Layer Protecting the Surface of Zirconium Used in Nuclear Reactors.

PubMed

Ashcheulov, Petr; Skoda, Radek; Skarohlíd, Jan; Taylor, Andrew; Fendrych, Frantisek; Kratochvílová, Irena

2016-01-01

Zirconium alloys have very useful properties for nuclear facilities applications having low absorption cross-section of thermal electrons, high ductility, hardness and corrosion resistance. However, there is also a significant disadvantage: it reacts with water steam and during this (oxidative) reaction it releases hydrogen gas, which partly diffuses into the alloy forming zirconium hydrides. A new strategy for surface protection of zirconium alloys against undesirable oxidation in nuclear reactors by polycrystalline diamond film has been patented- Czech patent 305059: Layer protecting the surface of zirconium alloys used in nuclear reactors and PCT patent: Layer for protecting surface of zirconium alloys (Patent Number: WO2015039636-A1). The zirconium alloy surface was covered by polycrystalline diamond layer grown in plasma enhanced chemical vapor deposition apparatus with linear antenna delivery system. Substantial progress in the description and understanding of the polycrystalline diamond/ zirconium alloys interface and material properties under standard and nuclear reactors conditions (irradiation, hot steam oxidation experiments and heating-quenching cycles) was made. In addition, process technology for the deposition of protective polycrystalline diamond films onto the surface of zirconium alloys was optimized. Zircaloy2 nuclear fuel pins were covered by 300 nm thick protective polycrystalline diamond layer (PCD) using plasma enhanced chemical vapor deposition apparatus with linear antenna delivery system. The polycrystalline diamond layer protects the zirconium alloy surface against undesirable oxidation and consolidates its chemical stability while preserving its functionality. PCD covered Zircaloy2 and standard Zircaloy2 pins were for 30 min. oxidized in 1100°C hot steam. Under these conditions α phase of zirconium changes to β phase (more opened for oxygen/hydrogen diffusion). PCD anticorrosion protection of Zircaloy nuclear fuel assemblies can significantly prolong lifetime of Zirconium alloy in nuclear reactors even above Zirconium phase transition temperatures. Even after ion beam irradiation (10 dpa, 3 MeV Fe(2+)) the diamond film still shows satisfactory structural integrity with both sp(3) and sp(2) carbon phases. Zircaloy2 under the carbon-based protective layer after hot steam oxidation test differed from the original Zircaloy2 material composition only very slightly, proving that the diamond coating increases the material resistance to high temperature oxidation. Zirconium alloys nuclear fuel pins' surfaces were covered by compact and homogeneous polycrystalline diamond layers consisting of sp(3) and sp(2) carbon phases with a high crystalline diamond content and low roughness. Diamond withstands very high temperatures, has excellent thermal conductivity and low chemical reactivity, it does not degrade over time and (important for the nuclear fuel cladding) being pure carbon, it has perfect neutron cross-section properties. Moreover, polycrystalline diamond layers consisting of crystalline (sp(3)) and amorphous (sp(2)) carbon phases could have suitable thermal expansion. Zirconium alloys coated with polycrystalline diamond film are protected against undesirable changes and processes. Further, the polycrystalline diamond layer prevents the reaction between the alloy surface and water vapor. During such reaction, water molecules dissociate and initiate formation of zirconium dioxide and hydrogen, accompanied by the release of large amount of heat. Thus the protective layer prevents the formation of hydrogen and the release of reaction heat. Few relevant patents to the topic have been reviewed and cited.

Zirconium-based alloys, nuclear fuel rods and nuclear reactors including such alloys, and related methods

DOEpatents

Mariani, Robert Dominick

2014-09-09

Zirconium-based metal alloy compositions comprise zirconium, a first additive in which the permeability of hydrogen decreases with increasing temperatures at least over a temperature range extending from 350.degree. C. to 750.degree. C., and a second additive having a solubility in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. At least one of a solubility of the first additive in the second additive over the temperature range extending from 350.degree. C. to 750.degree. C. and a solubility of the second additive in the first additive over the temperature range extending from 350.degree. C. to 750.degree. C. is higher than the solubility of the second additive in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. Nuclear fuel rods include a cladding material comprising such metal alloy compositions, and nuclear reactors include such fuel rods. Methods are used to fabricate such zirconium-based metal alloy compositions.

NEUTRON REACTOR FUEL ELEMENT UTILIZING ZIRCONIUM-BASE ALLOYS

DOEpatents

Saller, H.A.; Keeler, J.R.; Szumachowski, E.R.

1957-11-12

This patent relates to clad fuel elements for use in neutronic reactors and is drawn to such a fuel element which consists of a core of fissionable material, comprised of an alloy of zirconium and U/sup 235/ enriched uranium, encased in a jacket of a binary zirconium-tin alloy in which the tin content ranges between 1 and 15% by weight.

A simple spectrophotometric method for determination of zirconium or hafnium in selected molybdenum-base alloys

NASA Technical Reports Server (NTRS)

Dupraw, W. A.

1972-01-01

A simple analytical procedure is described for accurately and precisely determining the zirconium or hafnium content of molybdenum-base alloys. The procedure is based on the reaction of the reagent Arsenazo III with zirconium or hafnium in strong hydrochloric acid solution. The colored complexes of zirconium or hafnium are formed in the presence of molybdenum. Titanium or rhenium in the alloy have no adverse effect on the zirconium or hafnium complex at the following levels in the selected aliquot: Mo, 10 mg; Re, 10 mg; Ti, 1 mg. The spectrophotometric measurement of the zirconium or hafnium complex is accomplished without prior separation with a relative standard deviation of 1.3 to 2.7 percent.

Filler metal alloy for welding cast nickel aluminide alloys

DOEpatents

Santella, Michael L.; Sikka, Vinod K.

1998-01-01

A filler metal alloy used as a filler for welding east nickel aluminide alloys contains from about 15 to about 17 wt. % chromium, from about 4 to about 5 wt. % aluminum, equal to or less than about 1.5 wt. % molybdenum, from about 1 to about 4.5 wt. % zirconium, equal to or less than about 0.01 wt. % yttrium, equal to or less than about 0.01 wt. % boron and the balance nickel. The filler metal alloy is made by melting and casting techniques such as are melting the components of the filler metal alloy and east in copper chill molds.

78 FR 40200 - Duke Energy Carolinas, LLC, Oconee Nuclear Station Units 1, 2, and 3; Independent Spent Fuel...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-07-03

... breaches.'' Zircaloy is a type of zirconium alloy which includes both Zircaloy-2 and Zircaloy-4 cladding, but does not include M5 cladding. The M5 is a different type of zirconium alloy, which does not... ``zirconium alloy'' clad spent fuel assemblies in the 24PHB DSC, which would include both the ``zircaloy clad...

Copper-acrylic enamel serves as lubricant for cold drawing of refractory metals

NASA Technical Reports Server (NTRS)

Beane, C.; Karasek, F.

1966-01-01

Acrylic enamel spray containing metallic copper pigment lubricates refractory metal tubing during cold drawing operations so that the tubing surface remains free from scratches and nicks and does not seize in the die. Zirconium alloys, zirconium, tantalum alloys, niobium alloys, vanandium alloys and titanium alloys have been drawn using this lubricant.

High temperature mechanical properties of a zirconium-modified, precipitation- strengthened nickel, 30 percent copper alloy

NASA Technical Reports Server (NTRS)

Whittenberger, J. D.

1974-01-01

A precipitation-strengthened Monel-type alloy has been developed through minor alloying additions of zirconium to a base Ni-30Cu alloy. The results of this exploratory study indicate that thermomechanical processing of a solution-treated Ni-30Cu-0.2Zr alloy produced a dispersion of precipitates. The precipitates have been tentatively identified as a Ni5Zr compound. A comparison of the mechanical properties, as determined by testing in air, of the zirconium-modified alloy to those of a Ni-30Cu alloy reveals that the precipitation-strengthened alloy has improved tensile properties to 1200 K and improved stress-rupture properties to 1100 K. The oxidation characteristics of the modified alloy appeared to be equivalent to those of the base Ni-30Cu alloy.

Characterization of deformation mechanisms in zirconium alloys: effect of temperature and irradiation

NASA Astrophysics Data System (ADS)

Long, Fei

Zirconium alloys have been widely used in the CANDU (CANada Deuterium Uranium) reactor as core structural materials. Alloy such as Zircaloy-2 has been used for calandria tubes; fuel cladding; the pressure tube is manufactured from alloy Zr-2.5Nb. During in-reactor service, these alloys are exposed to a high flux of fast neutron at elevated temperatures. It is important to understand the effect of temperature and irradiation on the deformation mechanism of zirconium alloys. Aiming to provide experimental guidance for future modeling predictions on the properties of zirconium alloys this thesis describes the result of an investigation of the change of slip and twinning modes in Zircaloy-2 and Zr-2.5Nb as a function of temperature and irradiation. The aim is to provide scientific fundamentals and experimental evidences for future industry modeling in processing technique design, and in-reactor property change prediction of zirconium components. In situ neutron diffraction mechanical tests carried out on alloy Zircaloy-2 at three temperatures: 100¢ªC, 300¢ªC, and 500¢ªC, and described in Chapter 3. The evolution of the lattice strain of individual grain families in the loading and Poisson's directions during deformation, which probes the operation of slip and twinning modes at different stress levels, are described. By using the same type of in situ neutron diffraction technique, tests on Zr-2.5Nb pressure tube material samples, in either the fast-neutron irradiated or un-irradiated condition, are reported in Chapter 4. In Chapter 5, the measurement of dislocation density by means of line profile analysis of neutron diffraction patterns, as well as TEM observations of the dislocation microstructural evolution, is described. In Chapter 6 a hot-rolled Zr-2.5Nb with a larger grain size compared with the pressure tubing was used to study the development of dislocation microstructures with increasing plastic strain. In Chapter 7, in situ loading of heavy ion irradiated hot-rolled Zr-2.5Nb alloy is described, providing evidence for the interaction between moving dislocations and irradiation induced loops. Chapter 8 gives the effect on the dislocation structure of different levels of compressive strains along two directions in the hot-rolled Zr-2.5Nb alloy. By using high resolution neutron diffraction and TEM observations, the evolution of type and dislocation densities, as well as changes of dislocation microstructure with plastic strain were characterized.

Fretting wear behavior of zirconium alloy in B-Li water at 300 °C

NASA Astrophysics Data System (ADS)

Zhang, Lefu; Lai, Ping; Liu, Qingdong; Zeng, Qifeng; Lu, Junqiang; Guo, Xianglong

2018-02-01

The tangential fretting wear of three kinds of zirconium alloys tube mated with 304 stainless steel (SS) plate was investigated. The tests were conducted in an autoclave containing 300 °C pressurized B-Li water for tube-on-plate contact configuration. The worn surfaces were examined with scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS) and 3D microscopy. The cross-section of wear scar was examined with transmission electron microscope (TEM). The results indicated that the dominant wear mechanism of zirconium alloys in this test condition was delamination and oxidation. The oxide layer on the fretted area consists of outer oxide layer composed of iron oxide and zirconium oxide and inner oxide layer composed of zirconium oxide.

Manufacturing process to reduce large grain growth in zirconium alloys

DOEpatents

Rosecrans, Peter M.

1987-01-01

A method of treating cold-worked zirconium alloys to reduce large grain gth during thermal treatment at temperatures above the recrystallization temperature of the alloy comprising heating the cold-worked alloy between about 1300.degree.-1350.degree. F. for 1 to 3 hours prior to treatment above its recrystallization temperature.

Filler metal alloy for welding cast nickel aluminide alloys

DOEpatents

Santella, M.L.; Sikka, V.K.

1998-03-10

A filler metal alloy used as a filler for welding cast nickel aluminide alloys contains from about 15 to about 17 wt. % chromium, from about 4 to about 5 wt. % aluminum, equal to or less than about 1.5 wt. % molybdenum, from about 1 to about 4.5 wt. % zirconium, equal to or less than about 0.01 wt. % yttrium, equal to or less than about 0.01 wt. % boron and the balance nickel. The filler metal alloy is made by melting and casting techniques such as are melting the components of the filler metal alloy and cast in copper chill molds. 3 figs.

