Showing papers on "Zirconium alloy published in 1991"

Zirconium in the nuclear industry

Abstract: This book contains papers from the 9th International Symposium on Zirconium in the Nuclear Industry. Zirconium alloys, especially Zircaloy-2, Zircaloy-4, and Zr-2.5Nb, have been the main materials used in the core of nuclear power plants. Because of the recent trends in the nuclear industry towards higher fuel discharge burnups for better fuel utilization, and higher heat ratings and coolant temperatures for better thermal efficiency, interest in zirconium alloys has increased. It follows that corrosion-a topic of concern-has high coverage in this book. Other topics discussed are mechanical properties of zirconium alloys, and alloy fabrication.

Zirconium oxide and nitride coated prosthesis for wear and corrosion resistance

Abstract: Orthopedic implants of zirconium or zirconium-based alloy coated with blue or blue-black zirconium oxide or zirconium nitride to provide low friction, highly wear resistant coatings especially useful in artificial joints, such as hip joints, knee joints, elbows, etc. The invention zirconium oxide or nitride coated prostheses are also useful in that the coatings provide a barrier against implant corrosion caused by ionization of the metal prosthesis. Such protection can be extended by the use of oxidized or nitrided porous coatings of zirconium or zirconium alloy beads or wire mesh into which bone spicules may grow so that the prosthesis may be integrated into the living skeleton.

Kinetics of Formation of a Platelet-Reinforced Ceramic Composite Prepared by the Directed Reaction of Zirconium with Boron Carbide

Abstract: In this paper the kinetics of formation of a new class of ceramic composite material, zirconium diboride platelet-reinforced zirconium carbides, are discussed. These materials are prepared by the directed reaction of molten zirconium with boron carbide to form a ceramic material composed of zirconium diboride platelets approximately uniformly distributed in a zirconium carbide matrix containing a controlled amount of residual zirconium metal. Results from interrupted growth studies, differential thermal analysis, adiabatic reaction temperature calculations, and kinetic measurements have been used to study the kinetics of the process. The reaction is very fast and proceeds parabolically with time with a rate constant between 1.6 {times} 10{sup {minus}2} and 3.9 {times} 10{sup {minus}2} cm{sup 2}/s. The proposed mechanism suggests that when molten zirconium contacts boron carbide, the molten zirconium exothermically reacts with the boron carbide to form a boron-rick liquid. Further reaction is sustained by the continuous dissolution of the boron carbide as the boron-rich liquid is drawn into the boron carbide. The product is a zirconium diboride/zirconium carbide/zirconium composite which homogenizes quickly at the reaction temperature to yield a uniform product microstructure throughout the composite. Two alternative rate-limiting steps are discussed and the implications of each are explored.

The effect of copper, chromium, and zirconium on the microstructure and mechanical properties of Al-Zn-Mg-Cu alloys

Abstract: The present study evaluates the effect of the systematic variation of copper, chromium, and zirconium contents on the microstructure and mechanical properties of a 7000-type aluminum alloy. Fracture toughness and tensile properties are evaluated for each alloy in both the peak aging, T8, and the overaging, T73, conditions. Results show that dimpled rupture essentially characterize the fracture process in these alloys. In the T8 condition, a significant loss of toughness is observed for alloys containing 2.5 pct Cu due to the increase in the quantity of Al-Cu-Mg-rich S-phase particles. An examination of T8 alloys at constant Cu levels shows that Zr-bearing alloys exhibit higher strength and toughness than the Cr-bearing alloys. In the T73 condition, Cr-bearing alloys are inherently tougher than Zr-bearing alloys. A void nucleation and growth mechanism accounts for the loss of toughness in these alloys with increasing copper content.

Magnetic properties of ZrCo5.1−xFex alloys

Abstract: The results of magnetic measurements performed on ZrCo5.1−xFex alloys in the temperature range 4.2–900 K and fields up to 7 T are reported. The presence of solid solutions is evidenced in the composition range x<0.9. The Curie temperature and the saturation magnetization increase nearly linearly when cobalt is substituted by iron. In addition, the anisotropy field decreases and shows an unusual temperature dependence, with a maximum at T≂300 K.

