Papers by Keyword: Zircaloy-4

Abstract: In this paper, a series of bending cyclic tests under stress controlled were conducted at room temperature on Zircaloy-4 (Zr-4) to investigate its bending ratcheting behavior. The effects of mean stress and stress amplitude on the bending ratcheting behavior were experimentally studied, respectively. The experimental results show that the ratcheting strain of the material is very sensitive to mean stress and stress amplitude. It can be concluded that ratcheting strain level increases with increasing mean stress and stress amplitude.

Abstract: Zirconium-based dissolver vessels containing highly radioactive and concentrated corrosive nitric acid solution needs to be joined to the rest of fuel reprocessing plant made of stainless steel(SS), which demands high integrity and corrosion resistant dissimilar joints. Vacuum brazing joining process was proposed in the present work to join zircaloy-4 and 1Cr18Ni9Ti SS since fusion welding processes produce brittle intermetallic precipitates at the interface which reduce the mechanical strength as well as the corrosion resistance of the joint. It is observed that Ag-Cu eutectic phase structure was formed in the brazing seam, a reaction layer exhibited in the interface between zircaloy-4/Ag-Cu-Ti. The concentration of the dendrites increases with an increase in bonding temperature. The width of the reaction layer increases with the increase of the brazing temperature and holding time.

Abstract: During power transient conditions in nuclear reactors, uranium oxide pellets expand and crack due to the increase in temperature and their poor thermal conductivity. Moreover, the cladding undergoes creep because of the external pressure, and its diameter shortens. These antagonistic phenomena lead to the establishment of a contact between the pellet and the cladding, called the pellet-cladding interaction. The synergistic effect of the hoop tensile stress and strain imposed on the cladding by fuel thermal expansion and corrosion by iodine released from the UO₂ fuel as a fission product at the same time can lead to Iodine-induced Stress Corrosion Cracking (I-SCC) of the Zircaloy-4 cladding. I-SCC failures of zirconium alloys are usually described in three steps: initiation of cracks, intergranular subcritical propagation, and critical propagation with a brittle transgranular propagation mode [1]. Transgranular propagation occurs as soon as the stress intensity factor overshoots a threshold value K_I,SCC. It is the critical step and leads to the final ductile failure of the cladding. Transgranular cracks propagate by cleavage-like fracture on basal planes of the hexagonal lattice and fluting; it is the result of a competition between a plastic accommodation of the applied strain and the brittle fracture of basal planes by iodine assisted cleavage.

Abstract: The mechanical behaviour and texture evolution during uniaxial compression of Zircaloy-4 at different temperatures (25, 300, 500 C) has been studied. At room temperature and 300 C the texture evolution and strain-hardening behaviour observed are attributed to the activation of {10-12} tensile twinning, which can be identified in optical micrographs and electron backscatter diffraction (EBSD) data. The influence of twinning upon the texture evolution and hardening rate becomes less apparent with increasing temperature. Nevertheless twinning is still active at 500 C. Simulation of the texture evolution at 500 C using crystal plasticity finite element modelling (CPFEM) indicates that slip alone cannot explain the experimentally observed textures at this temperature.

Abstract: The oxidation of γ-Zr(Fe,Cr)2 intermetallic particles during the thermal exposition of Zircaloy-4 at 470°C in oxygen was investigated with PhotoElectroChemical techniques (PEC). Via the measurement of bandgap, haematite Fe2O3 (2.2 eV), rhomboedric solid solution (FexCr1-x)2O3 (2.6 eV) and chromia Cr2O3 (3.0 eV) phases were identified as components of oxidised particles. Evolution of size, lateral distribution and density of these particles was studied in function of zirconia scale thickness. During the first stage of oxidation, the density of oxidised particles increased with thickness but decreased during a second stage, highlighting in an innovative way the phenomenon of haematite and chromia dissolution in the zirconia matrix. It is concluded that PEC techniques represent a sensitive and powerful way to locally analyse the various semiconductor phases in an oxide scale at the micron scale.

