Showing papers on "Zirconium alloy published in 1988"

Effect of 0.1 at. pct Zirconium on the cyclic oxidation resistance of beta-NiAl

Abstract: The effect of 0.1 at. pct Zr on the cyclic oxidation of hipped beta-NiAl was studied. Oxidation testing was performed in static air at 1100-1200 C, using 1-hr exposure cycles for test times up to 3000 hr. The weight change versus time data were modeled with the COSP computer program to analyze and predict cyclic-oxidation behavior. Zr additions significantly change the nature of the scale-spalling process during cooling, so that the oxide spalls near the oxide-air interface at a relatively low depth within the scale. Without Zr, the predominantly alpha-Al2O3 scale tends to spall randomly to bare metal at relatively high effective-scale-loss rates, particularly at 1150 C and 1200 C. This leads to higher rates of Al consumption for the Zr-free aluminide and much earlier depletion of Al, leading to eventual breakaway (i.e., failure).

XPS studies of the thermal behaviour of passivated Zircaloy‐4 surfaces

Abstract: The composition and the chemical states of components of Zircaloy-4 (zirconium alloy) surfaces were studied in the temperature range between room temperature and 500°C. Each sample was kept at constant temperature (25, 100, 200, 300, 400, 500°C) for up to 16 hours. The changes of composition and chemical states of the Zircaloy-4 surface during heating were monitored by x-ray photoelectron spectroscopy (XPS). Originally, the components form well-defined layers elucidated by angle-resolved x-ray photoelectron spectroscopy (ARXPS). In contrast to depth profiling using ion sputtering, ARXPS is non-destrutive. However, it is applicable for layers of up to a few nanometres thickness only. The experiments showed a decomposition of the ZrO2 coverage above 200°C accompanied by oxygen diffusion into the bulk. These processes lead to the reduction of ZrO2 to metallic zirconium on the surface at 300°C and higher temperatures. The oxygen diffusion into the bulk was indicated by AES depth profiles. The layered structure observed up to a heating temperature of 200°C could not be seen at higher temperatures. After Zr metal appears at the surface during the heating process, a reaction with the adsorbed hydrocarbons takes place, leading to the formation of zirconium carbide. Though the depth resolution of an AES depth profile does not permit identification of layers with thicknesses in the nanometre region, the temperature-dependent behaviour of oxyen is reflected by its AES profiles, showing features in accordance with the results from ARXPS, especially with respect to the fact that well-defined layers vanish above 200°C.

The influence of hydride size and matrix strength on fracture initiation at hydrides in zirconium alloys

Abstract: The effects of matrix strength (yield stress) on hydride fracture and alloy ductility have been studied as a function of stress state, hydride content, hydride size, and precipitation stress. Uniaxial and triaxial states of stress were investigated by using smooth and notched tensile specimens, respectively, containing 0.18 or 0.90 at. pct H, with the longest hydride platelet dimension varying from 5 to 400 μm. The majority of the hydrides in the specimens had their plate normals oriented parallel to the tensile axis direction. Crack initiation at hydrides was monitored using acoustic emission, finiteelement calculations were employed to determine the stresses and strains in the notched specimens, and metallographic and fractographic analyses were carried out to determine the state of fractured hydrides/voids near and on the fracture surface. These techniques showed that, up to a hydride platelet length of ∼50 to 100 μm and regardless of the stress state, a critical plastic strain, independent of matrix strength, controls the initiation of fracture in hydrides. The amount of plastic strain needed to fracture hydrides decreases as (a) the average hydride length increases and (b) the axiality of stress increases. The equivalent plastic strain to fracture small hydrides is ∼ 1 pct under a triaxial as opposed to ∼5 pct under a uniaxial state of stress. When the average hydride platelet lengths are longer than ∼50 to 100 μm, negligible plastic deformation is required to fracture hydrides. A critical applied stress then is the governing factor in all three materials, ranging from 750 to 850 MPa, depending on the stress state.