METHOD AND ALLOY FOR BONDING TO ZIRCONIUM

DOEpatents

McCuaig, F.D.; Misch, R.D.

1960-04-19

A brazing alloy can be used for bonding zirconium and its alloys to other metals, ceramics, and cermets, and consists of 6 to 9 wt.% Ni, 6 to 9 wn~.% Cr, Mo, or W, 0 to 7.5 wt.% Fe, and the balance Zr.

PROCESS FOR DISSOLVING BINARY URANIUM-ZIRCONIUM OR ZIRCONIUM-BASE ALLOYS

DOEpatents

Jonke, A.A.; Barghusen, J.J.; Levitz, N.M.

1962-08-14

A process of dissolving uranium-- zirconium and zircaloy alloys, e.g. jackets of fuel elements, with an anhydrous hydrogen fluoride containing from 10 to 32% by weight of hydrogen chloride at between 400 and 450 deg C., preferably while in contact with a fluidized inert powder, such as calcium fluoride is described. (AEC)

Nuclear fuel element

DOEpatents

Meadowcroft, Ronald Ross; Bain, Alastair Stewart

1977-01-01

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.

Zirconium behaviour during electrorefining of actinide-zirconium alloy in molten LiCl-KCl on aluminium cathodes

NASA Astrophysics Data System (ADS)

Meier, R.; Souček, P.; Malmbeck, R.; Krachler, M.; Rodrigues, A.; Claux, B.; Glatz, J.-P.; Fanghänel, Th.

2016-04-01

A pyrochemical electrorefining process for the recovery of actinides from metallic nuclear fuel based on actinide-zirconium alloys (An-Zr) in a molten salt is being investigated. In this process actinides are group-selectively recovered on solid aluminium cathodes as An-Al alloys using a LiCl-KCl eutectic melt at a temperature of 450 °C. In the present study the electrochemical behaviour of zirconium during electrorefining was investigated. The maximum amount of actinides that can be oxidised without anodic co-dissolution of zirconium was determined at a selected constant cathodic current density. The experiment consisted of three steps to assess the different stages of the electrorefining process, each of which employing a fresh aluminium cathode. The results indicate that almost a complete dissolution of the actinides without co-dissolution of zirconium is possible under the applied experimental conditions.

Nickel aluminide alloy suitable for structural applications

DOEpatents

Liu, Chain T.

1998-01-01

Alloys for use in structural applications based upon NiAl to which are added selected elements to enhance room temperature ductility and high temperature strength. Specifically, small additions of molybdenum produce a beneficial alloy, while further additions of boron, carbon, iron, niobium, tantalum, zirconium and hafnium further improve performance of alloys at both room temperature and high temperatures. A preferred alloy system composition is Ni--(49.1.+-.0.8%)Al--(1.0.+-.0.8%)Mo--(0.7.+-.0.5%)Nb/Ta/Zr/Hf--(nearly zero to 0.03%)B/C, where the % is at. % in each of the concentrations. All alloys demonstrated good oxidation resistance at the elevated temperatures. The alloys can be fabricated into components using conventional techniques.

A study into the impact of interface roughness development on mechanical degradation of oxides formed on zirconium alloys

NASA Astrophysics Data System (ADS)

Platt, P.; Wedge, S.; Frankel, P.; Gass, M.; Howells, R.; Preuss, M.

2015-04-01

As a cladding material used to encapsulate nuclear fuel pellets, zirconium alloys are the primary barrier separating the fuel and a pressurised steam or lithiated water environment. Degradation mechanisms such as oxidation can be the limiting factor in the life-time of the fuel assembly. Key to controlling oxidation, and therefore allowing increased burn-up of fuel, is the development of a mechanistic understanding of the corrosion process. In an autoclave, the oxidation kinetics for zirconium alloys are typically cyclical, with periods of accelerated kinetics being observed in steps of ∼2 μm oxide growth. These periods of accelerated oxidation are immediately preceded by the development of a layer of lateral cracks near the metal-oxide interface, which may be associated with the development of interface roughness. The present work uses scanning electron microscopy to carry out a statistical analysis of changes in the metal-oxide interface roughness between three different alloys at different stages of autoclave oxidation. The first two alloys are Zircaloy-4 and ZIRLO™ for which analysis is carried out at stages before, during and after first transition. The third alloy is an experimental low tin alloy, which under the same oxidation conditions and during the same time period does not appear to go through transition. Assessment of the metal-oxide interface roughness is primarily carried out based on the root mean square of the interface slope known as the Rdq parameter. Results show clear trends with relation to transition points in the corrosion kinetics. Discussion is given to how this relates to the existing mechanistic understanding of the corrosion process, and the components required for possible future modelling approaches.

Superconductivity in zirconium-rhodium alloys

NASA Technical Reports Server (NTRS)

Zegler, S. T.

1969-01-01

Metallographic studies and transition temperature measurements were made with isothermally annealed and water-quenched zirconium-rhodium alloys. The results clarify both the solid-state phase relations at the Zr-rich end of the Zr-Rh alloy system and the influence upon the superconducting transition temperature of structure and composition.

Fabrication and Evaluation of Titanium and Zirconium based Wires for use during Extended, Deep Space, Missions

NASA Technical Reports Server (NTRS)

Grugel, Richard N.

2006-01-01

Novel materials and designs are necessary for transport vessels and propulsion systems to fulfill NASA's vision of easier access to space and the expansion of human exploration beyond low-earth orbit. Spacecraft components must necessarily be lighter and stronger than their predecessors and will likely be required to serve new purposes. Furthermore, they must be resilient to the thermal, vacuum, and radiation environment of space for extended periods of time and may need to perform in the near proximity of a nuclear fuel source. To this end research has been initiated to fabricate novel, composite, wires based on titanium and zirconium pearlitic alloys. It is expected that the fabricated wire will well endure in the space environment with application as tethers, sail components, fasteners, and a myriad of other (including earth-based) uses. A background on pearlitic wire, novel alloy development, microstructural characterization, and initial mechanical testing results will be presented and discussed.

ZIRCONIUM-TITANIUM-BERYLLIUM BRAZING ALLOY

DOEpatents

Gilliland, R.G.; Patriarca, P.; Slaughter, G.M.; Williams, L.C.

1962-06-12

A new and improved ternary alloy is described which is of particular utility in braze-bonding parts made of a refractory metal selected from Group IV, V, and VI of the periodic table and alloys containing said metal as a predominating alloying ingredient. The brazing alloy contains, by weight, 40 to 50 per cent zirconium, 40 to 50 per cent titanium, and the balance beryllium in amounts ranging from 1 to 20 per cent, said alloy having a melting point in the range 950 to 1400 deg C. (AEC)

PROCESS OF DISSOLVING ZIRCONIUM ALLOYS

DOEpatents

Shor, R.S.; Vogler, S.

1958-01-21

A process is described for dissolving binary zirconium-uranium alloys where the uranium content is about 2%. In prior dissolution procedures for these alloys, an oxidizing agent was added to prevent the precipitation of uranium tetrafluoride. In the present method complete dissolution is accomplished without the use of the oxidizing agent by using only the stoichiometric amount or slight excess of HF required by the zirconium. The concentration of the acid may range from 2M to 10M and the dissolution is advatageously carried out at a temperature of 80 deg C.

Nickel aluminide alloy suitable for structural applications

DOEpatents

Liu, C.T.

1998-03-10

Alloys are disclosed for use in structural applications based upon NiAl to which are added selected elements to enhance room temperature ductility and high temperature strength. Specifically, small additions of molybdenum produce a beneficial alloy, while further additions of boron, carbon, iron, niobium, tantalum, zirconium and hafnium further improve performance of alloys at both room temperature and high temperatures. A preferred alloy system composition is Ni--(49.1{+-}0.8%)Al--(1.0{+-}0.8%)Mo--(0.7 + 0.5%)Nb/Ta/Zr/Hf--(nearly zero to 0.03%)B/C, where the % is at. % in each of the concentrations. All alloys demonstrated good oxidation resistance at the elevated temperatures. The alloys can be fabricated into components using conventional techniques. 4 figs.

ZIRCONIUM-CLADDING OF THORIUM

DOEpatents

Beaver, R.J.

1961-11-21

A method of cladding thorium with zirconium is described. The quality of the bond achieved between thorium and zirconium by hot-rolling is improved by inserting and melting a thorium-zirconium alloy foil between the two materials prior to rolling. (AEC)

Process for massively hydriding zirconium--uranium fuel elements

DOEpatents

Katz, N.H.

1973-12-01

A method is described of hydriding uranium-zirconium alloy by heating the alloy in a vacuum, introducing hydrogen and maintaining an elevated temperature until occurrence of the beta--delta phase transformation and isobarically cooling the composition. (Official Gazette)

Titanium-Zirconium-Nickel Alloy Inside Marshall's Electrostatic Levitator (ESL)

NASA Technical Reports Server (NTRS)

2003-01-01

This is a close-up of a sample of titanium-zirconium-nickel alloy inside the Electrostatic Levitator (ESL) vacuum chamber at NASA's Marshall Space Flight Center (MSFC). The ESL uses static electricity to suspend an object (about 3-4 mm in diameter) inside a vacuum chamber allowing scientists to record a wide range of physical properties without the sample contracting the container or any instruments, conditions that would alter the readings. Once inside the chamber, a laser heats the sample until it melts. The laser is then turned off and the sample cools, changing from a liquid drop to a solid sphere. Since 1977, the ESL has been used at MSFC to study the characteristics of new metals, ceramics, and glass compounds. Materials created as a result of these tests include new optical materials, special metallic glasses, and spacecraft components.

CHARACTERISTICS OF ANODIC AND CORROSION FILMS ON ZIRCONIUM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Misch, R.D.

1960-05-01

Zirconium anodizes similarly to tungsten in respect to the change of interference colors with applied voltage. However, the oxide layer on tungsten cannot reach as great a thickness. Hafnium does not anodize in the same way as zirconium but is similar to tantalum. By measuring the interference color and capacitative thicknesses on zirconium (Grades I and III) and a 2.5 wt.% tin ailoy, the film was found to grow less rapidly in terms of capacitance than in terms of iaterference colors. This was interpreted to mean that cracks develop in the oxide as it thickens. The effect was most pronouncedmore » on Grade III zirconium and least pronounced on the tin alloy. The reduction in capacitative thickness was especially noticeable when white oxide appeared. Comparative measurements on Grade I zirconium and 2.5 wt.% tin alloy indicated that the thickness of the oxide film on the tin alloy (after 16 hours in water) increased more rapidly with temperature than the film on zirconium. Tin is believed to act in ways to counteract the tendency of the oxide to form cracks, and to produce vacancies which promote ionic diffusion. (auth)« less

Effects of hydrogen on thermal creep behaviour of Zircaloy fuel cladding

NASA Astrophysics Data System (ADS)

Suman, Siddharth; Khan, Mohd Kaleem; Pathak, Manabendra; Singh, R. N.

2018-01-01

Zirconium alloys are extensively used for nuclear fuel cladding. Creep is one of the most likely degradation mechanisms for fuel cladding during reactor operating and repository conditions. Fuel cladding tubes undergo waterside corrosion during service and hydrogen is produced as a result of it-a fraction of which is picked up by cladding. Hydrogen remains in solid solution up to terminal solid solubility and it precipitates as brittle hydride phase in the zirconium metal matrix beyond this limiting concentration. Hydrogen, either in solid solution or as precipitated hydride, alters the creep behaviour of zirconium alloys. The present article critically reviews the influence of hydrogen on thermal creep behaviour of zirconium alloys, develops the systematic understanding of this multifaceted phenomenon, and delineates the thrust areas which require further investigations.

Pu-Zr alloy for high-temperature foil-type fuel

DOEpatents

McCuaig, Franklin D.

