Predictions of Dry Storage Behavior of Zircaloy Clad Spent Fuel Rods Using Deformation and Fracture Map Analyses PDF Download
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Author: Publisher: ISBN: Category : Languages : en Pages :
Book Description
Predictions of the maximum initial allowable temperature required to achieve a 40-year life in dry storage are made for Zircaloy clad spent fuel. Maximum initial dry storage temperatures of 420°C for 1 year fuel cladding subjected to a constant stress of 70 MPa are predicted. The technique utilized in this work is based on the deformation and fracture map methodology. Maps are presented for temperatures between 50 and 850°C stresses between 5 and 500 MPa. These maps are combined with a life fraction rule to predict the time to rupture of Zircaloy clad spent Light Water Reactor (LWR) fuel subjected to various storage conditions.
Author: Publisher: ISBN: Category : Languages : en Pages :
Book Description
Predictions of the maximum initial allowable temperature required to achieve a 40-year life in dry storage are made for Zircaloy clad spent fuel. Maximum initial dry storage temperatures of 420°C for 1 year fuel cladding subjected to a constant stress of 70 MPa are predicted. The technique utilized in this work is based on the deformation and fracture map methodology. Maps are presented for temperatures between 50 and 850°C stresses between 5 and 500 MPa. These maps are combined with a life fraction rule to predict the time to rupture of Zircaloy clad spent Light Water Reactor (LWR) fuel subjected to various storage conditions.
Author: Publisher: ISBN: Category : Languages : en Pages :
Book Description
The Zircaloy cladding of water reactor fuel rods is susceptible to local breach-type failure, commonly known as pellet-cladding interaction (PCI) failure, during operational and off-normal power transients after the fuel has achieved a sufficiently high burnup. An optimization of power ramp procedures or fuel rod fabrication to minimize the cladding failure would result in a significant decrease in radiation exposure of plant personnel due to background and airborne radioactivity as well as an extension of core life in terms of allowable off-gas radioactivity. As part of a program to provide a better understanding of the fuel rod faiure phenomenon and to facilitate the formulation of a better failure criterion, a mechanistic study of the deformation and fracture behavior of high-burnup spent-fuel cladding is in progress under simulated PCI conditions.
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: Lakshman Santanam Publisher: ISBN: Category : Metals 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.
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.
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Author: Lakshman Santanam
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Author: Lakshman Santanam
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