Showing papers on "Zirconium alloy published in 1987"

Enhanced charge retention electrochemical hydrogen storage alloys and an enhanced charge retention electrochemical cell

Abstract: OF THE DISCLOSURE A high charge retention, reversible, multicomponent, multiphase, electrochemical hydrogen storage alloy comprising titanium, vanadium, zirconium, nickel, and chromium. The hydrogen storage alloy is capable of electrochemically charging and discharging hydrogen in alkaline aqueous media. In one preferred exemplification the hydrogen storage alloy comprises (Ti2-xZrxV4-yNiy)1-zCrz where 0.0 is less than x is less than 1.5, 0.6 is less than y is less than 3.5, and z is an effective amount less than 0.20. Also disclosed is a rechargeable, electrochemical cell utilizing a negative electrode formed of the alloy.

Reduction of Plastic Anisotropy of Zircaloy Cladding by Neutron Irradiation, (II)

Abstract: Initial plastic deformation behavior of zirconium alloy fuel cladding was described quantitatively by the deformation system of single crystal of α-zirconium, and a model was proposed to simulate the yield behavior of polycrystalline material. Based on the model, effects of crystallographic texture and stress state on the plastic deformation of the cladding were evaluated. Conclusions obtained from this investigation are: (1) The proposed model shows good agreement with the von Mises' yield criteria for a material with isotropic properties. (2) Plastic anisotropy of the cladding decreases when neutron irradiation affects prism slip more strongly than the other deformation systems. (3) Dominant deformation systems for axial tension or internal pressurization of the cladding are predicted to be prism slip or tensile twin, respectively, when the stress state of the cladding reaches the yield condition.

Zirconium alloy fuel cladding resistant to PCI crack propagation

Abstract: This invention relates to a tubular water reactor fuel cladding having an outer cylindrical layer composed of a conventional zirconium base alloy. Bonded to the outer layer is a second, inner layer composed of an alloy consisting essentially of: about 0.1 to 0.3 wt. % tin; about 0.05 to 0.2 wt. % iron; about 0.05 to 0.4 wt. % niobium; about 0.03 to 0.1 wt. % of either chromium or nickel, alone or in combination with each other; while keeping the sum of the iron chromium and nickel contents below 0.25 wt. %; 300 to 1200 ppm oxygen; and the balance essentially zirconium. The inner layer is characterized by excellent resistance to PCI crack propagation, excellent aqueous corrosion resistance and a fully recrystallized microstructure.

Corrosion-resistant zirconium alloy

Abstract: PURPOSE:To obtain the titled Zr alloy having excellent corrosion resistance, particularly nodular corrosion resistant-characteristics by specifying the compsn. consisting of Nb, Si, Ge and Zr. CONSTITUTION:The titled Zr alloy consists of, by weight, 0.1-2.5% Nb, at least one kind between 0.01-3% Si and Ge, furthermore consists of <=5% Ni, <=5% Fe, <=1.5% Sn, <=1% Cr and the balance consisting substantially of Zr. Said alloy has excellent corrosion resistance, particularly nodular corrosion resistant-characteristics. The above-mentioned alloy has improved corrosion resistance by finely and uniformly depositing and dispersing the intermetallic compounds between beta stabilizing elements such as Nb and Zr.

Thermomechanical processing of iron, titanium, and zirconium alloys in the BCC structure

Abstract: The body centered cubic (bcc) metals undergo a high level of dynamic recovery during elevated temperature straining so that the stress increases monotonically to a steady-state value σs. The strain rate and σs are related by means of the power, the exponential, or the sinh law with an Arrhenius temperature relationship. The activation energy for a iron has values of 250–280 kJ/mol, whereas for β titanium and β zirconium it is in the range 134–184 kJ/mol. The structure developed during hot working consists of elongated grains containing subgrains of dimension inversely proportional to σs. In warm working of α iron (limited to below 0.66T m), the textures are similar to those for cold working. In working β titanium and β zirconium which is limited to above 0.6T m except in β stabilized alloys or as matrix in α+ β processing, the bcc textures transform into α textures. The α iron relies principally on substructure strengthening in association with carbides. The β phases can be thermomechanically processed to provide equiaxed or lamellar a in a variety of dimensions and combinations, with or without substructure. Hot working of the bcc refractory metal alloys, principally molybdenum, is similar to hot working of α iron.

Pellet-clad interaction resistant nuclear fuel element

Abstract: Nuclear fuel elements are provided that are resistant to pellet-clad interaction. The closed end nuclear fuel elements comprise a zirconium or zirconium alloy tubular cladding (3) that has a layer of lubricant (7), preferably graphite, on the inner surface (5) thereof, and enriched uranium dioxide pellets (9) that have a coating (15) on the outer surface (13) thereof of a thick­ness sufficient to absorb fission products. The coating (15) on the pellets (9) may be a burnable absorber or a material that has a relatively low neutron absorption compared to a burnable absorber. The combination of the layer (7) on the tubular cladding (3) and the coating (15) on the pellets (9) reduces both the stress level and the concentration of damaging fission products that would contact and react with the layer of lubricant (7) about the inside surface (5) of the tubular cladding (3) and thus reduces conditions for pellet-clad interaction.

