Showing papers on "Zirconium alloy published in 1976"

Water radiolysis and its effect upon in-reactor zircaloy corrosion

Abstract: The well-known effects of dissolved oxygen and hydrogen in respectively increasing and reducing the corrosion rate of zircaloys irradiated in high temperature water have been attributed to changes in the course of water radiolysis.1 Computer simulation of the growth and decay of radioly tic species produced by β, γ and neutron irradiation of water at 25°C and 305°C has been performed. In general, neutron irradiation leads to greater decomposition of water, and the effect of dissolved oxygen at 305°C is to markedly increase the concentrations of oxidizing radiolytic species. Correlation of zircaloy corrosion behaviour under such conditions with predictions of water radiolysis suggests that either HO2 of O− 2 are significant contributors to the corrosion enhancement phenomenon. Possible enhancement mechanisms are briefly discussed.

Process for electrolytic deposition of metals on zirconium materials

Abstract: A process for the electrolytic deposition of a metal layer on an article comprised of zirconium or a zirconium alloy is disclosed. The article is activated in an aged aqueous solution comprising from about 10 to about 20 grams per liter ammonium bifluoride and from about 0.75 to about 2 grams per liter of sulfuric acid. The solution is aged by immersion of pickled zirconium in the solution for at least about 10 minutes. The loosely adhering film formed on the article in the activating step is removed and the article is contacted with an electrolytic plating solution containing the metal to be deposited on the article in the presence of an electrode receiving current.

Preparation of zirconium alloys

Abstract: High tensile strength, creep and corrosion resistant zirconium alloys containing 7.0-10.0 wt% aluminum, 0-3 wt% in total of one or more elements selected from the group consisting of magnesium, tin, chromium, iron, carbon, silicon, yttrium, niobium, molybdenum and beryllium, the balance zirconium and incidental impurities, are produced by annealing cast alloys at a temperature below about 992° C for a period of time sufficient to produce alloys having a substantially continuous matrix of the ordered intermetallic compound Zr3 Al. The alloys may initially be hot worked at temperatures above about 1000° C, while in the βZr-Zr2 Al two phase region, prior to annealing. Preferred alloys contain 7.5-9.5% aluminum, the balance zirconium and incidental impurities.

Process for improving the heat resistance of zirconium and its alloys

Abstract: A method for treating zirconium and zirconium alloys, in particular the zirconium alloys used in nuclear reactors which are water cooled, as a structural or casing material for fuels. The method consists in dissolving or maintaining a solid solution of the majority of carbon contained in these alloys by thermal or thermo-mechanical treatments carried out in the α + β range or if necessary in the β range followed by a rolling in α phase if necessary. The products obtained have a highly improved mechanical resistance under heat, in particular with regard to resistance to creep; they are particularly suitable for constructing casing tubes for fuels in water cooled nuclear reactors having a better resistance to distortion in relation to time.

Copper base alloys containing chromium, niobium and zirconium

Abstract: Copper base alloys containing chromium, zirconium and niobium are disclosed as well as a process of heat treating and mechanically working said alloys. The combination of alloying elements, hot and cold working, annealing and aging steps increases both the strength and electrical conductivity properties of the alloy without excessive cold working.

Internal Hydrogen Embrittlement of Gamma-Stabilized Uranium Alloys

Abstract: Relationships between the tensile ductility and fracture characteristics of as-quenched, gamma-stabilized uranium alloys (uranium-10 Wt% molybdenum, uranium-8.5 Wt% niobium, uranium-10 Wt%...

Zirconium metal-water oxidation kinetics. I. Thermometry

Abstract: A description is given of the thermometry techniques used in the Zirconium Metal--Water Oxidation Kinetics Program. Temperature measurements in the range 900 to 1500$sup 0$C are made in three experimental systems: two oxidation apparatuses and the annealing furnace used in a corollary study of the diffusion of oxygen in $beta$-Zircaloy. Carefully calibrated Pt vs Pt--10 percent Rh thermocouples are employed in all three apparatuses, while a Pt--6 percent Rh vs Pt-- 30 percent Rh thermocouple and an optical pyrometer are used in addition in the annealing furnace. Features of the experimental systems pertaining to thermocouple installation, temperature control, emf measurements, etc. are described, and potential temperature-measurement error sources are discussed in detail. The accuracy of the temperature measurements is analyzed.