Brittle fracture induced by hydrides in zircaloy-4

Abstract: Zircaloy-4 is used as a cladding material in the nuclear industry for fuel elements. Its mechanical properties can be drastically affected by the presence of hydrides, which form when hydrogen content exceeds the terminal solid solubility. This change often manifests itself as a reduction in ductility (elongation and reduction in area), coupled with the evolution of the fracture mode from ductile microvoid nucleation and coalescence to intergranular fracture. It has been found, at room temperature, that Zircaloy-4 undergoes a ductile to brittle transition when the hydrogen content (hydride volume fraction) in the specimen is higher than some critical value depending on the microstructure and the hydride morphology. Heat treatment of the material can shift the transition end point from 1050 ppm wt H for the stress-relieved state to 100-150 ppm wt H for the {beta} treated state, thus strongly suggesting that there may be some relationship between the microstructure (grain size and shape) and the ductile-brittle transition. It has also been reported that for the same hydriding condition, the hydrogen absorption rate is higher for the stress-relieved and recrystallized states and lower for the {beta} treated state. This phenomenon is very important for engineering applications because it is related tomore » the determination of the safe life. Insufficient attention has been drawn to the quantitative evaluation and the modelization of the influence of the microstructure on the ductile-brittle transition in hydrided Zircaloy-4, though there has been some general research on the boundary structural effect on intergranular fracture. The present authors attempt to modelize this influence by an upper-limit model using the results of image analysis on the microstructures and tensile tests on hydrided sheet specimens.« less

Processing zirconium alloy used in light water reactors for specified creep rate

Abstract: This invention is for the processing of a somewhat broader range of compositions, including ZIRLO material. It controls creep rate in an alloy having, by weight percent, 0.5-2.0 niobium, 0.7-1.5 tin, 0.07-0.28 of at least one of iron, nickel and chromium and up to 220 ppm carbon, and the balance essentially zirconium. The method is of a type which utilizes subjecting the material to a post extrusion anneal, a series of intermediate area reductions and intermediate recrystallization anneals, with one of the intermediate recrystallization anneals possibly being a late stage beta-quench, a final pass area reduction, and a final stress relief anneal. The creep rate is controlled to about the desired amount per hour by the use of an average intermediate recrystallization annealing temperature and a final pass area reduction combination selected from the designated area of a designated figure, with the figure being selected based on whether the post extrusion anneal was an alpha or a beta anneal; on whether the final anneal was a stress relief anneal or a recrystallization anneal, and whether or not a late stage beta-quench was utilized, and the desired creep rate range. The method may also comprise subjecting the material to an alpha post extrusion anneal and a final stress relief anneal, and controlling the creep rate by the use of certain intermediate recrystallization annealing temperatures and final pass true area reductions.

Wear resistant nuclear fuel assembly components

Abstract: Methods and apparatus for improving fretting resistance of zirconium alloy components formed into a shape for use in a nuclear reactor are disclosed in which at least a portion of the outer surface of a component is reacted with material selected from the group consisting of carbon, nitrogen, oxygen and combinations of the foregoing at a temperature below about 700 °C to form a wear resistant layer on the surface of the component.

Lattice misfit variation of Al3(Ti,V,Zr) in Al Ti V Zr alloys

Abstract: It seems likely that high-strength Al-base alloys useful up to 698K can be developed in view of the behavior of nickel base superalloys which resist degradation of mechanical properties to 75 pct of their absolute melting temperature. For high temperature Al alloys, the dispersed hardening phase must not undergo phase transformation to an undesirable phase during long time exposure at the temperature of interest. An additional factor to be considered is the stability of the hardening phase with respect to Ostwald ripening. This coarsening resistance is necessary so that the required strength level can be maintained after the long-time service at high temperatures. A good lattice matching between the intermetallic compound and the aluminum matrix is necessary to promote a low interfacial energy and thus, a low driving force for particle coarsening. In this paper, several Al-base systems containing transition metal elements for potential elevated temperature use were suggested based on low values of diffusivities and solubilities. The equilibrium crystal structures of Al{sub 3}Ti, Al{sub 3}V and Al{sub 3}Zr are tetragonal DO{sub 22}, DO{sub 22} and DO{sub 23}, respectively. At the temperatures of interest, around 698K, vanadium and titanium are mutually substitutable in the form of Al{sub 3}(Ti, V). Muchmore » of titanium and vanadium can be substituted for zirconium in the DO{sub 23}-type Al{sub 3}Zr compound, creating Al{sub 3}(Ti, Zr) and Al{sub 3}(V, Zr), respectively.« less

