Modeling of Zircaloy Cladding Degradation Under Repository Conditions PDF Download
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Author: Suresh Raghavan
Publisher:
ISBN:
Category : Nuclear fuel claddings
Languages : en
Pages : 138
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Book Description
Author: Suresh Raghavan
Publisher:
ISBN:
Category : Nuclear fuel claddings
Languages : en
Pages : 138
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Book Description
Author: Lakshman Santanam
Publisher:
ISBN:
Category : Metals
Languages : en
Pages : 28
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Book Description
Two potential degradation mechanisms, creep and stress corrosion cracking, of Zircaloy cladding during repository storage of spent nuclear fuel have been investigated. The deformation and fracture map methodology has been used to predict maximum allowable initial storage temperatures to achieve a thousand year life without rupture as a function of spent-fuel history. A stress analysis of fuel rods has been performed. Stresses in the outer zirconium oxide layer and the inner Zircaloy tube have been predicted for typical internal pressurization, oxide layer thickness, volume expansion from formation of the oxide layer and thermal expansion coefficients of the cladding and oxide. Stress relaxation occurring in-reactor has also been taken into account. The calculations indicate that for the anticipated storage conditions investigated, the outer zirconium oxide layer is in a state of compression thus making it unlikely that stress corrosion cracking of the exterior surface will occur.
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 28
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Book Description
Two potential degradation mechanisms, creep and stress corrosion cracking, of Zircaloy cladding during repository storage of spent nuclear fuel have been investigated. The deformation and fracture map methodology has been used to predict maximum allowable initial storage temperatures to achieve a thousand year life without rupture as a function of spent-fuel history. A stress analysis of fuel rods has been performed. Stresses in the outer zirconium oxide layer and the inner Zircaloy tube have been predicted for typical internal pressurization, oxide layer thickness, volume expansion from formation of the oxide layer and thermal expansion coefficients of the cladding and oxide. Stress relaxation occurring in-reactor has also been taken into account. The calculations indicate that for the anticipated storage conditions investigated, the outer zirconium oxide layer is in a state of compression thus making it unlikely that stress corrosion cracking of the exterior surface will occur. 20 refs., 6 figs., 9 tabs.
Author: Lakshman Santanam
Publisher:
ISBN:
Category : Nuclear fuel claddings
Languages : en
Pages : 218
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Book Description
Author: Lakshman Santanam
Publisher:
ISBN:
Category : Metals
Languages : en
Pages : 14
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Book Description
Creep, a potential degradation mechanism of Zircaloy cladding after repository disposal of spent nuclear fuel, has been investigated. The deformation and fracture map methodology has been used to predict maximum allowable initial storage temperatures to achieve a thousand year life without rupture as a function of spent-fuel history. Maximum allowable temperatures are 340°C (613 K) for typically stressed rods (70--100 MPa) and 300°C (573 K) for highly stressed rods (140--160 MPa).
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 16
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Book Description
Creep, a potential degradation mechanism of Zircaloy cladding after repository disposal of spent nuclear fuel, has been investigated. The deformation and fracture map methodology has been used to predict maximum allowable initial storage temperatures to achieve a thousand year life without rupture as a function of spent-fuel history. Maximum allowable temperatures are 340°C (613 K) for typically stressed rods (70--100 MPa) and 300°C (573 K) for highly stressed rods (140--160 MPa). 10 refs., 2 figs.
Author:
Publisher:
ISBN:
Category : Power resources
Languages : en
Pages : 676
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Book Description
Author:
Publisher:
ISBN:
Category : Power resources
Languages : en
Pages : 876
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Book Description
Author: Harry D. Smith
Publisher:
ISBN:
Category : Boiling water reactors
Languages : en
Pages : 22
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Book Description
This paper reports the results of an experiment designed to detect the influence of copper on Zircaloy spent fuel cladding degradation in one possible repository environment. Copper and copper alloys are being considered for use in a tuff repository. The compatibility of a copper waste package container and the Zircaloy cladding on spent fuel has been questioned essentially because copper ion has been observed to accelerate zirconium alloy corrosion in acid environments, as does ferric iron, and a phenomenon called "crud-induced localized corrosion" is observed in some Boiling Water Reactors where through-the-wall corrosion pits develop beneath copper-rich crud deposits.
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ISBN:
Category :
Languages : en
Pages : 26
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Book Description
This paper reports the results of an experiment designed to detect the influence of copper on Zircaloy spent fuel cladding degradation in one possible repository environment. Copper and copper alloys are being considered for use in a tuff repository. The compatibility of a copper waste package container and the Zircaloy cladding on spent fuel has been questioned essentially because copper ion has been observed to accelerate zirconium alloy corrosion in acid environments, as does ferric iron, and a phenomenon called "crud-induced localized corrosion" is observed in some Boiling Water Reactors where thorugh-the-wall corrosion pits develop beneath copper-rich crud deposits. 16 refs., 6 figs., 2 tabs.