X-ray photoelectron spectroscopy and Auger studies of selected Zr/Ni intermetallic alloy catalysts

Abstract: Various intermetallic alloys of Zr/Ni have been studied with regard to their surface composition using Auger electron spectroscopy (AES) and x‐ray photoelectron spectroscopy (XPS) as functions of elevated temperature exposure to hydrogen and oxygen gas. The intermetallic alloys (ZrNi5, Zr2Ni7, ZrNi3, ZrNi, and Zr2Ni) were prepared by arc melting techniques, and were studied either as powders or as slices of the arc melted sample button. The elevated temperature treatments were chosen to duplicate the sample treatment these alloys received prior to a recently reported catalytic hydrogenation study. The alloys were sequentially analyzed with XPS and/or Auger techniques after the following representative treatment cycle: as‐prepared; 400 °C H2; 400 °C O2; and a final 400 °C H2 exposure. In general, the as‐prepared alloys are depleted in nickel at the surface (as compared to the bulk), and the alloys have been decomposed into a mixed Ni (or NiO) and ZrO2 surface layer residing on the bulk alloy. After the ele...

Preparation, mechanical strengths, and thermal

Abstract: Ni-based amorphous wires with good bending ductility have been prepared for Ni75Si8B17 and Ni78P12B10 alloys containing 1 to 2 at pct Al or Zr by melt spinning in rotating water The enhancement of the wire-formation tendency by the addition of Al has been clarified to be due to the increase in the stability of the melt jet through the formation of a thin A12O3 film on the outer surface The maximum wire diameter is about 190 to 200 μm for the Ni-Si (or P)-B-Al alloys and increases to about 250 μm for the Ni-Si-B-Al-Cr alloys containing 4 to 6 at pct Cr The tensile fracture strength and fracture elongation are 2730 MPa and 29 pct for (Ni075Si008B01799Al1) wire and 2170 MPa and 24 pct for (Ni078P012B01)99Al1 wire These wires exhibit a fatigue limit under dynamic bending strain in air with a relative humidity of 65 pct; this limit is 050 pct for a Ni-Si-B-Al wire, which is higher by 015 pct than that of a Fe75Si10B15 amorphous wire Furthermore, the Ni-base wires do not fracture during a 180-deg bending even for a sample annealed at temperatures just below the crystallization temperature, in sharp contrast to high embrittlement tendency for Fe-base amorphous alloys Thus, the Ni-based amorphous wires have been shown to be an attractive material similar to Fe- and Co-based amorphous wires because of its high static and dynamic strength, high ductility, high stability to thermal embrittlement, and good corrosion resistance

Improving corrosion resistance of zirconium alloys

Abstract: A process for improving the corrosion resistance and hydrogen absorption resistance of zirconium alloy to be used, for example, in a light water nuclear reactor environment, comprises sputtering a layer of hafnium ions onto the zirconium alloy and implanting the hafnium ions with xenon ion doses of 3 x 1016 ions/ cm2.

Zirconium cladded pressurized water reactor nuclear fuel element

Abstract: A pressurized water reactor nuclear fuel element has a tubular cladding container formed from zirconium or a zirconium alloy material without a protective coating or liner therefor, the cladding material containing less than 4 percent of alloying elements, including an oxygen content of less than 600 parts per million. The cladding contains a sealed nuclear fuel and a pressurized helium atmosphere which fills the gap between the fuel material and the inner wall of the cladding, the helium pressurized to between 150 to 500 pounds per square inch.

Steady-state creep of the N-1 alloy in the 650–870°K range

Abstract: At the present time, the N-1 alloy (which is used as a jacket material in the power reactors working with an aqueous coolant) is studied intensively for establishing the mechanisms controlling the plastic deformation process during creep and for giving a mathematical description of this process. This paper deals with a study of the creep process of the N-1 alloy at higher temperatures that are characteristic of emergency situations. Statistical processing of the results shows that in the 10-150 MPa stress range, the activation energy for creep does not depend on the magnitude of stress and is equal to (240 +/- 10) kJ/mole. This value is close to the activation energy for self-diffusion of ..cap alpha..-zirconium. The actual activation energy for creep (that was obtained taking the temperature dependence of the modulus of elasticity into account) is equal to (235 +/- 10) kJ/mole and it virtually coincides with the apparent activation energy (considering the experimental scatter).

Preparation of reactive metal hydrides

Abstract: Transition metal, notably zirconium alloy, is melted (e.g., by a plasma) to form droplets within the size range of minus 20, plus 60 mesh. The droplets are exposed to a hydrogen atmosphere for a short period while cooling through the hydriding temperature range (e.g., 600° C. to 400° C.). A friable particulate of uniform size and hydrogen content, suitable for sintering or forming components, is recovered.