Zirconium alloy

About: Zirconium alloy is a research topic. Over the lifetime, 6548 publications have been published within this topic receiving 78954 citations. The topic is also known as: zircaloy.

Effect of Water Chemistry and Composition on Microstructural Evolution of Oxide on Zr Alloys

Abstract: The microstructure of oxide films formed on Zircaloy-4 and Alloy No. 3, which has a composition similar to ZIRLO™, was investigated by high resolution transmission and scanning electron microscopy, and by scanning probe microscopy after corrosion tests performed at 360°C/18.6 MPa in deionized water or lithiated water with 0.01 M LiOH. The microstructural evolution of the oxide films was analyzed by comparing the microstructure at different depths in the oxide layer. The defects, consisting of vacancies and interstitials, such as points, lines, planes, and volumes, were produced during the oxide growth. Monoclinic, tetragonal, cubic, and amorphous phases were detected and their coherent relationships were identified. The characteristic of oxide with such microstructure had an internal cause, and the temperature and time were the external causes that induced the microstructural evolution during the corrosion process. The diffusion, annihilation, and condensation of vacancies and interstitials under the action of stress, temperature, and time caused stress relaxation and phase transformation. It was observed, in the middle of the oxide layer, that the vacancies absorbed by grain boundaries formed pores to weaken the bonding strength between grains. Pores formed under compressive stress lined up along the direction parallel to the compressive stress. Thus, cracks developed from the pores were parallel to the oxide/metal interface. Li+ and OH− incorporated in oxide films were adsorbed on the wall of pores or entered into vacancies to reduce the surface free energy of the zirconium oxide during exposure in lithiated water. As a result, the diffusion of vacancies and the formation of pores were enhanced, inducing the degradation of the corrosion resistance. The relationship between the corrosion resistance of zirconium alloys and the microstructural evolution of oxide films affected by water chemistry and composition is also discussed.

Thermal Conductivity and Heat Capacity of Zircaloy-2, -4 and Unalloyed Zirconium

Abstract: (1975). Thermal Conductivity and Heat Capacity of Zircaloy-2, −4 and Unalloyed Zirconium. Journal of Nuclear Science and Technology: Vol. 12, No. 10, pp. 661-662.

Electrochemical studies on the stability and corrosion resistance of new zirconium‐based alloys for biomedical applications

Abstract: The biocompatibility of commercially pure zirconium and its alloys is closely related to their surface properties with both the composition of the protecting oxide film and the surface topography playing an important role. This article is a study of corrosion behavior of new zirconium alloys for orthopedic implants, which are supposed to be used instead of some implant materials that have a higher citotoxicity. For this reason, zirconium and its alloys will be employed widely in biomedical applications. The higher stability and corrosion resistance exhibited by zirconium are due to the spontaneous formation of a passive zirconium oxide film, which protects the metal from further oxidation. Potentiodynamic polarization and electrochemical impedance spectroscopy (EIS) are used as electrochemical techniques. Measurements were carried out to investigate the corrosion behavior of zirconium and Zr2.5Nb, Zr3Ta, and Zr2.5Nb3Ta alloys in aerated Hank solution at 37 ± 0.2 °C. The results of EIS were compared with those obtained by potentiodynamic polarization techniques. Impedance spectra were represented both in complex impedance diagram (Nyquist plot) and Bode plots. The EIS measurements have confirmed this stability range and pointed out the formation of oxide layers on the electrode surfaces.

Development of LWR Fuels with Enhanced Accident Tolerance

Abstract: Significant progress was made on the technical, licensing, and business aspects of the Westinghouse Electric Company’s Enhanced Accident Tolerant Fuel (ATF) by the Westinghouse ATF team. The fuel pellet options included waterproofed U15N and U3Si2 and the cladding options SiC composites and zirconium alloys with surface treatments. Technology was developed that resulted in U3Si2 pellets with densities of >94% being achieved at the Idaho National Laboratory (INL). The use of U3Si2 will represent a 15% increase in U235 loadings over those in UO₂ fuel pellets. This technology was then applied to manufacture pellets for 6 test rodlets which were inserted in the Advanced Test Reactor (ATR) in early 2015 in zirconium alloy cladding. The first of these rodlets are expected to be removed in about 2017. Key characteristics to be determined include verification of the centerline temperature calculations, thermal conductivity, fission gas release, swelling and degree of amorphization. Waterproofed UN pellets have achieved >94% density for a 32% U3Si2/68% UN composite pellet at Texas A&M University. This represents a U235 increase of about 31% over current UO2 pellets. Pellets and powders of UO2, UN, and U3Si2the were tested by Westinghouse and Los Alamos National Laboratory (LANL) using differential scanning calorimetrymore » to determine what their steam and 20% oxygen corrosion temperatures were as compared to UO2. Cold spray application of either the amorphous steel or the Ti2AlC was successful in forming an adherent ~20 micron coating that remained after testing at 420°C in a steam autoclave. Tests at 1200°C in 100% steam on coatings for Zr alloy have not been successful, possibly due to the low density of the coatings which allowed steam transport to the base zirconium metal. Significant modeling and testing has been carried out for the SiC/SiC composite/SiC monolith structures. A structure with the monolith on the outside and composite on the inside was developed which is the current baseline structure and a SiC to SiC tube closure approach. Permeability tests and mechanical tests were developed to verify the operation of the SiC cladding. Steam autoclave (420°C), high temperature (1200°C) flowing steam tests and quench tests were carried out with minimal corrosion, mechanical or hermeticity degradation effect on the SiC cladding or end plug closure. However, in-reactor loop tests carried out in the MIT reactor indicated an unacceptable degree of corrosion, likely due to the corrosive effect of radiolysis products which attacked the SiC.« less