Zirconium based alloy plate of low irradiation growth, method of manufacturing the same, and use of the same

Abstract: A zirconium based alloy plate of low irradiation growth, containing not more than 5 wt % Sn and/or not more 5 wt % Nb, and the balance Zr of not less than 90 wt %. The alloy plate has a texture that orientation (Fr value) with respect to direction perpendicular to the surface of the plate ranges from 0.20 to 0.50. This alloy plate is effective in being used to form a fuel channel box. Also a fuel assembly using this channel box is provided in which the crystal orientation of the zirconium alloy is made random by a heat treatment. Specifically, by the heat treatment, the Fr, Ft, and Fl values thereof are set to 0.25 to 0.50, 0.25 to 0.36, and 0.25 to 0.36, respectively.

Surface and Corrosion Study of Laser-Processed Zirconium Alloys

Abstract: Reactor Grade Zirconium (Zircaloy-2) was laser-glazed and laser-alloyed with nickel (Ni) or chromium (Cr) powders. Laser alloying produced a surface that was macroscopically, chemically ho...

Zirconium-bismuth-niobium alloy for nuclear fuel cladding barrier

Abstract: A corrosion resistant zirconium alloy is comprised of, in weight percent, about 0.1 to less than 0.5 percent bismuth, about 0.1 to less than 0.5 percent niobium, and the balance substantially zirconium. Preferably, niobium is about 0.1 to 3 weight percent. The alloys have improved corrosion resistance as compared to the moderate-purity sponge zirconium while maintaining a ductility comparable to sponge zirconium.

Effect of microstructure factors and cold work on the hydride precipitation in zircaloy-4 sheet.

Abstract: The effects of microstructure factors and manufacturing process on the precipitation of hydrides in Zircaloy-4 plate have been studied. A quantitative technique described in Ref. 1 has been used to evaluate the susceptibility to radial hydride formation under the influence of texture, residual stress and externally applied tensile stress. Image treatment was carried out to determine the average orientation of hydrides. The hydride oriented threshold stresses were estimated, following an improved approach which had been firstly developed by Sauthoff and then applied to Zr-H system by Puls.2 Both analytical and experimental results indicate that the threshold stress is influenced by the microstructure and thermomechanical treatment of Zircaloy-4, and by the hydrogen concentration in a very complex manner. After examining the hydride morphologies of formerly cold-rolled sheets of two different fabrication routes in stress-relieved, recrystallized and β treated states, it may be concluded that the distribution morphology of hydrides can be selected, and their noxiousness and effective contribution to the embrittlement can be thus minimized by controlling texture, grain structure, and manufacturing process.

Method for annealing zircaloy to improve nodular corrosion resistance

Abstract: A method for annealing a Zircaloy member having a cold worked or beta quenched crystal structure to mitigate the reduction in nodular corrosion resistance caused by the anneal comprises, annealing the member in an atmosphere comprising oxygen and the balance an inert atmosphere to form an adherent black oxide on the member.

Migration of hydrogen through thin films of ZrO2 on Zr-Nb alloy

Abstract: The barrier properties of ZrO2 to inward migration of deuterium have been investigated with a view to understanding the hydriding mechanisms of a Zr‐2.5% Nb alloy used in CANDU nuclear reactors fuel channels. Thin film oxide specimens, grown in steam to ∼1 μm thickness, have been heated to 350 °C and exposed to deuterium gas at pressures ranging from 6×10−3 Pa to 101 kPa (1 atm) and times from 10 to 810 min. Some irreversible uptake can be measured for all exposures using secondary ion mass spectrometry. At low exposures, the shape of the deuterium concentration profile is can be fitted to a Fickian relationship. During longer exposures, the rate of deuterium ingress is sharply curtailed, presumably due to passivated outer oxide surface. Reactions between D2O vapor and the thin film oxide in the 10−3 Pa pressure region and above show a sharply higher uptake of deuterium than in the equivalent pressure of D2 gas. This is ascribed to a more efficient decomposition of D2O on the ZrO2 surface compared to D2.