Showing papers on "Zirconium alloy published in 1985"

The influence of multiaxial states of stress on the hydrogen embrittlement of zirconium alloy sheet

Abstract: The ductility of ZIRCALOY *-2 sheets containing 21-615 wt ppm hydrogen has been investigated at room temperature over a range of stress states from uniaxial to equibiaxial tension. Data based on locally determined fracture strains show a decrease in ductility with both increasing hydrogen content and increasing degree of biaxiality of the stress state. Metallographic and fractographic examinations indicate that the embrittlement is a consequence of void nucleation (due to hydride fracture), void growth, and void link-up. The influence of hydrogen content and stress state on each of the sequential stages of ductile fracture is determined. These results indicate that the primary cause for the influence of stress state on the hydrogen embrittlement of the ZIRCALOY sheet is that void link-up is initiated at a much lower critical void density in equibiaxial tension than in uniaxial tension. This appears to be a result of equibiaxial deformation enhancing (a) direct participation of previously unfractured hydrides in providing a fracture path linking up voids and (b) a localized shear instability process which is triggered by the nucleation of voids.

Process for fabricating a zirconium-niobium alloy and articles resulting therefrom

Abstract: Articles, such as tubing, which have excellent corrosion resistance to steam at elevated temperatures and to hydriding, are produced from zirconium alloys containing 0.5 to 2.0 percent niobium, up to 1.5 percent tin, and up to 0.25 percent of a third alloying element such as iron, chromium, molybdenum, vanadium, copper, nickel and tungsten. The articles are formed by beta-treating the alloy, initially deforming the same at a temperature below 650° C. and further deforming the same through cold working stages also below 650° C., annealing the material between the cold working stages at a temperature between 500°-650° C., and final annealing the same at a temperature below 650° C. to provide articles having a microstructure of fine precipitates of less than about 800 Å, homogeneously dispersed throughout the zirconium.

Amorphous iron zirconium formed by solid state reaction

Abstract: We report on the first iron based amorphous alloy formed by the process of solid state reaction. Amorphous iron zirconium was observed to result from thermal anneals of layered crystalline elemental constituents. Starting material was electron beam evaporated with about 50 monolayers of each constituent per layer, which results in an average composition of Fe61Zr39 and a compositional wavelength of 22 nm. Thermal anneals at 623 K were performed in vacuum, and x‐ray diffraction was used to determine the structure. A calculated metastable free energy diagram shows the reaction is driven by a large thermodynamic driving force (several kcal/mole).

Crystallization characteristics of Co‐Zr metallic glasses from Co52Zr48 to Co20Zr80

Abstract: We present a crystallization study of melt‐spun Co‐Zr metallic glasses. The primary crystallization products are all metastable or unstable except around the equiatomic composition where the stable phase CoZr is produced. Alloys at the composition Co25Zr75 show a crystallization characteristic that is dependent on heating rate.

Co hydrogenation over an amorphous gold-zirconium alloy

Abstract: An amorphous Au25Zr75 alloy showed a high catalytic activity for hydrogenation of carbon monoxide. Thus, the rate of methanation was 2.3 μmol s−1 g−1 under 6.0 MPa at 523 K, being much higher than that over Au/ZrO2 or ZrO2. It was found that the alloy was oxidized into metallic gold and ZrO2 under the reaction conditions.

Process for forming seamless tubing of zirconium or titanium alloys from welded precursors

Abstract: Seamless tubing is formed from welded precursors of zirconium or titanium material by heating successive axial segments of welded tubing completely through the wall of the tubing to convert the material to the beta phase and then rapidly quenching the segments, with the heating and quenching effected so as to prevent the growth of beta grains within the material.

Processes for producing improved wear resistant coatings on zirconium shapes

Abstract: Process for producing zirconium oxide blue-black coatings on zirconium alloy substrates by treating said alloys with molten salts containing small amounts of oxidizing compounds. The molten salts are sodium cyanide and combinations of sodium chloride and potassium chloride.

Wear resistant zirconium base alloy article for water reactors

Abstract: Slender water displacer rods for use in water reactors are provided with rings of a wear resistant coating spaced along the length of the rod. Each coating contains Cr2 C3 and is metallurgically applied and bonded to the zirconium base alloy forming the outer portion of the rod by electrospark-deposition (ESD) technique.