1977-01-01

A nuclear reactor fuel alloy consists essentially of from slightly greater than 7 to about 4 w/o zirconium, balance plutonium, and is characterized in that the alloy is castable and is rollable to thin foils. A preferred embodiment of about 7 w/o zirconium, balance plutonium, has a melting point substantially above the melting point of plutonium, is rollable to foils as thin as 0.0005 inch thick, and is compatible with cladding material when repeatedly cycled to temperatures above 650.degree. C. Neutron reflux densities across a reactor core can be determined with a high-temperature activation-measurement foil which consists of a fuel alloy foil core sandwiched and sealed between two cladding material jackets, the fuel alloy foil core being a 7 w/o zirconium, plutonium foil which is from 0.005 to 0.0005 inch thick.

Pu-ZR Alloy high-temperature activation-measurement foil

DOEpatents

McCuaig, Franklin D.

1977-08-02

A nuclear reactor fuel alloy consists essentially of from slightly greater than 7 to about 4 w/o zirconium, balance plutonium, and is characterized in that the alloy is castable and is rollable to thin foils. A preferred embodiment of about 7 w/o zirconium, balance plutonium, has a melting point substantially above the melting point of plutonium, is rollable to foils as thin as 0.0005 inch thick, and is compatible with cladding material when repeatedly cycled to temperatures above 650.degree. C. Neutron flux densities across a reactor core can be determined with a high-temperature activation-measurement foil which consists of a fuel alloy foil core sandwiched and sealed between two cladding material jackets, the fuel alloy foil core being a 7 w/o zirconium, plutonium foil which is from 0.005 to 0.0005 inch thick.

Finite element analysis of the tetragonal to monoclinic phase transformation during oxidation of zirconium alloys

NASA Astrophysics Data System (ADS)

Platt, P.; Frankel, P.; Gass, M.; Howells, R.; Preuss, M.

2014-11-01

Corrosion is a key limiting factor in the degradation of zirconium alloys in light water reactors. Developing a mechanistic understanding of the corrosion process offers a route towards improving safety and efficiency as demand increases for higher burn-up of fuel. Oxides formed on zirconium alloys are composed of both monoclinic and meta-stable tetragonal phases, and are subject to a number of potential mechanical degradation mechanisms. The work presented investigates the link between the tetragonal to monoclinic oxide phase transformation and degradation of the protective character of the oxide layer. To achieve this, Abaqus finite element analysis of the oxide phase transformation has been carried out. Study of the change in transformation strain energy shows how relaxation of oxidation induced stress and fast fracture at the metal-oxide interface could destabilise the tetragonal phase. Central to this is the identification of the transformation variant most likely to form, and understanding why twinning of the transformed grain is likely to occur. Development of transformation strain tensors and analysis of the strain components allows some separation of dilatation and shear effects. Maximum principal stress is used as an indication of fracture in the surrounding oxide layer. Study of the stress distributions shows the way oxide fracture is likely to occur and the differing effects of dilatation and shape change. Comparison with literature provides qualitative validation of the finite element simulations.

Iron aluminide alloys with improved properties for high temperature applications

DOEpatents

McKamey, Claudette G.; Liu, Chain T.

1990-01-01

An improved iron aluminide alloy of the DO.sub.3 type that has increased room temperature ductility and improved high elevated temperature strength. The alloy system further is resistant to corrosive attack in the environments of advanced energy corrosion systems such as those using fossil fuels. The resultant alloy is relatively inexpensive as contrasted to nickel based and high nickel steels currently utilized for structural components. The alloy system consists essentially of 26-30 at. % aluminum, 0.5-10 at. % chromium, 0.02-0.3 at. % boron plus carbon, up to 2 at. % molybdenum, up to 1 at. % niobium, up to 0.5 at. % zirconium, up to 0.1 at. % yttrium, up to 0.5 at. % vanadium and the balance iron.

Iron aluminide alloys with improved properties for high temperature applications

DOEpatents

McKamey, C.G.; Liu, C.T.

1990-10-09

An improved iron aluminide alloy of the DO[sub 3] type is described that has increased room temperature ductility and improved high elevated temperature strength. The alloy system further is resistant to corrosive attack in the environments of advanced energy conversion systems such as those using fossil fuels. The resultant alloy is relatively inexpensive as contrasted to nickel based and high nickel steels currently utilized for structural components. The alloy system consists essentially of 26--30 at. % aluminum, 0.5--10 at. % chromium, 0.02--0.3 at. % boron plus carbon, up to 2 at. % molybdenum, up to 1 at. % niobium, up to 0.5 at. % zirconium, up to 0.1 at. % yttrium, up to 0.5 at. % vanadium and the balance iron. 3 figs.

Titanium-Zirconium-Nickel Alloy Inside Marshall's Electrostatic Levitator (ESL)

NASA Technical Reports Server (NTRS)

2003-01-01

This Photo, which appeared on the July cover of `Physics Today', is of the Electrostatic Levitator (ESL) at NASA's Marshall Space Flight Center (MSFC). The ESL uses static electricity to suspend an object (about 3-4 mm in diameter) inside a vacuum chamber allowing scientists to record a wide range of physical properties without the sample contracting the container or any instruments, conditions that would alter the readings. Once inside the chamber, a laser heats the sample until it melts. The laser is then turned off and the sample cools, changing from a liquid drop to a solid sphere. In this particular shot, the ESL contains a solid metal sample of titanium-zirconium-nickel alloy. Since 1977, the ESL has been used at MSFC to study the characteristics of new metals, ceramics, and glass compounds. Materials created as a result of these tests include new optical materials, special metallic glasses, and spacecraft components.

Nickel aluminide alloy for high temperature structural use

DOEpatents

Liu, Chain T.; Sikka, Vinod K.

1991-01-01

The specification discloses nickel aluminide alloys including nickel, aluminum, chromium, zirconium and boron wherein the concentration of zirconium is maintained in the range of from about 0.05 to about 0.35 atomic percent to improve the ductility, strength and fabricability of the alloys at 1200.degree. C. Titanium may be added in an amount equal to about 0.2 to about 0.5 atomic percent to improve the mechanical properties of the alloys and the addition of a small amount of carbon further improves hot fabricability.

Phase Transformation Temperatures and Solute Redistribution in a Quaternary Zirconium Alloy

NASA Astrophysics Data System (ADS)

Cochrane, C.; Daymond, M. R.

2018-05-01

This study investigates the phase stability and redistribution of solute during heating and cooling of a quaternary zirconium alloy, Excel (Zr-3.2Sn-0.8Mo-0.8Nb). Time-of-flight neutron diffraction data are analyzed using a novel Vegard's law-based approach to determine the phase fractions and location of substitutional solute atoms in situ during heating from room temperature up to 1050 °C. It is seen that this alloy exhibits direct nucleation of the β Zr phase from martensite during tempering, and stable retention of the β Zr phase to high temperatures, unlike other two-phase zirconium alloys. The transformation strains resulting from the α \\leftrightarrow β transformation are shown to have a direct impact on the development of microstructure and crystallographic texture.

Nuclear fuel alloys or mixtures and method of making thereof

DOEpatents

Mariani, Robert Dominick; Porter, Douglas Lloyd

2016-04-05

Nuclear fuel alloys or mixtures and methods of making nuclear fuel mixtures are provided. Pseudo-binary actinide-M fuel mixtures form alloys and exhibit: body-centered cubic solid phases at low temperatures; high solidus temperatures; and/or minimal or no reaction or inter-diffusion with steel and other cladding materials. Methods described herein through metallurgical and thermodynamics advancements guide the selection of amounts of fuel mixture components by use of phase diagrams. Weight percentages for components of a metallic additive to an actinide fuel are selected in a solid phase region of an isothermal phase diagram taken at a temperature below an upper temperature limit for the resulting fuel mixture in reactor use. Fuel mixtures include uranium-molybdenum-tungsten, uranium-molybdenum-tantalum, molybdenum-titanium-zirconium, and uranium-molybdenum-titanium systems.

Acidic ammonothermal growth of gallium nitride in a liner-free molybdenum alloy autoclave

NASA Astrophysics Data System (ADS)

Malkowski, Thomas F.; Pimputkar, Siddha; Speck, James S.; DenBaars, Steven P.; Nakamura, Shuji

2016-12-01

This paper discusses promising materials for use as internal, non-load bearing components as well as molybdenum-based alloys for autoclave structural components for an ammonothermal autoclave. An autoclave was constructed from the commercial titanium-zirconium-molybdenum (TZM) alloy and was found to be chemically inert and mechanically stable under acidic ammonothermal conditions. Preliminary seeded growth of GaN was demonstrated with negligible incorporation of transition metals (including molybdenum) into the grown material (<1017 cm-3). Molybdenum and TZM were exposed to a basic ammonothermal environment, leading to slight degradation through formation of molybdenum nitride powders on their surface at elevated temperatures (T>560 °C). The possibility of a 'universal', inexpensive, liner-free ammonothermal autoclave capable of exposure to basic and acidic chemistry is demonstrated.

[Influence of different types of posts and cores on color of IPS-Empress 2 crown].

PubMed

Li, Dong-fang; Yang, Jing-yuan; Yang, Xing-mei; Yang, Liu; Xu, Qiang; Guan, Hong-yu; Wan, Qian-bing

2007-10-01

To evaluate the influence of different types of posts and cores on the final color of the IPS-Emperss 2 crown. Five types of posts and cores (Cerapost with Empress cosmo, Cerapost with composite resin, gilded Ni-Cr alloy, gold alloy and Ni-Cr alloy) were made. The shifts in color of three points of IPS-Empress 2 crown surface (cervical, middle and incisal) with different posts and cores was measured with a spectroradiometer (PR-650). The L* a* b* values of zirconium oxide and gilded Ni-Cr alloy posts and cores with ceramic crown were the highest. The L* a* values of zirconium oxide posts composite cores were higher while the b* values were lower. The L* a* b* values of Ni-Cr alloy were lower than that of gold alloy and were the lowest. In combination with IPS-Empress 2 crown, zirconium oxide posts are suitable for routine use in the anterior dentition, and gilded Ni-Cr alloy and gold alloy posts and cores can be recommended for clinical practice. Ni-Cr alloy posts and cores can not be recommended for clinical practice.

THE ANALYSIS OF URANIUM-ZIRCONIUM ALLOYS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Milner, G.W.C.; Skewies, A.F.

1953-03-01

A satisfactory procedure is described for the analysis of uranium-zirconium alloys containing up to 25% zirconium. It is based on the separation of the zirconium from the uranium by dissolving the cupferron complex of the former element into chloroform. After the evaporation of the solvent from the combined organic extracts, the residue is ignited to zirconium oxide. The latter is then re-dissolved and zirconium is separated from other elements co-extracted in the solvent extraction procedure by precipitation with mandelic acid. The zirconium mandelate is finally ignited to oxide at 960 deg C. The uranium is separated from the aqueous solutionmore » remaining from the cupferron extraction by precipitating with tannin at a pH of 8; the precipitate being removed by filtration and then ignited a t 800 deg C. The residue is dissolved in nitric acid and the uranium is finally determined by precipitating as ammonium diuranate and then igniting to U{sub 3}O{sub 8}. (auth)« less

About structural phase state of coating based on zirconium oxide formed by microplasma oxidation method

NASA Astrophysics Data System (ADS)

Gubaidulina, Tatiana A.; Sergeev, Viktor P.; Kuzmin, Oleg S.; Fedorischeva, Marina V.; Kalashnikov, Mark P.

2017-12-01

The oxide-ceramic coating based of zirconium oxide is formed by the method of microplasma oxidation. The producing modes of the oxide layers on E110 zirconium alloy are under testing. It was found that using microplasma treatment of E110 zirconium in aluminosilicate electrolyte makes possible the formation of porous oxide-ceramic coatings based on zirconium alloyed by aluminum and niobium. The study is focused on the modes how to form heat-shielding coatings with controlled porosity and minimal amount of microcracks. The structural-phase state of the coating is studied by X-ray diffraction analysis and scanning electron microscopy (SEM). It was found that the ratio of the monoclinic and tetragonal phases changes with the change occurring in the coating formation modes.

