Fracture of Zircaloy Cladding by Interactions with Uranium-Dioxide Pellets in Water Reactor Fuel Rods PDF Download
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Author: E. Smith
Publisher:
ISBN:
Category : Fuel-cladding bonding
Languages : en
Pages : 14
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The paper summarizes the main features of a detailed theoretical study of Zircaloy cladding fracture that is caused by the growth of stress corrosion cracks which form near fuel pellet cracks as a consequence of a power increase after a sufficiently high burn up. It is shown that unless stress corrosion crack growth in irradiated Zircaloy is able to proceed at very low stress intensifications, fracture is unlikely when there are uniform friction effects at the fuel-cladding interface, when the interfacial friction coefficient is less than unity, when a symmetric distribution of fuel cracks exists, and when symmetric interfacial slippage occurs (that is, "uniform" conditions). Otherwise, the observed fuel rod failures must be due to departures from uniform conditions, and this investigation shows that a very high interfacial friction coefficient, and particularly fuel-cladding bonding, are means of providing sufficient stress intensification at a cladding crack tip to explain the occurrence of cladding fractures.
Author: E. Smith
Publisher:
ISBN:
Category : Fuel-cladding bonding
Languages : en
Pages : 14
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Book Description
The paper summarizes the main features of a detailed theoretical study of Zircaloy cladding fracture that is caused by the growth of stress corrosion cracks which form near fuel pellet cracks as a consequence of a power increase after a sufficiently high burn up. It is shown that unless stress corrosion crack growth in irradiated Zircaloy is able to proceed at very low stress intensifications, fracture is unlikely when there are uniform friction effects at the fuel-cladding interface, when the interfacial friction coefficient is less than unity, when a symmetric distribution of fuel cracks exists, and when symmetric interfacial slippage occurs (that is, "uniform" conditions). Otherwise, the observed fuel rod failures must be due to departures from uniform conditions, and this investigation shows that a very high interfacial friction coefficient, and particularly fuel-cladding bonding, are means of providing sufficient stress intensification at a cladding crack tip to explain the occurrence of cladding fractures.
Author: E. Smith
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
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Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :
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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: ASTM International
ISBN:
Category :
Languages : en
Pages : 681
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Author: D. Cubicciotti
Publisher:
ISBN:
Category : Fission-product iodine
Languages : en
Pages : 21
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A model for pellet-cladding interaction (PCI) fracture of light-water reactor (LWR) fuel rods is presented, the basis of which is that Zircaloy cladding fails by iodine stress corrosion cracking (SCC). Laboratory data on iodine SCC of irradiated Zircaloy provide the primary input to the model, but unirradiated Zircaloy SCC data and theoretical analyses are utilized to broaden the regime of validity to encompass the various power reactor observations.
Author: Bernardo N. Nobrega
Publisher:
ISBN:
Category : Nuclear fuel claddings
Languages : en
Pages : 270
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Author:
Publisher:
ISBN:
Category : Nuclear energy
Languages : en
Pages : 640
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Author: United States. Energy Research and Development Administration
Publisher:
ISBN:
Category : Medicine
Languages : en
Pages : 1104
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Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :
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For a better understanding of Zircaloy fuel-rod failure by the pellet-cladding interaction (PCI) phenomenon, a mechanistic study of deformation and fracture behavior of spent power reactor fuel cladding under simulated PCI conditions was conducted. Zircaloy-2 cladding specimens, obtained from fuel assemblies of operating power reactors, were deformed to fracture at 325°C by internal gas pressurization in the absence of fission product simulants. Fracture characteristics and microstructures were examined via SEM, TEM, and HVEM. Numerous dislocation tangles and cell structures, observed in TEM specimens of cladding tubes that failed in a ductile manner, were consistent with SEM observations of a limited number of dimples characteristic of microvoid coalescence. A number of brittle-type failures were produced without the influence of fission product simulants. The brittle cracks occurred near the areas compressed by the Swagelok fittings of the internally pressurized tube and propagated from the outer to the inner surface. Since the outer surface was isolated and maintained under a flowing stream of pure helium, it is unlikely that the brittle-type failure was influenced by any fission product traces. SEM fractography of the brittle-type failure revealed a large area of transgranular pseudocleavage with limited areas of ductile fluting, which were similar in appearance to the surfaces produced by in-reactor PCI-type failures. A TEM evaluation of the cladding in the vicinity of the through-wall crack revealed numerous locations that contained an extensive amount of second-phase precipitate (Zr3O). We believe that the brittle-type failures of the irradiated spent fuel cladding in the stress rupture experiments are associated with segregation of oxygen, which leads to the formation of the order structure, an immobilization of dislocations, and minimal plastic deformation in the material.
Author:
Publisher:
ISBN:
Category : Nuclear reactors
Languages : en
Pages : 700
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