-4 clad fuel rods under various reactor loss-of-coolant accident (LOCA) conditions. Transient-heating burst tests were conducted at Argonne National Laboratory, Battelle Columbus Laboratories, and Oak Ridge National Laboratory as part of that program. A brief description of the testing methods, specimens, equipment, and procedures illustrates the similarities and differences in tests conducted at each laboratory. The data obtained from tests conducted in steam and utilizing test specimens with internal heaters or specimens containing alumina mandrels or pellets were compared for heating rates of 5, 28, and 55 C/sec. Both burst pressure and circumferential failure strain data versus burst temperature were plotted and compared in the range of 650 to 1250 C. It was found that testing methods, specimen oxidation layer, and irradiation effected the transient-heating burst test data.
Author: AA. Bauer
Publisher:
ISBN:
Category : Annealing
Languages : en
Pages : 10
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Book Description
A program is being conducted to determine the mechanical behavior of irradiated Zircaloy-4 fuel rod cladding subjected to anneals under temperature and time conditions that simulate various off-normal and transient reactor operating conditions. This paper reports the initial phase of the program, which is a study of changes in hardness and tensile properties that accompany isothermal and transient anneals. A special transient annealing apparatus was built to subject specimens to the desired temperature/time conditions. Hardness and tensile strength decreased with increase in the maximum annealing temperature from 800°F (425°C) to 1300°F (705°C). Changes above 1100°F (595°C) were related to recrystallization and grain growth.
Author:
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ISBN:
Category : Nuclear fuel claddings
Languages : en
Pages : 40
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Author: U.S. Nuclear Regulatory Commission
Publisher:
ISBN:
Category : Nuclear energy
Languages : en
Pages : 40
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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:
Publisher:
ISBN:
Category : Metals
Languages : en
Pages : 452
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Book Description
Author: Larry M. Lowry
Publisher:
ISBN:
Category : Metals
Languages : en
Pages : 68
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Author:
Publisher: ASTM International
ISBN:
Category : Metals
Languages : en
Pages : 491
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