Co-deposition of zirconium with carbon by the bromide process

Abstract: Zirconium was co-deposited with carbon, at temperatures below 1400 °C, onto plates of graphite, alumina, and silicon carbide, respectively, in a flow process, in order to obtain deposits with compositions between zirconium carbide and carbon. Zirconium tetrabromide and methane served as sources of zirconium and carbon, respectively. The effect of the gas-phase composition on the deposition behavior was investigated. The deposits could be divided into four regions, plain, mosaic, fiber and granular, on the basis of the appearance of the surface. The regions showed characteristic patterns in the X-ray diffraction examination. In most cases, the densities of the deposits were considerably lower than the theoretical values.

Effect of Pre-Hydriding on Post-Irradiation Ring-Tensile Properties of Zircaloy-2 Cladding Tubes

Abstract: Zircaloy-2 tubes were hydrided up to a nominal content of 200 ppm and irradiated as fuel claddings in HBWR. Post-irradiation ring-tensile testing revealed that hydrogen enhances the irradiation-induced decrease of elongation and wall thickness reduction at room temperature. On the other hand, no effect of hydrogen was observed on ultimate tensile strength. With testings at 300°C, the effect was negligible on elongation too. From the evaluation of the test results including metallographic observation of ring specimens after fracture, it was concluded that segregation of hydrides due to thermal diffusion of hydrogen during irradiation was at least a part responsible to the above effect of hydrogen enhancing embrittlement.

Nuclear fuel element

Abstract: A nuclear fuel element wherein a tubular cladding of zirconium or a zirconium alloy has a fission gas plenum chamber which is held against collapse by the loops of a spacer in the form of a tube which has been deformed inwardly at three equally spaced, circumferential positions to provide three loops A heat resistant disc of, say, graphite separates nuclear fuel pellets within the cladding from the plenum chamber The spacer is of zirconium or a zirconium alloy

Behavior of Zircaloy Fuel Cladding Tubes

Abstract: A series of experiments has been performed to obtain information on the behavior of zircaloy fuel cladding tube, indispensable for predicting fuel performance in water reactors. The results obtaine...

Oxidation and Corrosion of Tin‐Coated Zircaloy‐4

Abstract: Because of the lack of the adhesion of the growing oxide layer on zirconium alloys, tin-coated Zircaloy-4 has been oxidized in 250 torr oxygen at 500/sup 0/C and corroded in high pressure water at 349/sup 0/C and 176 atm. The rate of both reactions was decreased with increasing thickness of the tin layer (up to 0.1 ..mu..m). Simultaneously, the transition point was shifted to longer periods. Presumably this enhancement was caused by the tin layer which exists between the alloy and the oxide operating as liquid agent. (auth)

Thermo mechanical treatment of zirconium ALLOYS - for nuclear reactors by annealing and subjecting to tensile creep deformation

Abstract: An alloy comprising >=95% Zr and at least one of Cr, Sn Nb, Fe, Ta, Ni and Al which is thoroughly annealed is treated by subjecting the alloy to a tensile creep strain of 3-10% while holding in a temp. range at which creep will occur yet which is below the temp. for significant recovery by loading at a strain rate of 0.001-0.01 inch/inch/minute until the tensile creep strain is obtd. and holding at the final load for at least 15 mins. The treated alloy has high strength, ductility and corrosion resistance to steam. Tubes of the alloy are heated on the upper part of the alpha-range for Zr to anneal and then treated as described. Treated tubes of the alloy are used in nuclear reactors.

Zirconium hydride as a storage medium for tritium

Abstract: Conditions necessary for rapid production (approximately 20 minutes preparation time) have been determined for a 6-g sample of zirconium hydride with a composition of ZrH/sub 1.5/ to ZrH/sub 1.9/. Two alternate sets of conditions were found to produce a hydride of suitable physical integrity for tritium storage: the first condition involves isothermal absorption of H/sub 2/ at 760 torr and 600/sup 0/C; the second involves addition of H/sub 2/ at 760 torr as the temperature is increased from 600/sup 0/C until absorption ceases. The latter method appears to produce a hydride which is essentially crack-free. Small amounts of air in the hydrogen were found to have a very deleterious effect on the hydriding reaction. The cost of zirconium is a disadvantage. However, the use of scrap metal may make the method more attractive and the possible use of irradiated Zircaloy cladding hulls would be even more economically favorable. Questions to be answered mainly concern the actinide and other residual activity remaining in the hulls after decladding. Zircaloy-2 tubing, cladding scrap, was studied and found to be very easily hydrided. Hydrides were produced from 0.25- and 0.50-in. diameter Zircaloy rods, 0.50-in. diameter Zircaloy tubing, and a 0.25-in. diameter zirconium rod.