Initial results of metal waste form development activities at ANL-West

DOE Office of Scientific and Technical Information (OSTI.GOV)

Keiser, D.D. Jr.; Westphal, B.R.; Hersbt, R.S.

1997-10-01

Argonne National Laboratory is developing a metal alloy to contain metallic waste constituents from the electrometallurgical treatment of spent nuclear fuel. This alloy will contain stainless steel (from stainless steel-clad fuel elements), {approximately}15 wt.% zirconium (from alloy fuel), fission products noble to the process (e.g., Ru, Pd, Tc, etc.), and minor amounts of actinides. The alloy will serve as a final waste form for these components and will be disposed of in a geologic repository. The alloy ingot is produced in an induction furnace situated in a hot cell using Ar cover gas. This paper discusses results from the meltingmore » campaigns that have been initiated at ANL-West to generate the metal waste form using actual process materials. In addition, metal waste form samples have been doped with Tc and selected actinides and are described in the context of how elements of interest partition between various phases in the alloy and how this distribution of elements in the alloy may affect the leaching behavior of the components in an aqueous environment. 3 refs.« less

Initial results of metal waste-form development activities at ANL-West

DOE Office of Scientific and Technical Information (OSTI.GOV)

Keiser, D.D. Jr.; Westphal, B.R.; Herbst, R.S.

1997-12-01

Argonne National Laboratory (ANL) is developing a metal alloy to contain metallic waste constituent residual from the electrometallurgical treatment of spent nuclear fuel. This alloy will contain stainless steel (from stainless steel-clad fuel elements), {approximately} 15 wt% zirconium (from alloy fuel), fission products noble to the process (e.g., ruthenium, palladium, technetium, etc.), and minor amounts of actinides. The alloy will serve as a final waste form for these components and will be disposed of in a geologic repository. The alloy ingot is produced in an induction furnace situated in a hot cell using argon cover gas. This paper discusses resultsmore » from the melting campaigns that have been initiated at ANL-West to generate the metal waste form using actual process materials. In addition, metal waste form samples have been doped with technetium and selected actinides and are described in the context of how elements of interest partition between various phases in the alloy and how this distribution of elements in the alloy may affect the leaching behavior of the components in an aqueous environment.« less

Zirconium alloys with small amounts of iron and copper or nickel show improved corrosion resistance in superheated steam

NASA Technical Reports Server (NTRS)

Greenberg, S.; Youngdahl, C. A.

1967-01-01

Heat treating various compositions of zirconium alloys improve their corrosion resistance to superheated steam at temperatures higher than 500 degrees C. This increases their potential as fuel cladding for superheated-steam nuclear-fueled reactors as well as in autoclaves operating at modest pressures.

METHOD FOR ANNEALING AND ROLLING ZIRCONIUM-BASE ALLOYS

DOEpatents

Picklesimer, M.L.

1959-07-14

A fabrication procedure is presented for alpha-stabilized zirconium-base alloys, and in particular Zircaloy-2. The alloy is initially worked at a temperature outside the alpha-plus-beta range (810 to 970 deg ), held at a temperature above 970 deg C for 30 minutes and cooled rapidly. The alloy is then cold-worked to reduce the size at least 20% and annealed at a temperature from 700 to 810 deg C. This procedure serves both to prevent the formation of stringers and to provide a randomly oriented crystal structure.

The Effect of Boron and Zirconium on the Structure and Tensile Properties of the Cast Nickel-Based Superalloy ATI 718Plus

NASA Astrophysics Data System (ADS)

Hosseini, Seyed Ali; Abbasi, Seyed Mehdi; Madar, Karim Zangeneh

2018-04-01

The effects of boron and zirconium on cast structure, hardness, and tensile properties of the nickel-based superalloy 718Plus were investigated. For this purpose, five alloys with different contents of boron and zirconium were cast via vacuum induction melting and then purified via vacuum arc remelting. Microstructural analysis by light-optical microscope and scanning electron microscope equipped with energy-dispersive x-ray spectroscopy and phase studies by x-ray diffraction analysis were performed. The results showed that boron and zirconium tend to significantly reduce dendritic arm spacing and increase the amount of Laves, Laves/gamma eutectic, and carbide phases. It was also found that boron led to the formation of B4C and (Cr, Fe, Mo, Ni, Ti)3B2 phases and zirconium led to the formation of intermetallic phases and ZrC carbide. In the presence of boron and zirconium, the hardness and its difference between dendritic branches and inter-dendritic spaces increased by concentrating such phases as Laves in the inter-dendritic spaces. These elements had a negative effect on tensile properties of the alloy, including ductility and strength, mainly because of the increase in the Laves phase. It should be noted that the largest degradation of the tensile properties occurred in the alloys containing the maximum amount of zirconium.

Effect of Copper and Zirconium Addition on Properties of Fe-Co-Si-B-Nb Bulk Metallic Glasses

NASA Astrophysics Data System (ADS)

Ikram, Haris; Khalid, Fazal Ahmad; Akmal, Muhammad; Abbas, Zameer

2017-07-01

In this research work, iron-based bulk metallic glasses (BMGs) have been fabricated, characterized and compared with Fe-Si alloy. BMG alloys of composition ((Fe0.6Co0.4)0.75B0.20Si0.05)96Nb4) were synthesized by suction casting technique using chilled copper die. Effect of copper and zirconium addition on magnetic, mechanical, thermal and electrochemical behavior of ((Fe0.6Co0.4)0.75B0.20Si0.05)96Nb4 BMGs was investigated. Furthermore, effect of annealing on nano-crystallization and subsequently on magnetic and mechanical behavior was also analyzed. Amorphousness of structure was evidenced by XRD analysis and microscopic visualization, whereas nano-crystallization behavior was identified by peak broadening of XRD patterns. Magnetic properties, measured by vibrating sample magnetometer, were found to be improved for as-cast BMG alloys by copper addition and further enhanced by nano-crystallization after annealing. Mechanical properties were observed to be increased by zirconium addition while slightly declined by copper addition. Potentiodynamic polarization analysis manifested the positive role of zirconium in enhancing corrosion resistance of BMGs in acidic, basic and brine mediums. Moreover, mechanical properties and corrosion analysis results affirmed the superiority of BMG alloys over Fe-Si alloy.

Filler wire for aluminum alloys and method of welding

NASA Technical Reports Server (NTRS)

Bjorkman, Jr., Gerald W. O. (Inventor); Cho, Alex (Inventor); Russell, Carolyn K. (Inventor)

2003-01-01

A weld filler wire chemistry has been developed for fusion welding 2195 aluminum-lithium. The weld filler wire chemistry is an aluminum-copper based alloy containing high additions of titanium and zirconium. The additions of titanium and zirconium reduce the crack susceptibility of aluminum alloy welds while producing good weld mechanical properties. The addition of silver further improves the weld properties of the weld filler wire. The reduced weld crack susceptibility enhances the repair weldability, including when planishing is required.

Mechanical properties of zirconium alloys and zirconium hydrides predicted from density functional perturbation theory

DOE PAGES

Weck, Philippe F.; Kim, Eunja; Tikare, Veena; ...

2015-10-13

Here, the elastic properties and mechanical stability of zirconium alloys and zirconium hydrides have been investigated within the framework of density functional perturbation theory. Results show that the lowest-energy cubic Pn-3m with combining macron]m polymorph of δ-ZrH 1.5 does not satisfy all the Born requirements for mechanical stability, unlike its nearly degenerate tetragonal P4 2/ mcm polymorph. Elastic moduli predicted with the Voigt–Reuss–Hill approximations suggest that mechanical stability of α-Zr, Zr-alloy and Zr-hydride polycrystalline aggregates is limited by the shear modulus. According to both Pugh's and Poisson's ratios, α-Zr, Zr-alloy and Zr-hydride polycrystalline aggregates can be considered ductile. The Debyemore » temperatures predicted for γ-ZrH, δ-ZrH 1.5 and ε-ZrH 2 are θ D = 299.7, 415.6 and 356.9 K, respectively, while θ D = 273.6, 284.2, 264.1 and 257.1 K for the α-Zr, Zry-4, ZIRLO and M5 matrices, i.e. suggesting that Zry-4 possesses the highest micro-hardness among Zr matrices.« less

Influence of chemical composition of zirconium alloy E110 on embrittlement under LOCA conditions - Part 1: Oxidation kinetics and macrocharacteristics of structure and fracture

NASA Astrophysics Data System (ADS)

Nikulin, S. A.; Rozhnov, A. B.; Belov, V. A.; Li, E. V.; Glazkina, V. S.

2011-11-01

Exploratory investigations of the influence of alloying and impurity content in the E110 alloy cladding tubes on the behavior under conditions of Loss of Coolant Accidents (LOCA) has been performed. Three alloys of E110 type have been tested: E110 alloy of nominal composition Zr-1%Nb (E110), E110 alloy of modified composition Zr-1%Nb-0.12%Fe-0.13%O (E110M), E110 alloy of nominal composition Zr-1%Nb with reduced impurity content (E110G). Alloys E110 and E110M were manufactured on the electrolytic basis and alloy E110G was manufactured on the basis of zirconium sponge. The high temperature oxidation tests in steam ( T = 1100 °C, 18% of equivalent cladding reacted (ECR)) have been conducted, kinetics of oxidation was investigated. Quantitative research of structure and fracture macrocharacteristics was performed by means of optical and electron microscopy. The results received were compared with the residual ductility of specimens. The results of the investigation showed the existence of "breakaway oxidation" kinetics and white spalling oxide in E110 and E110M alloys while the specimen oxidation kinetics in E110G alloy was characterized by a parabolic law and specimens had a dense black oxide. Oxygen and iron alloying in the E110 alloy positively changed the macrocharacteristics of structure and fracture. However, in general, it did not improve the resistance to embrittlement in LOCA conditions apparently because of a strong impurity influence caused by electrolytic process of zirconium production.

Capture of Tritium Released from Cladding in the Zirconium Recycle Process

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spencer, Barry B.; Walker, T. B.; Bruffey, S. H.

2016-08-31

Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when themore » solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using nonradioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.« less

Capture of Tritium Released from Cladding in the Zirconium Recycle Process

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spencer, Barry B.; Walker, T. B.; Bruffey, Stephanie H.

2016-08-31

This report is issued as the first revision to FCRD-MRWFD-2016-000297. Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-basedmore » cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using non-radioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.« less

Development of Cold Spray Coatings for Accident-Tolerant Fuel Cladding in Light Water Reactors

NASA Astrophysics Data System (ADS)

Maier, Benjamin; Yeom, Hwasung; Johnson, Greg; Dabney, Tyler; Walters, Jorie; Romero, Javier; Shah, Hemant; Xu, Peng; Sridharan, Kumar

2018-02-01

The cold spray coating process has been developed at the University of Wisconsin-Madison for the deposition of oxidation-resistant coatings on zirconium alloy light water reactor fuel cladding with the goal of improving accident tolerance during loss of coolant scenarios. Coatings of metallic (Cr), alloy (FeCrAl), and ceramic (Ti2AlC) materials were successfully deposited on zirconium alloy flats and cladding tube sections by optimizing the powder size, gas preheat temperature, pressure and composition, and other process parameters. The coatings were dense and exhibited excellent adhesion to the substrate. Evaluation of the samples after high-temperature oxidation tests at temperatures up to 1300°C showed that the cold spray coatings significantly mitigate oxidation kinetics because of the formation of thin passive oxide layers on the surface. The results of the study indicate that the cold spray coating process is a viable near-term option for developing accident-tolerant zirconium alloy fuel cladding.

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

NASA Technical Reports Server (NTRS)

Wagner, John A.; Shenoy, R. N.

1991-01-01

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.

THE CREEP BEHAVIOUR OF THE MAGNESIUM-ZIRCONIUM ALLOY ZA AT 400 AND 450 C IN CARBON DIOXIDE CONTAINING /approximately equals/200 PPM MOISTURE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kent, R.P.; Wells, T.C.