Fuel rod for a nuclear reactor fuel assembly

Abstract: A fuel rod for a nuclear reactor fuel assembly includes a cladding tube having an outer surface and a given total wall thickness. Nuclear fuel is disposed in the cladding tube. The cladding tube is formed of a first zirconium alloy which may have alloy components of from 1.2 to 2% by weight of tin, 0.07 to 0.2% by weight of iron, 0.05 to 0.15% by weight of chromium, 0.03 to 0.08% by weight of nickel, 0.07 to 0.15% by weight of oxygen, and a total percent by weight for the alloy components of iron, chromium and nickel in a range of from 0.18 to 0.38% by weight. The first zirconium alloy may also have alloy components of from 1.2 to 2% by weight of tin, 0.18 to 0.24% by weight of iron, 0.07 to 0.13% by weight of chromium, 0.10 to 0.16% by weight of oxygen, and a total percentage by weight for the components of iron and chromium in a range of from 0.28 to 0.37% by weight. A surface layer which is disposed on the outer surface of the cladding tube is formed of a second zirconium alloy having a layer thickness in a range of from 5 to 20% of the given total wall thickness of the cladding tube, the second zirconium alloy being formed of at least one alloy component from the group consisting of iron, chromium, nickel and tin having a total percentage by weight of the alloy components of the group of from 0.4 to 1% by weight and/or having from 0.2 to 3% by weight of niobium as an alloy component.

Nuclear fuel element having oxidation resistant cladding

Abstract: Nuclear fuel elements for use in the core of a nuclear reactor include an improved composite cladding (17) having a zirconium barrier layer (22) metallurgically bonded on the inside surface of a zirconium alloy tube (21), wherein the inside surface of the barrier is alloyed with preselected elemental impurities to improve oxidation resistance. The zirconium barrier layer (22) forms a shield between the zirconium alloy tube (21) and a core of nuclear fuel material (16) enclosed in the composite cladding. The alloy layer formed on the barrier surface acts to inhibit cracking during the tube production fabrication step and limits oxidation in the event that the cladding is breached during operation of the reactor, allowing the entrance of water or steam into the fuel element.

Hydrogen storage characteristics of mechanically alloyed amorphous metals

Abstract: The hydrogen storage properties of a series of mechanically alloyed (MA) amorphous Ni1xZrx alloys are studied, using both gas phase and electrochemical techniques, and are compared to H storage of rapidly quenched (RQ) amorphous Ni1−xZrx. In the MA alloys, hydrogen resides in the Ni4−nZrn (n = 4,3,2) tetrahedral interstitial sites, with a maximum hydrogen-to-metal ratio of 1.9(4 n)xn(1 − x)4 − n. These features are identical to those of the RQ alloys and indicate that the topological and chemical order of the MA and RQ materials are essentially the same. However, the typical binding energy of hydrogen in a Ni4−nZrn site, En, is shifted in the MA alloys relative to the RQ alloys and the distribution of binding energies centered on En is significantly broader in the MA samples. Thus, the MA and RQ alloys are not identical and sample annealing does not alter this subtle distinction. The sensitivity of H storage to the presence of chemical order in binary alloys are analyzed theoretically and the data are found to be most consistent with little or no chemical order (random alloys).

Ductile irradiated zirconium alloy

Abstract: A stabilized alpha metal matrix provides an improved ductility after irradation without loss of corrosion resistance in a "Zircaloy" alloy modified with measurable amounts of up to 0.6 percent by weight of niobium or 0.1 percent by weight of molybdenum. Tin is present in the Zircaloy in the range of 1.2 to 1.70 percent by weight and the oxygen level is in the range of from 1000 to 1600 ppm. Iron and chromium alloying element levels are those of typical Zircaloys. The average intermetallic precipitates' particle sizes are in the range of from 1200 to 1800 angstroms, thereby providing optimum corrosion resistance of the improved alloy in both boiling water and pressurized water reactors.

Nuclear reactor fuel assembly

Abstract: A nuclear reactor fuel assembly includes a fuel rod containing nuclear fuel in a cladding tube formed of an iron-containing zirconium alloy A fuel assembly skeleton to which the fuel rod is attached has a structural part formed of the iron-containing zirconium alloy The iron-containing zirconium alloy has an oxygen content of from 01 to 016% by weight and contains alloy components of from 0 to 1% by weight of niobium, 0 to 08% by weight of tin, at least two metals from the group consisting of iron, chromium and vanadium having from 02 to 08% by weight of iron, 0 to 04% by weight of chromium and 0 to 03% by weight of vanadium, a total percent by weight of iron, chromium and vanadium of from 025 to 1% by weight, and a total percent by weight for niobium and tin in the range from 0 to 1% by weight