Abstract: This work was carried out to obtain sound welds and to select the most suitable binary metal joint among three different dissimilar metal combinations such as Zr-4/Ta, Mo/Ta and Ti/Ta (seal tube/sensor sheath) joints for an instrumented nuclear fuel irradiation test. To do this, the Taguchi experimental method was employed to optimize the experimental data. In addition, metallography, micro-focus x-ray radiography and a hardness test were conducted to examine the welds. From the weld bead appearance, penetration depth and bead width as well as the weld defects standpoint, the Zr-4/Ta joint is suggested for a circumferential joining between a seal tube and a sensor sheath. The optimized welding parameters based on the Zr-4/Ta joint are suggested as well.

Abstract: The formation of hydrides in zirconium alloy has been one of the essential matters of discussion to maintain mechanical strength of nuclear fuel cladding tubes. In this work, we examined the precipitation process of zirconium hydride by transmission electron microscopy under hydrogen ion irradiation. Zircaloy-4, which has been used extensively as nuclear fuel cladding, was irradiated with hydrogen ion at room temperature to achieve enough hydrogen concentration for precipitation. The growth of hydrides accompanied with dislocations around hydrides was observed under hydrogen implantation. The observed hydride was the γ-hydride phase with fct structure and the orientation relationship was <110>γ ||<1120>α as reported previously. As the hydride grew, the dislocations were generated gradually. This process can be explained using a ratchet mechanism suggested by Carpenter. The growth rate became lower according to the approach of other hydrides. This behavior is considered to be influenced by the strain field caused by other hydrides.

Abstract: Neutron incoherent scattering measurements were conducted on Zircaloy-4 round bars. The specimens were charged in a tube furnace at 430 °C, using a 12.5 vol. % hydrogen in an argon mixture for 30, 60, and 90 minutes at 13.8 kPa pressure. The volume-average neutron diffraction measurements showed the presence of the face-centered-cubic delta zirconium hydride (δ-ZrH2) phase in the hydrogenated specimens. The assessment of the background in the diffraction profiles due to the incoherent scattering from the hydrogen atoms was carried out by performing inelastic scans around zero energy transfer and at a fixed two-theta value for which there was only flat background and no coherent scattering. To estimate the relative amount of hydrogen in the Zircaloy-4 samples, the increase in incoherent scattering intensities with hydrogen content was calibrated using samples for which the hydrogen content was known. Measurement of the background scattering from locations within the round bar was also performed to map the distribution of hydrogen content.

Abstract: In order to study the nucleation and growth of cracks in the outer oxide scale which expand into the underlying alloy, deformation in creep in oxygen or in vacuum of zirconium and Zircaloy-4 has been studied mainly at 500°C. Influence of applied stresses, atmosphere and alloy’s grade on the deformation and oxidation processes are especially analyzed. The results underline the presence of two distinct deformation domains for both alloys grades, depending on the applied stress value. The presence of the oxide scale leads only to slight modifications on the deformation mechanism but it induces an increase of the deformation rate. This enhancement is especially observed in the case of the pre-oxidized Zircaloy-4 whose cracks remain mainly located in the outer part of the oxide. In opposite, the pre-oxidized zirconium shows cracks located down to the underlying metal. Acoustic emission is used to follow, in situ, in temperature the damage process of the outer zirconia layer during creep, and precisions about the oxidation mechanism and the effect of applied stress on oxygen diffusion and oxide growth rate are obtained thanks to the use of 18O as a marker.

Abstract: Zr-4 alloy is the material of nuclear fuel shell in nuclear power plant’s PWR. This paper presents this material’s general mechanical property and fatigue behavior that are tested in accordance with ASTM in room temperature and 380°C condition. The test results show that there is no cyclic hardening or cyclic softening phenomena for Zr-4 alloy applied by cyclic loading in room temperature condition. The fatigue design curve is obtained by processing fatigue test results with adopting ASME Sec.Ⅲ based on the test results of strain fatigue property. The research result shows the fatigue design data at different temperature may be corrected by elastic modulus with room temperature curve. This paper’s result may be used in PWR component design.