1963-03-01

The creep behavior of the magnesium-zirconium alloy ZA was studied in tests of up to 5600 hr duration at 400 deg C and up to 12 600 hr duration at 450 deg C, in an atmosphere of carbon dioxide containing approximately 200 ppm water. The accompanying microstructural changes were observed by optical and electron microscopy. The alloy is stronger at 450 deg C than at 400 deg C and additional strengthening obtains from prestraining at 250 deg C prior to creep-testing. In stress rupture tests at 200 deg C subsequent to creep-testing, the time to rupture and the rupture ductilitymore » are lower in specimens previously tested at 450 deg C than in those tested at 400 deg C. The increase in creep strength at 450 deg C, and subsequent loss of ductility, are attributed principally to the precipitation of a zirconium-rich phase, tentatively identified as epsilon - zirconium hydride, which forms both intragranularly (as ribbons and thin hexagonal plates) and as intergranular particles. (auth)« less

TERNARY ALLOY-CONTAINING PLUTONIUM

DOEpatents

Waber, J.T.

1960-02-23

Ternary alloys of uranium and plutonium containing as the third element either molybdenum or zirconium are reported. Such alloys are particularly useful as reactor fuels in fast breeder reactors. The alloy contains from 2 to 25 at.% of molybdenum or zirconium, the balance being a combination of uranium and plutonium in the ratio of from 1 to 9 atoms of uranlum for each atom of plutonium. These alloys are prepared by melting the constituent elements, treating them at an elevated temperature for homogenization, and cooling them to room temperature, the rate of cooling varying with the oomposition and the desired phase structure. The preferred embodiment contains 12 to 25 at.% of molybdenum and is treated by quenching to obtain a body centered cubic crystal structure. The most important advantage of these alloys over prior binary alloys of both plutonium and uranium is the lack of cracking during casting and their ready machinability.

METHOD OF DISSOLVING REFRACTORY ALLOYS

DOEpatents

Helton, D.M.; Savolainen, J.K.

1963-04-23

This patent relates to the dissolution of alloys of uranium with zirconium, thorium, molybdenum, or niobium. The alloy is contacted with an anhydrous solution of mercuric chloride in a low-molecular-weight monohydric alcohol to produce a mercury-containing alcohol slurry. The slurry is then converted to an aqueous system by adding water and driving off the alcohol. The resulting aqueous slurry is electrolyzed in the presence of a mercury cathode to remove the mercury and produce a uranium-bearing aqueous solution. This process is useful for dissolving irradiated nuclear reactor fuels for radiochemical reprocessing by solvent extraction. In addition, zirconium-alloy cladding is selectively removed from uranium dioxide fuel compacts by this means. (AEC)

Superalloys with improved weldability for high temperature applications

DOEpatents

Seth, Brij B.; George, Easo P.; Babu, Sudarsanum S.; Goodwin, Gene M.; David, Stanislaus A.; Moyer, Carol E.

2001-01-01

A cast nickel-base superalloy component (10) is made having a composition containing small amounts of both boron and zirconium which are effective in combination to provide increased weldability, where such alloy is adapted for welding by weld (18) to a second superalloy piece, where the two pieces are firmly bonded together and have a Sigmajig transverse stress value (16) greater than 137.9 million Newtons per square meter.

Electrochemical Study of Corrosion Phenomena in Zirconium Alloys

DTIC Science & Technology

2005-06-01

required reaction rates [1.1]. Based predominantly on this fact, zirconium alloys were chosen to sheath, or clad, the fuel. With the increasing demand...cathode or anode. As the oxidation and reduction reactions occur, a galvanic cell is developed. The electrons on the anode are released from the metallic...matrix as the ions are released into the aqueous solution in the initial half-cell reaction . The second half-cell reaction , taking place on the

Optimisation and characterisation of tungsten thick coatings on copper based alloy substrates

NASA Astrophysics Data System (ADS)

Riccardi, B.; Montanari, R.; Casadei, M.; Costanza, G.; Filacchioni, G.; Moriani, A.

2006-06-01

Tungsten is a promising armour material for plasma facing components of nuclear fusion reactors because of its low sputter rate and favourable thermo-mechanical properties. Among all the techniques able to realise W armours, plasma spray looks particularly attractive owing to its simplicity and low cost. The present work concerns the optimisation of spraying parameters aimed at 4-5 mm thick W coating on copper-chromium-zirconium (Cu,Cr,Zr) alloy substrates. Characterisation of coatings was performed in order to assess microstructure, impurity content, density, tensile strength, adhesion strength, thermal conductivity and thermal expansion coefficient. The work performed has demonstrated the feasibility of thick W coatings on flat and curved geometries. These coatings appear as a reliable armour for medium heat flux plasma facing component.

ZIRCONIUM ALLOY

DOEpatents

Wilhelm, H.A.; Ames, D.P.

1959-02-01

A binary zirconiuin--antimony alloy is presented which is corrosion resistant and hard containing from 0.07% to 1.6% by weight of Sb. The alloys have good corrosion resistance and are useful in building equipment for the chemical industry.

Study of metal corrosion using ac impedance techniques in the STS launch environment

NASA Technical Reports Server (NTRS)

Calle, Luz M.

1989-01-01

AC impedance measurements were performed to investigate the corrosion resistance of 19 alloys under conditions similar to the STS launch environment. The alloys were: Zirconium 702, Hastelloy C-22, Inconel 625, Hastelloy C-276, Hastelloy C-4, Inconel 600, 7Mo + N, Ferralium 255, Inco Alloy G-3, 20Cb-3, SS 904L, Inconel 825, SS 304LN, SS 316L, SS 317L, ES 2205, SS 304L, Hastelloy B-2, and Monel 400. AC impedance data were gathered for each alloy after one hour immersion time in each of the following three electrolyte solutions: 3.55 percent NaCl, 3.55 percent NaCl-0.1N HCl, and 3.55 percent NaCl-1.0N HCl. The data were analyzed qualitatively using the Nyquist plot and quantitatively using the Bode plot. Polarization resistance, Rp, values were obtained using the Bode plot. Zirconium 702 was the most corrosion resistant alloy in the three electrolytes. The ordering of the other alloys according the their resistance to corrosion varied as the concentration of hydrochloric acid in the electrolyte increased. The corrosion resistance of Zirconium 702 and Ferralium 255 increased as the concentration of hydrochloric acid in the electrolyte increased. The corrosion resistance of the other 17 alloys decreased as the concentration of the hyrdochloric acid in the electrolyte increased.

Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Qu, Jun; Cooley, Kevin M.; Shaw, Austin H.

In the cores of pressurized water nuclear reactors, water-flow induced vibration is known to cause claddings on the fuel rods to rub against their supporting grids. Such grid-to-rod-fretting (GTRF) may lead to fretting wear-through and the leakage of radioactive species. The surfaces of actual zirconium alloy claddings in a reactor are inevitably oxidized in the high-temperature pressurized water, and some claddings are even pre-oxidized. As a result, the wear process of the surface oxide film is expected to be quite different from the zirconium alloy substrate. In this paper, we attempt to measure the wear coefficients of zirconium claddings withoutmore » and with pre-oxidation rubbing against grid samples using a bench-scale fretting tribometer. Results suggest that the volumetric wear coefficient of the pre-oxidized cladding is 50 to 200 times lower than that of the untreated cladding. In terms of the linear rate of wear depth, the pre-oxidized alloy wears about 15 times more slowly than the untreated cladding. Finally, fitted with the experimentally-determined wear rates, a stage-wise GTRF engineering wear model demonstrates good agreement with in-reactor experience in predicting the trend of cladding lives.« less

Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation

DOE PAGES

Qu, Jun; Cooley, Kevin M.; Shaw, Austin H.; ...

2016-03-16

In the cores of pressurized water nuclear reactors, water-flow induced vibration is known to cause claddings on the fuel rods to rub against their supporting grids. Such grid-to-rod-fretting (GTRF) may lead to fretting wear-through and the leakage of radioactive species. The surfaces of actual zirconium alloy claddings in a reactor are inevitably oxidized in the high-temperature pressurized water, and some claddings are even pre-oxidized. As a result, the wear process of the surface oxide film is expected to be quite different from the zirconium alloy substrate. In this paper, we attempt to measure the wear coefficients of zirconium claddings withoutmore » and with pre-oxidation rubbing against grid samples using a bench-scale fretting tribometer. Results suggest that the volumetric wear coefficient of the pre-oxidized cladding is 50 to 200 times lower than that of the untreated cladding. In terms of the linear rate of wear depth, the pre-oxidized alloy wears about 15 times more slowly than the untreated cladding. Finally, fitted with the experimentally-determined wear rates, a stage-wise GTRF engineering wear model demonstrates good agreement with in-reactor experience in predicting the trend of cladding lives.« less

High-intensity low energy titanium ion implantation into zirconium alloy

NASA Astrophysics Data System (ADS)

Ryabchikov, A. I.; Kashkarov, E. B.; Pushilina, N. S.; Syrtanov, M. S.; Shevelev, A. E.; Korneva, O. S.; Sutygina, A. N.; Lider, A. M.

2018-05-01

This research describes the possibility of ultra-high dose deep titanium ion implantation for surface modification of zirconium alloy Zr-1Nb. The developed method based on repetitively pulsed high intensity low energy titanium ion implantation was used to modify the surface layer. The DC vacuum arc source was used to produce metal plasma. Plasma immersion titanium ions extraction and their ballistic focusing in equipotential space of biased electrode were used to produce high intensity titanium ion beam with the amplitude of 0.5 A at the ion current density 120 and 170 mA/cm2. The solar eclipse effect was used to prevent vacuum arc titanium macroparticles from appearing in the implantation area of Zr sample. Titanium low energy (mean ion energy E = 3 keV) ions were implanted into zirconium alloy with the dose in the range of (5.4-9.56) × 1020 ion/cm2. The effect of ion current density, implantation dose on the phase composition, microstructure and distribution of elements was studied by X-ray diffraction, scanning electron microscopy and glow-discharge optical emission spectroscopy, respectively. The results show the appearance of Zr-Ti intermetallic phases of different stoichiometry after Ti implantation. The intermetallic phases are transformed from both Zr0.7Ti0.3 and Zr0.5Ti0.5 to single Zr0.6Ti0.4 phase with the increase in the implantation dose. The changes in phase composition are attributed to Ti dissolution in zirconium lattice accompanied by the lattice distortions and appearance of macrostrains in intermetallic phases. The depth of Ti penetration into the bulk of Zr increases from 6 to 13 μm with the implantation dose. The hardness and wear resistance of the Ti-implanted zirconium alloy were increased by 1.5 and 1.4 times, respectively. The higher current density (170 mA/cm2) leads to the increase in the grain size and surface roughness negatively affecting the tribological properties of the alloy.

On the control of the crystallographic texture in cladding tubes from Zr-based alloys for nuclear reactor

NASA Astrophysics Data System (ADS)

Isaenkova, M.; Perlovich, Yu.; Fesenko, V.

2016-10-01

This paper summarizes researches of authors, directed to the development of the methodological basis of X-ray studies as applied to zirconium alloys and on the systematization of new experimental results obtained using developed methods. The paper describes regularities of crystallographic texture formation in cladding tubes from zirconium alloys and their substructure inhomogeneity at various stages of manufacture, i.e. at hot and cold deformation, recrystallization, phase transformations and interaction of the above processes. The special attention is payed to possibilities of control the crystallographic texture of tubes at successive stages of their technological treatment.

Systematic technology evaluation program for SiC/SiC composite-based accident-tolerant LWR fuel cladding and core structures: Revision 2015

DOE Office of Scientific and Technical Information (OSTI.GOV)

Katoh, Yutai; Terrani, Kurt A.

2015-08-01

Fuels and core structures in current light water reactors (LWR’s) are vulnerable to catastrophic failure in severe accidents as unfortunately evidenced by the March 2011 Fukushima Dai-ichi Nuclear Power Plant Accident. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures. Zr alloys are the primary material in LWR cores except for the fuel itself. Therefore, alternative materials with reduced oxidation kinetics as compared to zirconium alloys are sought to enable enhanced accident-tolerant fuels and cores.