Electrochemical-metallothermic reduction of zirconium in molten salt solutions

Abstract: This is a method of reducing zirconium chloride to a metal product by introducing zirconium chloride into a molten salt bath containing at least one alkali metal chloride and at least one alkaline earth metal chloride; and electrochemically reducing alkaline earth metal chloride to a metallic alkaline earth metal in the molten salt bath, with the reduced alkaline earth metal reacting with the zirconium chloride to produce zirconium metal. By using this electrochemical-metallothermic reduction, zirconium metal is produced and insoluble subchlorides of zirconium in the metal product are generally avoided. Preferably, the molten salt in the molten salt bath consists essentially of a mixture of lithium chloride, potassium chloride, magnesium chloride and zirconium or hafnium chloride. The method is especially useful as part of a distillation system for separating hafnium from zirconium, possibly after the zirconium chloride is removed from the distillation system, but especially where the distillation system has an alkali metal chloride and alkaline earth metal chloride recirculating solvent and the electrochemical-metallothermic reduction is used to strip the solvent of zirconium chloride. This process can also be used for hafnium and titanium, especially when a powder metal product is desired.

Corrosion resistant zirconium alloys containing bismuth

Abstract: Zirconium-based corrosion resistant alloys for use primarily as a cladding material for fuel rods in a boiling water nuclear reactor which consist essentially of 0.5 to 2.5 weight percent bismuth, or alternatively, 0.5 to 2.5 weight percent of a mixture of tin and bismuth, 0.5-1.0 weight percent of a solute composed of a member selected from the group consisting of niobium, molybdenum, tellurium, and mixture thereof, alternatively, the solute will be composed of tellurium and will be in the range of 0.3-1.0 weight percent, and the balance being zirconium.

Water reactor fuel cladding

Abstract: This invention describes a tubular water reactor fuel cladding 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 the group of zirconium base alloys consisting of: about 0.19 to 0.6 wt. % tin, about 0.19 to 0.5 wt. % iron, and about 100 to 700 ppm oxygen; or about 0.4 to 0.6 wt. % tin, about 0.1 to 0.3 wt. % iron, about 0.1 to 0.3 wt. % nickel, and about 100 to 700 ppm oxygen.

Properties of amorphous Zr3RhHx prepared from glassy and crystalline alloys.

Abstract: The electronic, magnetic, and thermal properties have been measured on the amorphous hydride phases prepared from originally glassy or crystalline Zr/sub 3/Rh alloys. The a-Zr/sub 3/RhH/sub x/ samples with xless than or equal to5.5 were studied by x-ray diffraction, proton nuclear magnetic resonance, magnetic susceptibility, low-temperature heat capacity, differential scanning calorimetry, and inelastic neutron scattering. All hydride samples are amorphous with similar properties which are shown to be independent of the structure for the initial alloy. Namely, the solid-state reaction of hydrogen with crystalline c-Zr/sub 3/Rh appeared to produce an equivalent amorphous phase to hydrogenated glassy alloys. The changes in the electronic and magnetic properties upon hydrogenation imply a systematic decrease in the Fermi-level density of states in a-Zr/sub 3/RhH/sub x/ as the hydrogen content x increases. The thermal stabilities of the amorphous hydrides (with respect to the irreversible formation of crystalline ZrH/sub x/ phases) also decrease with increasing hydrogen stoichiometry. Whereas the hydrogen atoms predominantly occupy tetrahedral interstitial sites coordinated with zirconium atoms, there is strong circumstantial evidence for occupancies of different and less stable sites for compositions with higher hydrogen content

Gettering of hydrogen isotopes by Zr-Ni alloys

Abstract: Activation process and absorption/desorption of hydrogen isotopes were studied for Zr-Ni alloys by means of XPS-SIMS and thermal desorption spectroscopy Alloying of Ni to Zr gave rise to considerable modification of the getter properties: it caused the changes in the activation temperature, the activation energy for hydrogen absorption, the heat of absorption, selective pumping property, and the kinetic isotope effect The results indicate that one can develop Zr-Ni getters applicable to the various unit processes in the fuel handling systems of thermonuclear reactors

Tube, bar, sheet or strip made from zirconium alloy resistant both to uniform and nodular corrosion

Abstract: The invention relates to a tube, bar, sheet or strip resistant both to uniform and nodular corrosion, of a zirconium-base alloy of a composition (% by weight) of Fe 0.1 to 0.35, V 0.07 to 0.4, Q 0.05 to 0.3, Sn below 0.25, Nb below 0.25, trace impurities and balance Zr. The invention also relates to a process for producing these products which have a greatly increased service life. It also relates to composite tube having an internal and/or external alloy sheathing.

Alloy coated fuel cladding

Abstract: A fuel element for a nuclear reactor having a zirconium tin alloy cladding tube, with a thin coating of enriched boron-containing alloy burnable poison, such as nickel-thallium-boron deposited from a reducing agent by an electroless plating process.