Process for producing high purity zirconium and hafnium

Abstract: This is a process for producing high quality zirconium or hafnium sponge. It is especially useful for producing the very high purity zirconium which may be used to line the inner surface zirconium of Zircaloy tubing for nuclear fuel rods. The process uses a combination reduction-distillation vessel which is directly fed with zirconium or hafnium tetrachloride from a molten salt sublimer. A precharge of magnesium chloride is used to prevent reduction of metal outside the inner liner, and thereby facilitate removal of the inner liner after reduction and distillation. The agitator for the molten salt sublimer preferably utilizes a molten metal seal around the agitator shaft.

Zirconium metal matrix-silicon carbide composite nuclear reactor components

Abstract: A metallic composite material and nuclear components such as fuel cladding, rod guide thimbles, grids and channels made therefrom. The metallic composite material comprises 90-60 volume percent of a metal matrix of zirconium or a zirconium alloy containing homogeneously incorporated, throughout the matrix, 10-40 volume percent of silicon carbide whiskers.

Composite nuclear fuel cladding tubing and other core internal structures with improved corrosion resistance

Abstract: An elongated nuclear fuel cladding tube is provided with at least two annular concentric layers which are bonded to each other The inner layer is composed of a zirconium base alloy The outer layer is composed of a titanium base material exhibiting excellent corrosion resistance in elevated temperature and pressure aqueous environments

A review of zircaloy fuel cladding behavior in a loss-of-coolant accident

Abstract: The paper reviews the state-of-the-art experimental work performed in several countries with respect to the acceptance criteria established for emergency core cooling (ECC) in a loss-of-coolant accident (LOCA) of light water reactors (LWRs). It covers in detail oxidation, embrittlement, plastic deformation and coolability of deformed rod bundles. The main test results are discussed on the basis of research work performed at the Karlsruhe Nuclear Research Center (KfK) within the framework of the Nuclear Safety Project (PNS) and reference is made to test data obtained in other countries. The conclusion reached in the paper is that the major mechanisms and consequences of oxidation, deformation and emergency core cooling are sufficiently investigated in order to provide a reliable data base for safety assessments and licensing of LWRs. All test data prove that the ECC-criteria are conservative and that the coolability of an LWR and the public safety can be maintained in a LOCA. Ein Uberblick Uber das Zircaloy-HUllrohrverhalten beim KUhlmittelverlust stör fall

Hydrogen permeation protection for metals

Abstract: A method is proposed to attenuate hydrogen diffusion through fusion reactor components via controlled oxidation of prealloyed surfaces. Alloys of zirconium, vanadium, titanium, niobium, tantalum, tungsten, nickel and iron may be protected from hydrogen isotope permeation by using vanadium as an intermediate layer and either chromium, nickel-chromium, silicon or aluminum as the oxide forming component. This method may also be used to prevent hydrogen embrittlement of metals from external hydrogen environments and is applicable to the majority of alloy systems exhibiting hydrogen embrittlement phenomena.

Annealing of zirconium based articles by induction heating

Abstract: Processes for the rapid alpha annealing of zirconium based articles are described. These processes utilize induction heating to rapidly heat a worked zirconium based article to an elevated temperature after which it is then cooled. Time at the selected elevated temperature is less than about 1 second, and preferably essentially zero. Stress relieving of cold pilgered Zircaloy may be performed by induction heating to a temperature between about 540° and about 650° C. Partial recrystallization annealing of cold pilgered Zircaloy may be performed by induction heating to a temperature between about 650° and about 760° C. Full alpha recrystallization annealing of cold pilgered Zircaloy may be performed by induction heating to a temperature between about 760° and about 900° C.

Process for coating the internal surface of zirconium tubes with neutron absorbers

Abstract: A process for forming a boron-containing coating on the internal surface of a zirconium or zirconium alloy hollow tube by heating the internal surface to a temperature of between 200°-450° C. and passing through the tube a mixture of a volatilized boron compound in helium or argon, such that the boron compound decomposes to form an integral boron containing coating on the internal surface.

Nuclear fuel cladding containing a burnable absorber

Abstract: A nuclear fuel rod cladding having a burnable absorber integrally incorporated therein has an outer tubular layer of a zirconium alloy; an intermediate layer, of a thickness less than the outer tubular layer, of a zirconium alloy containing a boron-containing burnable absorber; and an inner layer, of a thickness less than the intermediate layer, of zirconium metal. The layers are bonded together to form a cladding for the containment of nuclear fuel.