Abrasive wear behavior of in-situ RZ5-10wt%TiC composite

NASA Astrophysics Data System (ADS)

Mehra, Deepak; Mahapatra, M. M.; Harsha, S. P.

2018-05-01

RZ5 Magnesium alloys containing zinc, rare earth and zirconium are well-known to have high specific strength, good creep resistance widely used in aerospace components. The incorporation of hard ceramic strengthens RZ5 mg alloy. The RZ5-10wt%TiC composite has been fabricated in situ using RZ5 mg alloy as matrix and TiC as reinforcement by self propagating high temperature synthesis (SHS) technique. This paper investigates the abrasive wear behavior of RZ5-10wt%TiC. Tests were performed using pin-on-disc apparatus against 600 grit abrasive paper by varying the sliding distance and applied load. The results showed improvement in the wear resistance of testing composite as compared to the unreinforced RZ5 Mg alloy. The coefficient of friction and weight loss increased linearly as applied load and sliding distance increased. The field emission scanning electron microscopic (FESEM) showed dominate wear mechanisms: abrasion, ploughing grooves.

High temperature, low-cycle fatigue of copper-base alloys in argon. Part 1: Preliminary results for 12 alloys at 1000 F (538 C)

NASA Technical Reports Server (NTRS)

Conway, J. B.; Stentz, R. H.; Berling, J. T.

1973-01-01

Short-term tensile evaluations at room temperature and 538 C and low-cycle fatigue evaluations at 538 C are presented for the following materials: Zirconium copper-annealed, Zirconium copper-1/4 hard, Zirconium copper-1/2 hard, Tellurium copper-1/2 hard, Chromium copper-SA and aged, OFHC copper-hard, OFHC copper-1/4 hard, OFHC copper-annealed, Silver-as drawn, Zr-Cr-Mg copper-SA, CW and aged, Electroformed copper-30-35 ksi, and Co-Be-Zr- copper-SA, aged. A total of 50 tensile tests and 76 low-cycle fatigue tests were performed using a strain rate of 0.2 percent per second.

Acoustic emission characteristics of copper alloys under low-cycle fatigue conditions

NASA Technical Reports Server (NTRS)

Krampfner, Y.; Kawamoto, A.; Ono, K.; Green, A.

1975-01-01

The acoustic emission (AE) characteristics of pure copper, zirconium-copper, and several copper alloys were determined to develop nondestructive evaluation schemes of thrust chambers through AE techniques. The AE counts rms voltages, frequency spectrum, and amplitude distribution analysis evaluated AE behavior under fatigue loading conditions. The results were interpreted with the evaluation of wave forms, crack propagation characteristics, as well as scanning electron fractographs of fatigue-tested samples. AE signals at the beginning of a fatigue test were produced by a sample of annealed alloys. A sample of zirconium-containing alloys annealed repeatedly after each fatigue loading cycle showed numerous surface cracks during the subsequent fatigue cycle, emitting strong-burst AE signals. Amplitude distribution analysis exhibits responses that are characteristic of certain types of AE signals.

Submersion Quenching of Undercooled Liquid Metals in an Electrostatic Levitator

NASA Technical Reports Server (NTRS)

SanSoucie, Michael P.; Rogers, Jan R.

2016-01-01

The NASA Marshall Space Flight Center (MSFC) electrostatic levitation (ESL) laboratory has a long history of providing materials research and thermophysical property data. The laboratory has recently added a new capability, a rapid quench system. This system allows samples to be dropped into a quench vessel that can be filled with a low melting point material, such as a gallium or indium alloy. Thereby allowing rapid quenching of undercooled liquid metals and alloys. This is the first submersion quench system inside an electrostatic levitator. The system has been tested successfully with samples of zirconium, iron-cobalt alloys, titanium-zirconium-nickel alloys, and silicon-cobalt alloys. This rapid quench system will allow materials science studies of undercooled materials and new materials development, including studies of metastable phases and transient microstructures. In this presentation, the system is described and some initial results are presented.

Impact Toughness and Heat Treatment for Cast Aluminum

NASA Technical Reports Server (NTRS)

Lee, Jonathan A (Inventor)

2016-01-01

A method for transforming a cast component made of modified aluminum alloy by increasing the impact toughness coefficient using minimal heat and energy. The aluminum alloy is modified to contain 0.55%-0.60% magnesium, 0.10%-0.15% titanium or zirconium, less than 0.07% iron, a silicon-tomagnesium product ratio of 4.0, and less than 0.15% total impurities. The shortened heat treatment requires an initial heating at 1,000deg F. for up to I hour followed by a water quench and a second heating at 350deg F. to 390deg F. for up to I hour. An optional short bake paint cycle or powder coating process further increase.

High temperature, low-cycle fatigue of copper-base alloys in argon. Part 2: Zirconium-copper at 482, 538 and 593 C

NASA Technical Reports Server (NTRS)

Conway, J. B.; Stentz, R. H.; Berling, J. T.

1973-01-01

Zirconium-copper (1/2 hard) was tested in argon over the temperature range from 482 to 593 C in an evaluation of short-term tensile and low-cycle fatigue behavior. The effect of strain rate on the tensile properties was evaluated at 538 C and in general it was found that the yield and ultimate strengths increased as the strain rate was increased from 0.0004 to 0.01/sec. Ductility was essentially insensitive to strain rate in the case of the zirconium-copper alloy. Strain-rate and hold-time effects on the low cycle fatigue behavior of zirconium-copper were evaluated in argon at 538 C. These effects were as expected in that decreased fatigue life was noted as the strain rate decreased and when hold times were introduced into the tension portion of the strain-cycle. Hold times in compression were much less detrimental than hold times in tension.

Boron and Zirconium from Crucible Refractories in a Complex Heat-Resistant Alloy

NASA Technical Reports Server (NTRS)

Decker, R F; Rowe, John P; Freeman, J W

1958-01-01

In a laboratory study of the factors involved in the influence of induction vacuum melting on 55ni-20cr-15co-4mo-3ti-3al heat resistant alloy, it was found that the major factor was the type of ceramic used as the crucible. The study concluded that trace amounts of boron or zirconium derived from reaction of the melt with the crucible refactories improved creep-rupture properties at 1,600 degrees F. Boron was most effective and, in addition, markedly improved hot-workability.

Continuum model for hydrogen pickup in zirconium alloys of LWR fuel cladding

NASA Astrophysics Data System (ADS)

Wang, Xing; Zheng, Ming-Jie; Szlufarska, Izabela; Morgan, Dane

2017-04-01

A continuum model for calculating the time-dependent hydrogen pickup fractions in various Zirconium alloys under steam and pressured water oxidation has been developed in this study. Using only one fitting parameter, the effective hydrogen gas partial pressure at the oxide surface, a qualitative agreement is obtained between the predicted and previously measured hydrogen pickup fractions. The calculation results therefore demonstrate that H diffusion through the dense oxide layer plays an important role in the hydrogen pickup process. The limitations and possible improvement of the model are also discussed.

URANIUM ALLOYS

DOEpatents

Seybolt, A.U.

1958-04-15

Uranium alloys containing from 0.1 to 10% by weight, but preferably at least 5%, of either zirconium, niobium, or molybdenum exhibit highly desirable nuclear and structural properties which may be improved by heating the alloy to about 900 d C for an extended period of time and then rapidly quenching it.

Oxidised zirconium versus cobalt alloy bearing surfaces in total knee arthroplasty: 3D laser scanning of retrieved polyethylene inserts.

PubMed

Anderson, F L; Koch, C N; Elpers, M E; Wright, T M; Haas, S B; Heyse, T J

2017-06-01

We sought to establish whether an oxidised zirconium (OxZr) femoral component causes less loss of polyethylene volume than a cobalt alloy (CoCr) femoral component in total knee arthroplasty. A total of 20 retrieved tibial inserts that had articulated with OxZr components were matched with 20 inserts from CoCr articulations for patient age, body mass index, length of implantation, and revision diagnosis. Changes in dimensions of the articular surfaces were compared with those of pristine inserts using laser scanning. The differences in volume between the retrieved and pristine surfaces of the two groups were calculated and compared. The loss of polyethylene volume was 122 mm 3 (standard deviation (sd) 87) in the OxZr group and 170 mm 3 (sd 96) in the CoCr group (p = 0.033). The volume loss in the OxZr group was also lower in the medial (72 mm 3 (sd 67) versus 92 mm 3 (sd 60); p = 0.096) and lateral (49 mm 3 (sd 36) versus 79 mm 3 (sd 61); p = 0.096) compartments separately, but these differences were not significant. Our results corroborate earlier findings from in vitro testing and visual retrieval analysis which suggest that polyethylene volume loss is lower with OxZr femoral components. Since both OxZr and CoCr are hard surfaces that would be expected to create comparable amounts of polyethylene creep, the differences in volume loss may reflect differences in the in vivo wear of these inserts. Cite this article: Bone Joint J 2017;99-B:793-8. ©2017 The British Editorial Society of Bone & Joint Surgery.

Amorphous metal alloy

DOEpatents

Wang, R.; Merz, M.D.

1980-04-09

Amorphous metal alloys of the iron-chromium and nickel-chromium type have excellent corrosion resistance and high temperature stability and are suitable for use as a protective coating on less corrosion resistant substrates. The alloys are stabilized in the amorphous state by one or more elements of titanium, zirconium, hafnium, niobium, tantalum, molybdenum, and tungsten. The alloy is preferably prepared by sputter deposition.

TEM study on the initial oxidation of Zircaloy-4 thin foil specimens heated in a low vacuum air condition at 280-300 °C

NASA Astrophysics Data System (ADS)

Wang, Zhen; Zhou, Bang-xin; Zhu, Wei; Wen, Bang; Yao, Mei-yi; Li, Qiang; Wu, Lu; Zhang, Jin-long; Fang, Zhong-qiang

2017-04-01

As one of the important structural materials in nuclear industry, the corrosion resistance of zirconium alloy limits their in-pile application. Therefore, it is necessary to investigate the corrosion mechanism of zirconium alloys. The zirconium-oxygen reaction at the O/M interface is one of the factors that affect the oxidation process. There are few reports in this regard. Ideally, the reaction process at the O/M interface has certain relevance with the initiation oxidation of zirconium, which provided a new way to investigate the reaction process by observing the initiation oxidation behaviours. To investigate the oxidation behaviours of zirconium alloy at the initial stage, in this paper, zircaloy-4 TEM thin foil specimens in 3 mm diameter were studied by TEM observation after heating in air condition with a vacuum of 3 Pa at 280 °C, 290 °C and 300 °C for 30 min exposures. The results show that, ZrO2 begin to nucleate at a size of 3-5 nm at a high Zr/O ratio of 10.4 and oxide layer formed while Zr/O was 4.6. As a result of stress caused by the P.B ratio of Zr, slip bands formed and a bcc structure sub-oxide b-ZrOx (a = 0.51 nm) grew up along with the slip bands was observed. At both sides of b-ZrOx, two hcp structure sub-oxides having the same a-axis lattice parameter and different c-axis lattice parameter were detected.

DISSOLUTION OF ZIRCONIUM AND ALLOYS THEREFOR

DOEpatents

Swanson, J.L.

1961-07-11

The dissolution of zirconium cladding in a water solution of ammonium fluoride and ammonium nitrate is described. The method finds particular utility in processing spent fuel elements for nuclear reactors. The zirconium cladding is first dissolved in a water solution of ammonium fluoride and ammonium nitrate; insoluble uranium and plutonium fiuorides formed by attack of the solvent on the fuel materiai of the fuel element are then separated from the solution, and the fuel materiai is dissolved in another solution.

Numerical assessment of bone remodeling around conventionally and early loaded titanium and titanium-zirconium alloy dental implants.

