This paper describes specimen geometries that have proven useful in characterizing the mechanical behavior of irradiated Zircaloy components. In many cases, small plane strain-type specimens obtained from the component of interest have been chosen. Specimen fabrication techniques are discussed. Failure strains for the various types of specimens are compared and analyzed in terms of the microdeformation mechanisms that occur in irradiated Zircaloy. Finite element analyses of specimens are conducted to determine details of the strain distribution just before the initiation of plastic instability. Results of these analyses are used to determine the maximum principal plastic strain at the failure initiation point.
DIP-COATING PROCESS OF ZIRCALOY-2 FUEL CLADDING WITH COLLOIDAL GRAPHITE. The intensive researchs on high discharge burn-up of Light Water Reactor (LWR) fuel element have been continously performed due to the extension of fuel element's utility life. One of these researches was allowing for alteration of the existing zirconium-based clad system through coating. A coating technique with the coating layer thickness of 10-30 µm will improve the corrosion resistance of cladding without changing the dimension of cladding. The scope of this current research is to obtain the zircaloy-2 cladding coated with ZrC layer by dipping process of zircaloy-2 specimens in colloidal graphite at room temperature. The dip-coated specimens undergo heating process at 700 o C, 900 o C and 1100 o C respectively in Argon gas atmosphere for 1 hour and are subsequently characterized by optical microscope and XRD. The optical microscope images show that the coating layers thickness are increased as the heating temperature increased. The coating layers thickness are 10 μm, 20-40 μm and 100 μm for the specimens heated at 700 o C, 900 o C and 1100 o C respectively. The calculated diffusivity of carbon into zircaloy-2 cladding for the coated specimens at 700 o C, 900 o C and 1100 o C are 3,10216E-11 cm 2 s-1 ; 3,60479E-11 cm 2 s-1 and 4,00613E-11 cm 2 s-1 respectively. From XRD examination analysis reveals that the ZrC phase appears in the specimens heated at 1100 o C but it is not the case for specimens heated at both 700 o C and 900 o C. The coating layer of specimens heated at both 700 o C and 900 o C mostly consists of carbon. At these heating temperatures, carbon atoms have diffused into zircaloy-2 and substituted the zirconium atoms with a limited occupation to form C-Zr solid solution. At the temperature of 1100 o C, due to the increase in vibration energy, the carbon atoms have enough energy to diffuse to form the carbide phase. Heating process at higher than 700 o C, however, will degrade the zircaloy-2 cladding.. It is concluded that the dip-coating process of zircaloy-2 cladding in graphite colloid with subsequent high temperature heating is not the proper method to gain the ZrC-coated zircaloy-2 cladding. Therefore, as for the future works of this research, the others method should be searched and investigated to obtain the proper ZrC coating process on LWR zircaloy cladding which fulfills the dimension and quality requirements.
This metallurgical study of Zr-barrier fuel cladding evaluates the importance of three salient attributes: (1) metallurgical bond between the zirconium liner and the Zircaloy substrate, (2) liner thickness (roughly 10% of the total cladding wall), and (3) softness (purity). The effect that each of these attributes has on the pellet-cladding interaction (PCI) resistance of the Zr-barrier fuel was studied by a combination of analytical model calculations and laboratory experiments using an expanding mandrel technique. Each of the attributes is shown to contribute to PCI resistance. The effect of the zirconium liner on fuel behavior during off-normal events in which steam comes in contact with the zirconium surface was studied experimentally. Simulations of loss-of-coolant accident (LOCA) showed that the behavior of Zr-barrier cladding is virtually indistinguishable from that of conventional Zircaloy cladding. If steam contacts the zirconium liner surface through a cladding perforation and the fuel rod is operated under normal power conditions, the zirconium liner is oxidized more rapidly than is Zircaloy, but the oxidation rate returns to the rate of Zircaloy oxidation when the oxide phase reaches the zirconium-Zircaloy metallurgical bond.
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