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Author: Harry D. Smith Publisher: ISBN: Category : Boiling water reactors Languages : en Pages : 22
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.
Author: Harry D. Smith Publisher: ISBN: Category : Boiling water reactors Languages : en Pages : 22
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.
Author: Publisher: ISBN: Category : Languages : en Pages : 26
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.
Author: A. J. Rothman Publisher: ISBN: Category : Nuclear fuel claddings Languages : en Pages : 49
Book Description
A literature review and analysis were made of corrosion and degradation processes applicable to Zircaloy cladding on spent nuclear fuel in a tuff repository. In particular, lifetime sought for the Zircaloy is 10,000 years. Among the potential failure mechanisms examined were: oxidation by steam, air, and water, including the effects of ions whose presence is anticipated in the water; mechanical overload; stress (creep) rupture; stress-corrosion cracking (SCC); and delayed failure due to hydride cracking. The conclusion is that failure due to oxidation is not credible, although a few experiments are suggested to confirm the effect of aqueous fluoride on the Zircaloy cladding. Mechanical overload is not a problem, and failure from stress-rupture does not appear likely based on a modified Larson-Miller analysis. Analysis shows that delayed hydride cracking is not anticipated for the bulk of spent fuel pins. However, for a minority of pins under high stress, there is some uncertainty in the analysis as a result of: (1) uncertainty about crack depths in spent fuel claddings and (2) the effect of slow cooling on the formation of radially oriented hydride precipitates. Experimental resolution is called for. Finally, insufficient information is currently available on stress-corrosion cracking. While evidence is presented that SCC failure is not likely to occur, it is difficult to demonstrate this conclusively because the process is not clearly understood and data are limited. Further experimental work on SCC susceptibility is especially needed.
Author: Publisher: ISBN: Category : Languages : en Pages : 53
Book Description
A literature review and analysis were made of corrosion and degradation processes applicable to Zircaloy cladding on spent nuclear fuel in a tuff repository. In particular, lifetime sought for the Zircaloy is 10,000 years. Among the potential failure mechanisms examined were: oxidation by steam, air, and water, including the effects of ions whose presence is anticipated in the water; mechanical overload; stress (creep) rupture; stress-corrosion cracking (SCC); and delayed failure due to hydride cracking. The conclusion is that failure due to oxidation is not credible, although a few experiments are suggested to confirm the effect of aqueous fluoride on the Zircaloy cladding. Mechanical overload is not a problem, and failure from stress-rupture does not appear likely based on a modified Larson-Miller analysis. Analysis shows that delayed hydride cracking is not anticipated for the bulk of spent fuel pins. However, for a minority of pins under high stress, there is some uncertainty in the analysis as a result of: (1) uncertainty about crack depths in spent fuel claddings and (2) the effect of slow cooling on the formation of radially oriented hydride precipitates. Experimental resolution is called for. Finally, insufficient information is currently available on stress-corrosion cracking. While evidence is presented that SCC failure is not likely to occur, it is difficult to demonstrate this conclusively because the process is not clearly understood and data are limited. Further experimental work on SCC susceptibility is especially needed.
Author: Lakshman Santanam Publisher: ISBN: Category : Metals Languages : en Pages : 14
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
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 : Languages : en Pages : 28
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: Harry D. Smith
Author: Harry D. Smith
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Author: A. J. Rothman
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Author: Lakshman Santanam
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