PubMed

Akça, Kıvanç; Eser, Atılım; Çavuşoğlu, Yeliz; Sağırkaya, Elçin; Çehreli, Murat Cavit

2015-05-01

The aim of this study was to investigate conventionally and early loaded titanium and titanium-zirconium alloy implants by three-dimensional finite element stress analysis. Three-dimensional model of a dental implant was created and a thread area was established as a region of interest in trabecular bone to study a localized part of the global model with a refined mesh. The peri-implant tissues around conventionally loaded (model 1) and early loaded (model 2) implants were implemented and were used to explore principal stresses, displacement values, and equivalent strains in the peri-implant region of titanium and titanium-zirconium implants under static load of 300 N with or without 30° inclination applied on top of the abutment surface. Under axial loading, principal stresses in both models were comparable for both implants and models. Under oblique loading, principal stresses around titanium-zirconium implants were slightly higher in both models. Comparable stress magnitudes were observed in both models. The displacement values and equivalent strain amplitudes around both implants and models were similar. Peri-implant bone around titanium and titanium-zirconium implants experiences similar stress magnitudes coupled with intraosseous implant displacement values under conventional loading and early loading simulations. Titanium-zirconium implants have biomechanical outcome comparable to conventional titanium implants under conventional loading and early loading.

Numerical Simulations on the Laser Spot Welding of Zirconium Alloy Endplate for Nuclear Fuel Bundle Assembly

NASA Astrophysics Data System (ADS)

Satyanarayana, G.; Narayana, K. L.; Boggarapu, Nageswara Rao

2018-03-01

In the nuclear industry, a critical welding process is joining of an end plate to a fuel rod to form a fuel bundle. Literature on zirconium welding in such a critical operation is limited. A CFD model is developed and performed for the three-dimensional non-linear thermo-fluid analysis incorporating buoyancy and Marnangoni stress and specifying temperature dependent properties to predict weld geometry and temperature field in and around the melt pool of laser spot during welding of a zirconium alloy E110 endplate with a fuel rod. Using this method, it is possible to estimate the weld pool dimensions for the specified laser power and laser-on-time. The temperature profiles will estimate the HAZ and microstructure. The adequacy of generic nature of the model is validated with existing experimental data.

Iron-chrome-aluminum alloy cladding for increasing safety in nuclear power plants

NASA Astrophysics Data System (ADS)

Rebak, Raul B.

2017-12-01

After a tsunami caused plant black out at Fukushima, followed by hydrogen explosions, the US Department of Energy partnered with fuel vendors to study safer alternatives to the current UO2-zirconium alloy system. This accident tolerant fuel alternative should better tolerate loss of cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. General electric, Oak ridge national laboratory, and their partners are proposing to replace zirconium alloy cladding in current commercial light water power reactors with an iron-chromium-aluminum (FeCrAl) cladding such as APMT or C26M. Extensive testing and evaluation is being conducted to determine the suitability of FeCrAl under normal operation conditions and under severe accident conditions. Results show that FeCrAl has excellent corrosion resistance under normal operation conditions and FeCrAl is several orders of magnitude more resistant than zirconium alloys to degradation by superheated steam under accident conditions, generating less heat of oxidation and lower amount of combustible hydrogen gas. Higher neutron absorption and tritium release effects can be minimized by design changes. The implementation of FeCrAl cladding is a near term solution to enhance the safety of the current fleet of commercial light water power reactors.

Amorphous metal alloy and composite

DOEpatents

Wang, Rong; Merz, Martin D.

1985-01-01

Amorphous metal alloys of the iron-chromium and nickel-chromium type have excellent corrosion resistance and high temperature stability and are suitable for use as a protective coating on less corrosion resistant substrates. The alloys are stabilized in the amorphous state by one or more elements of titanium, zirconium, hafnium, niobium, tantalum, molybdenum, and tungsten. The alloy is preferably prepared by sputter deposition.

Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors

DOE PAGES

Yamamoto, Yukinori; Pint, Bruce A.; Terrani, Kurt A.; ...

2015-10-19

Development of nuclear grade, iron-based wrought FeCrAl alloys has been initiated for light water reactor (LWR) fuel cladding to serve as a substitute for zirconium-based alloys with enhanced accident tolerance. Ferritic alloys with sufficient chromium and aluminum additions can exhibit significantly improved oxidation kinetics in high-temperature steam environments when compared to zirconium-based alloys. In the first phase, a set of model FeCrAl alloys containing 10–20Cr, 3–5Al, and 0–0.12Y in weight percent, were prepared by conventional arc-melting and hot-working processes to explore the effect of composition on the properties of FeCrAlY alloys. It was found that the tensile properties were insensitivemore » to the alloy compositions studied; however, the steam oxidation resistance strongly depended on both the chromium and the aluminum contents. The second phase development focused on strengthening Fe-13Cr-5Al with minor alloying additions of molybdenum, niobium, and silicon. Combined with an optimized thermo-mechanical treatment, a thermally stable microstructure was produced with improved tensile properties at temperatures up to 741°C.« less

Primary radiation damage of Zr-0.5%Nb binary alloy: atomistic simulation by molecular dynamics method

NASA Astrophysics Data System (ADS)

Tikhonchev, M.; Svetukhin, V.; Kapustin, P.

2017-09-01

Ab initio calculations predict high positive binding energy (˜1 eV) between niobium atoms and self-interstitial configurations in hcp zirconium. It allows the expectation of increased niobium fraction in self-interstitials formed under neutron irradiation in atomic displacement cascades. In this paper, we report the results of molecular dynamics simulation of atomic displacement cascades in Zr-0.5%Nb binary alloy and pure Zr at the temperature of 300 K. Two sets of n-body interatomic potentials have been used for the Zr-Nb system. We consider a cascade energy range of 2-20 keV. Calculations show close estimations of the average number of produced Frenkel pairs in the alloy and pure Zr. A high fraction of Nb is observed in the self-interstitial configurations. Nb is mainly detected in single self-interstitial configurations, where its fraction reaches tens of percent, i.e. more than its tenfold concentration in the matrix. The basic mechanism of this phenomenon is the trapping of mobile self-interstitial configurations by niobium. The diffusion of pure zirconium and mixed zirconium-niobium self-interstitial configurations in the zirconium matrix at 300 K has been simulated. We observe a strong dependence of the estimated diffusion coefficients and fractions of Nb in self-interstitials produced in displacement cascades on the potential.

Observations on the brittle to ductile transition temperatures of B2 nickel aluminides with and without zirconium

NASA Technical Reports Server (NTRS)

Raj, S. V.; Noebe, R. D.; Bowman, R.

1989-01-01

The effect of a zirconium addition (0.05 at. pct) to a stoichiometric NiAl alloy on the brittle-to-ductile transition temperature (BDTT) of this alloy was investigated. Constant velocity tensile tests were conducted to fracture between 300 and 1100 K under initial strain rate 0.00014/sec, and the true stress and true strain values were determined from plots of load vs time after subtracting the elastic strain. The inelastic strain was measured under a traveling microscope. Microstructural characterization of as-extruded and fractured specimens was carried out by SEM and TEM. It was found that, while the addition of 0.05 at. pct Zr strengthened the NiAl alloy, it increased its BDTT; this shift in the BDTT could not be attributed either to variations in grain size or to impurity contents. Little or no room-temperature ductility was observed for either alloy.

Rapid Quench in an Electrostatic Levitator

NASA Technical Reports Server (NTRS)

SanSoucie, Michael P.; Rogers, Jan R.; Matson, Douglas M.

2016-01-01

The Electrostatic Levitation (ESL) Laboratory at the NASA Marshall Space Flight Center (MSFC) is a unique facility for investigators studying high-temperature materials. The ESL laboratory's main chamber has been upgraded with the addition of a rapid quench system. This system allows samples to be dropped into a quench vessel that can be filled with a low melting point material, such as a gallium or indium alloy, as a quench medium. Thereby allowing rapid quenching of undercooled liquid metals. Up to eight quench vessels can be loaded into a wheel inside the chamber that is indexed with control software. The system has been tested successfully with samples of zirconium, iron-cobalt alloys, titanium-zirconium-nickel alloys, and a silicon-cobalt alloy. This new rapid quench system will allow materials science studies of undercooled materials and new materials development. In this presentation, the system is described and some initial results are presented.

Rapid Quench in an Electrostatic Levitator

NASA Technical Reports Server (NTRS)

SanSoucie, Michael P.; Rogers, Jan R.; Matson, Michael M.

2016-01-01

The Electrostatic Levitation (ESL) Laboratory at the NASA Marshall Space Flight Center (MSFC) is a unique facility for investigators studying high-temperature materials. The ESL laboratory’s main chamber has been upgraded with the addition of a rapid quench system. This system allows samples to be dropped into a quench vessel that can be filled with a low melting point material, such as a gallium or indium alloy, as a quench medium. Thereby allowing rapid quenching of undercooled liquid metals. Up to eight quench vessels can be loaded into a wheel inside the chamber that is indexed with control software. The system has been tested successfully with samples of zirconium, iron-cobalt alloys, iron-chromium-nickel, titanium-zirconium-nickel alloys, and a silicon-cobalt alloy. This new rapid quench system will allow materials science studies of undercooled materials and new materials development. The system is described and some initial results are presented.

THE DETERMINATION OF TRACES OF BORON IN ZIRCONIUM METAL AND ZIRCONIUM ALLOYS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hayes, M.R.; Metcalfe, J.

1962-01-01

A general procedure is given for the determination of B, down to 0.2 ppm, in Zr and Zr alloys. Separation of the B is not necessary, the B-curcumin complex being formed directly in an aliquot of the metal sulfate solution. An interference effect has been noted when analyzing Zr alloys containing Sn. The interference is caused by an insoluble compound of curcumin which separates and has similar properties to the B-curcumin complex. This source of interference is, however, readily eliminated during the procedure for the determination of B. The procedure has been applied to the determination of B in puremore » Zr, zr--0.5% Cu-- 0.5% MO, and Zr--1.5% Sn--0.1% Fe--0.1% Cr--0.05% Ni alloys. Results are comparable with those obtained by methods requiring the separation of the B as methyl borate. (auth)« less

Data on the effect of homogenization heat treatments on the cast structure and tensile properties of alloy 718Plus in the presence of grain-boundary elements.

PubMed

Hosseini, Seyed Ali; Madar, Karim Zangeneh; Abbasi, Seyed Mehdi

2017-08-01

The segregation of the elements during solidification and the direct formation of destructive phases such as Laves from the liquid, result in in-homogeneity of the cast structure and degradation of mechanical properties. Homogenization heat treatment is one of the ways to eliminate destructive Laves from the cast structure of superalloys such as 718Plus. The collected data presents the effect of homogenization treatment conditions on the cast structure, hardness, and tensile properties of the alloy 718Plus in the presence of boron and zirconium additives. For this purpose, five alloys with different contents of boron and zirconium were cast by VIM/VAR process and then were homogenized at various conditions. The microstructural investigation by OM and SEM and phase analysis by XRD were done and then hardness and tensile tests were performed on the homogenized alloys.

Influence of Crucible Materials on High-temperature Properties of Vacuum-melted Nickel-chromium-cobalt Alloy

NASA Technical Reports Server (NTRS)

Decker, R F; Rowe, John P; Freeman, J W

1957-01-01

A study of the effect of induction-vacuum-melting procedure on the high-temperature properties of a titanium-and-aluminum-hardened nickel-base alloy revealed that a major variable was the type of ceramic used as a crucible. Reactions between the melt and magnesia or zirconia crucibles apparently increased high-temperature properties by introducing small amounts of boron or zirconium into the melts. Heats melted in alumina crucibles had relatively low rupture life and ductility at 1,600 F and cracked during hot-working as a result of deriving no boron or zirconium from the crucible.