Oxidation resistance of Ni+10% Cr alloy with additions of silicon, zirconium, hafnium, or rare earth metals

Abstract: 1. Alloying of Ni-10% Cr with silicon along with small amounts of zirconium, hafnium, or REM leads to a sharp increase in the oxidation resistance of the alloy during isothermal and thermal cycling tests. 2. Small additions of hafnium, zirconium, lanthanum, or samarium to Ni-10% Cr-1.8% Si increase the time to failure several times in thermal cycling tests. The effect of these additions on the oxidation resistance of the alloy is negligible during isothermal oxidation, but increases with temperature in the range of 1000–1200°. 3. The increase in the oxidation resistance of Ni-10% Cr when alloyed with silicon along with zirconium, hafnium, or REM is due to formation of a protective layer of SiO2 in the scale and to good adherence of the scale to the alloy, ensured by precipitation of oxides of the second alloying element in the grain boundaries of the alloy beneath the scale.

Electrolytic hydrogen in beta titanium. Final report

Abstract: The vacuum processes of ion bombardment etching followed by thin film vapor deposition of palladium have been employed to produce oxide-free titanium specimens suitable for electrochemical hydrogen permeation and embrittlement studies. The alloy chosen for study was Ti - 11.5 Mo - 6 Zr - 4.5 Sn in the solution treated and quenched condition. Electrochemical permeation experiments which were conducted with basic solutions containing cyanide additions exhibited well behaved hydrogen permeation transientes corresponding to simple diffusion behavior. These experiments yielded a diffusivity at 21 C for hydrogen through the beta alloy of 5.6 (+ or - 1.92) X 10 sq cm/sec. Anomalous permeation behavior occurred when high hydrogen chemical potentials associated with acidic and basic solutions without cyanide, existed at the input side during charging. The anomalous behavior occurs at the later stages of the approach to steady state. This behavior is shown to be consistent with the trapping model of hydrogen in metals of McNabb and Foster. Plastic deformation and spontaneous cracking at the wetted portion of the specimen have been observed under extreme conditions during this anomalous behavior. A substantial part of the deformation is found to be reversible; i.e., interstitial or trapped hydrogen atoms diffuse out ofmore » the metal upon heating or standing. A series of mandrel bend experiments were performed to further define the embrittlement phenomenon. (GRA)« less

An electron microscopy study of the metastable structures in splat-cooled erbium-zirconium alloys

Abstract: The metastable structures produced by splat-cooling in Er-Zr alloys were investigated by using transmission electron microscopy and X ray analysis in order to deduce the mechanism by which a continuous series of supersaturated solid solutions are formed between Er and Zr in their low temperature h c p allotropic form. The study revealed that the microstructure of alloys with extended solid solubility consisted of an h c p phase exhibiting predominantly martensitic character together with small amounts of a metastable beta phase. The volume fraction of the martensitic phase increased and that of the beta phase decreased with increase in Zr content of the alloy. The phase transformations involved in the formation of the various metastable phases in the Er-Zr system during splat cooling and the stability of these phases during subsequent ageing treatments are discussed in relation to the equilibrium phase diagram for this system.

Oxidation and corrosion of tin-coated zircaloy-4

Abstract: Because of the lack of the adhesion of the growing oxide layer on zirconium alloys, tin-coated Zircaloy-4 has been oxidized in 250 torr oxygen at 500/sup 0/C and corroded in high pressure water at 349/sup 0/C and 176 atm. The rate of both reactions was decreased with increasing thickness of the tin layer (up to 0.1 ..mu..m). Simultaneously, the transition point was shifted to longer periods. Presumably this enhancement was caused by the tin layer which exists between the alloy and the oxide operating as liquid agent. (auth)