Cladding tube for nuclear fuel and nuclear fuel element having this cladding tube

Abstract: 57 Disclosed are a cladding tube for a nuclear fuel and a nuclear fuel element incorporating the cladding tube. The cladding tube consists of an inner zirconium liner layer (6;34) and an outer zirconium alloy layer (1). The cladding tube has at least one of the following features. (I) the ratio a/b of the oxygen content a to iron content b in the zirconium liner layer (6; 34) is greater than 1.0, (II) the zirconium liner layer (34) is made of a zirconium into the matrix of which impurities are dissolved, and (III) the second phase particles having microscopic sizes and dispersed in the inner surface of the zirconium liner layer (6) and/or the outer surface of the zirconium alloy layer (1) have been removed substantially. Owing to these features, undesirable stress corrosion cracking and local corrosion are remarkably suppressed in the cladding tube (6, 1; 34, 1) and the nuclear fuel element of the invention.

Water reactor fuel cladding tubes

Abstract: Tubular water reactor fuel cladding tubes having an outer cylindrical layer composed of a conventional zirconium base alloy. Bonded to the outer cylindrical layer is a second layer composed of an alloy selected from zirconium base alloys consisting essentially of (A) 0.19 to 0.6 wt.% tin, and 0.19to 0.5 wt.% iron; or (B) 0.4 to 0.6 wt.% tin, 0.1 to 0.3 wt.% iron, and 0.1 to 0.3 wt.% nickel; plus 100 to 700 ppm oxygen and a balance of zirconium.

Hydrogen absorbing zirconium alloy

Abstract: A zirconium alloy having a plateau region of equilibrium hydrogen dissociation pressure and a small hysteresis, whose hydrogen dissociation pressure is in a range of 06 to 3 atm at 20° C, 1 to 5 atm at 40° C, and 1 to 17 atm at 80° C, the activation of which can be effected according to a single procedure of vacuum evacuation at room temperature and pressurization under a hydrogen pressure of 30 atm at room temperature, and which is comparable or superior in the rates of hydrogen absorption and release, the maximum hydrogen absorbing capacity and effective hydrogen releasing capacity to conventional zirconium alloys A hydrogen absorbing and releasing material comprising an alloy as mentioned above and thin films of one element selected from Pd, Cu, and Ni, which allows only hydrogen to permeate, and covering the surfaces of particles of the alloy

Process for the manufacture of a strip made of zirconium zircaloy 2 or restored zircaloy 4 alloy, and strip obtained

Abstract: Process for the manufacture of a strip made of zircaloy 2 or zircaloy 4, with an oxygen content of between 900 and 1600 ppm and with mechanical characteristics meeting the three conditions: E0.2 at 315 DEG C >/= 250 MPa A % distributed "long" at 20 DEG C >/= 4 and A % distributed "across" at 20 DEG C >/= 4. After the final intermediate annealing the strip of the invention is rolled with a degree of distortion of between 30 and 50 %, and is then heat-treated at a temperature of between 490 and 580 DEG C for 1 to 10 min. The heat treatment is either performed in air and is then followed by a slight polishing and a pickling, or performed under protective atmosphere. The strip of the invention is partially recrystallised over 0.5 to 5 % of its volume. It is employed for the manufacture of spacing grids employed in nuclear fuel elements.

Effects of hydrogen on magnetic properties of amorphous zirconium alloys

Abstract: The magnetic susceptibilities χ(T) of the amorphous ternary hydrides of the paramagnetic metallic glasses a‐Zr2Pd and a‐Zr3Rh have been measured between 7 and 300 K. It has been found that χ(T) of both a‐Zr2PdHx and a‐Zr3RhHx decrease substantially with increasing hydrogen content. This behavior is primarily attributed to reductions in the densities of the Fermi level electronic states. Although χ(T) for crystalline Zr2Pd (MoSi2 structure type) is smaller than the susceptibility of the a‐Zr2Pd glass, χ(T) for a crystallized Zr3Rh sample was found to increase relative to the initial susceptibility for this glass.

High strength aluminum-base alloy containing lithium and zirconium and methods of preparation

Abstract: Novel Al-Li-Zr alloys of the formula Al bal Li b Zr c wherein "b" is about 1.9 through about 4.5 weight percent and "c" is about 0.70 through about 4.0 weight percent are disclosed. These alloys can be formed by mixing Al alloy, Li alloy and Zr alloy, heating to high temperatures and rapidly cooling, followed by heat treatment.