Effects of alloying elements on the formation of < c >-component loops in Zr alloy Excel under heavy ion irradiation.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Idrees, Yasir; Francis, Elisabeth M.; Yao, Zhongwen

2015-05-14

We report here the microstructural changes occurring in the zirconium alloy Excel (Zr-3.5 wt% Sn-0.8Nb-0.8Mo-0.2Fe) during heavy ion irradiation. In situ irradiation experiments were conducted at reactor operating temperatures on two Zr Excel alloy microstructures with different states of alloying elements, with the states achieved by different solution heat treatments. In the first case, the alloying elements were mostly concentrated in the beta (beta) phase, whereas, in the second case, large Zr-3(Mo,Nb,Fe)(4) secondary phase precipitates (SPPs) were grown in the alpha (alpha) phase by long term aging. The heavy ion induced damage and resultant compositional changes were examined using transmissionmore » electron microscopy (TEM) in combination with scanning transmission electron microscope (STEM)-energy dispersive x-ray spectroscopy (EDS) mapping. Significant differences were seen in microstructural evolution between the two different microstructures that were irradiated under similar conditions. Nucleation and growth of < c >-component loops and their dependence on the alloying elements are a major focus of the current investigation. It was observed that the < c >-component loops nucleate readily at 100, 300, and 400 degrees C after a threshold incubation dose (TID), which varies with irradiation temperature and the state of alloying elements. It was found that the TID for the formation of < c >-component loops increases with decrease in irradiation temperature. Alloying elements that are present in the form of SPPs increase the TID compared to when they are in the beta phase solid solution. Dose and temperature dependence of loop size and density are presented. Radiation induced redistribution and clustering of alloying elements (Sn, Mo, and Fe) have been observed and related to the formation of < c >-component loops. It has been shown that at the higher temperature tests, irradiation induced dissolution of precipitates occurs whereas irradiation induced amorphization occurs at 100 degrees C. Furthermore, dose and temperature seem to be the main factors governing the dissolution of SPPs and redistribution of alloying elements, which in turn controls the nucleation and growth of < c >-component loops. The correlation between the microstructural evolution and microchemistry has been found by EDS and is discussed in detail.« less

Nuclear fuel element

DOEpatents

Armijo, Joseph S.; Coffin, Jr., Louis F.

1980-04-29

A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has an improved composite cladding comprised of a moderate purity metal barrier of zirconium metallurgically bonded on the inside surface of a zirconium alloy tube. The metal barrier forms a shield between the alloy tube and a core of nuclear fuel material enclosed in the composite cladding. There is a gap between the cladding and the core. The metal barrier forms about 1 to about 30 percent of the thickness of the composite cladding and has low neutron absorption characteristics. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the alloy tube from contact and reaction with such impurities and fission products. Methods of manufacturing the composite cladding are also disclosed.

Effect of oxidation on transport properties of zirconium-1% niobium alloy

NASA Astrophysics Data System (ADS)

Peletsky, V. E.; Musayeva, Z. A.

1995-11-01

The thermal conductivity and electrical resistivity of zirconium-1 wt% niobium samples were measured before and after the process of their oxidation in air. A special procedure was used to dissolve the gas and to smooth out its concentration in the alloy. The basic experiments were performed under high vacuum under steady-state temperature conditions. The temperature range was 300 1600 K. for the pure alloy and 300 1100 K for the samples containing oxygen. It was found that the thermal conductivity—oxygen concentration relation reverses its sign from negative at low and middle temperatures to positive at temperatures above 900 K. The relation between the electrical resistivity and the oxygen content does not show this feature. The Lorenz function was found to have an anomalous temperature dependence.

Method for fabricating uranium alloy articles without shape memory effects

DOEpatents

Banker, John G.

1985-01-01

Uranium-rich niobium and niobium-zirconium alloys possess a characteristic known as shape memory effect wherein shaped articles of these alloys recover their original shape when heated. The present invention circumvents this memory behavior by forming the alloys into the desired configuration at elevated temperatures with "cold" matched dies and maintaining the shaped articles between the dies until the articles cool to ambient temperature.

Method for fabricating uranium alloy articles without shape memory effects

DOEpatents

Banker, J.G.

1980-05-21

Uranium-rich niobium and niobium-zirconium alloys possess a characteristic known as shape memory effect wherein shaped articles of these alloys recover their original shape when heated. The present invention circumvents this memory behavior by forming the alloys into the desired configuration at elevated temperatures with cold matched dies and maintaining the shaped articles between the dies until the articles cool to ambient temperature.

As-cast uranium-molybdenum based metallic fuel candidates and the effects of carbon addition

NASA Astrophysics Data System (ADS)

Blackwood, Van Stephen

The objective of this research was to develop and recommend a metallic nuclear fuel candidate that lowered the onset temperature of gamma phase formation comparable or better than the uranium-10 wt. pct. molybdenum alloy, offered a solidus temperature as high or higher than uranium-10 wt. pct. zirconium (1250°C), and stabilized the fuel phase against interaction with iron and steel at least as much as uranium-10 wt. pct. zirconium stabilized the fuel phase. Two new as-cast alloy compositions were characterized to assess thermal equilibrium boundaries of the gamma phase field and the effect of carbon addition up to 0.22 wt. pct. The first system investigated was uranium- x wt. pct. M where x ranged between 5-20 wt. pct. M was held at a constant ratio of 50 wt. pct. molybdenum, 43 wt. pct. titanium, and 7 wt. pct. zirconium. The second system investigated was the uranium-molybdenum-tungsten system in the range 90 wt. pct. uranium - 10 wt. pct. molybdenum - 0 wt. pct. tungsten to 80 wt. pct. uranium - 10 wt. pct. molybdenum - 10 wt. pct. tungsten. The results showed that the solidus temperature increased with increased addition of M up to 12.5 wt. pct. for the uranium-M system. Alloy additions of titanium and zirconium were removed from uranium-molybdenum solid solution by carbide formation and segregation. The uranium-molybdenum-tungsten system solidus temperature increased to 1218°C at 2.5 wt. pct. with no significant change in temperature up to 5 wt. pct. tungsten suggesting the solubility limit of tungsten had been reached. Carbides were observed with surrounding areas enriched in both molybdenum and tungsten. The peak solidus temperatures for the alloy systems were roughly the same at 1226°C for the uranium-M system and 1218°C for the uranium-molybdenum-tungsten system. The uranium-molybdenum-tungsten system required less alloy addition to achieve similar solidus temperatures as the uranium-M system.

Microstructure and mechanical properties of zirconium doped NiAl/Cr(Mo) hypoeutectic alloy prepared by injection casting

NASA Astrophysics Data System (ADS)

Sheng, L. Y.; Du, B. N.; Guo, J. T.

2017-01-01

NiAl based materials has been considered as most potential candidate of turbine blade, due to its excellent high-temperature properties. However the bad room-temperature properties handicap its application. In the present paper, the zirconium doped NiAl/Cr(Mo) hypoeutectic alloy is fabricated by conventional casting and injection casting technology to improve its room-temperature properties. The microstructure and compressive properties at different temperatures of the conventionally-cast and injection-cast were investigated. The results exhibit that the conventionally-cast alloy comprises coarse primary NiAl phase and eutectic cell, which is dotted with irregular Ni2AlZr Heusler phase. Compared with the conventionally-cast alloy, the injection-cast alloy possesses refined the primary NiAl, eutectic cell and eutectic lamella. In addition, the Ni2AlZr Heusler phase become smaller and distribute uniformly. Moreover, the injection casting decrease the area fraction of primary NiAl phase at the cell interior or cell boundaries. The compressive ductility and yield strength of the injection-cast alloy at room temperature increase by about 100% and 35% over those of conventionally-cast alloy, which should be ascribed to the microstructure optimization.

TERNARY ALLOYS OF URANIUM, COLUMBIUM, AND ZIRCONIUM

DOEpatents

Foote, F.G.

1960-08-01

Ternary alloys of uranium are described which are useful as neutron- reflecting materials in a fast neutron reactor. They are especially resistant to corrosion caused by oxidative processes of gascous or aqueous origin and comprise uranium as the predominant metal with zirconiunn and niobium wherein the total content of the minor alloying elements is between 2 and 8% by weight.

Superalloy material with improved weldability

DOEpatents

Allen, David B.; Wagner, Gregg P.; Seth, Brij B.

2004-02-24

A fusion weldable superalloy containing 0.005-0.5 wt. % scandium. In one embodiment, the superalloy may have a composition similar to IN-939 alloy, but having added scandium and having only 0.005-0.040 wt. % zirconium. A gas turbine component may be formed by an investment casting of such a scandium-containing superalloy, and may include a fusion weld repaired area. A scandium-containing nickel-based superalloy coated with an MCrAlY bond coat will have improved cyclic oxidation resistance due to the sulfur-gettering effect of the scandium.

Interfacing VPSC with finite element codes. Demonstration of irradiation growth simulation in a cladding tube

DOE Office of Scientific and Technical Information (OSTI.GOV)

Patra, Anirban; Tome, Carlos

This Milestone report shows good progress in interfacing VPSC with the FE codes ABAQUS and MOOSE, to perform component-level simulations of irradiation-induced deformation in Zirconium alloys. In this preliminary application, we have performed an irradiation growth simulation in the quarter geometry of a cladding tube. We have benchmarked VPSC-ABAQUS and VPSC-MOOSE predictions with VPSC-SA predictions to verify the accuracy of the VPSCFE interface. Predictions from the FE simulations are in general agreement with VPSC-SA simulations and also with experimental trends.

Evaluation of a Zirconium Recycle Scrubber System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Spencer, Barry B.; Bruffey, Stephanie H.

2017-04-01

A hot-cell demonstration of the zirconium recycle process is planned as part of the Materials Recovery and Waste Forms Development (MRWFD) campaign. The process treats Zircaloy® cladding recovered from used nuclear fuel with chlorine gas to recover the zirconium as volatile ZrCl4. This releases radioactive tritium trapped in the alloy, converting it to volatile tritium chloride (TCl). To meet regulatory requirements governing radioactive emissions from nuclear fuel treatment operations, the capture and retention of a portion of this TCl may be required prior to discharge of the off-gas stream to the environment. In addition to demonstrating tritium removal from amore » synthetic zirconium recycle off-gas stream, the recovery and quantification of tritium may refine estimates of the amount of tritium present in the Zircaloy cladding of used nuclear fuel. To support these objectives, a bubbler-type scrubber was fabricated to remove the TCl from the zirconium recycle off-gas stream. The scrubber was fabricated from glass and polymer components that are resistant to chlorine and hydrochloric acid solutions. Because of concerns that the scrubber efficiency is not quantitative, tests were performed using DCl as a stand-in to experimentally measure the scrubbing efficiency of this unit. Scrubbing efficiency was ~108% ± 3% with water as the scrubber solution. Variations were noted when 1 M NaOH scrub solution was used, values ranged from 64% to 130%. The reason for the variations is not known. It is recommended that the equipment be operated with water as the scrubbing solution. Scrubbing efficiency is estimated at 100%.« less

Recycle of Zirconium from Used Nuclear Fuel Cladding: A Major Element of Waste Reduction

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collins, Emory D; DelCul, Guillermo D; Terekhov, Dmitri

2011-01-01

Feasibility tests were initiated to determine if the zirconium in commercial used nuclear fuel (UNF) cladding can be recovered in sufficient purity to permit re-use, and if the recovery process can be operated economically. Initial tests are being performed with unirradiated, non-radioactive samples of various types of Zircaloy materials that are used in UNF cladding to develop the recovery process and determine the degree of purification that can be obtained. Early results indicate that quantitative recovery can be accomplished and product contamination with alloy constituents can be controlled sufficiently to meet purification requirements. Future tests with actual radioactive UNF claddingmore » are planned. The objective of current research is to determine the feasibility of recovery and recycle of zirconium from used fuel cladding wastes. Zircaloy cladding, which contains 98+% of hafnium-free zirconium, is the second largest mass, on average {approx}25 wt %, of the components in used U.S. light-water-reactor fuel assemblies. Therefore, recovery and recycle of the zirconium would enable a large reduction in geologic waste disposal for advanced fuel cycles. Current practice is to compact or grout the cladding waste and store it for subsequent disposal in a geologic repository. This paper describes results of initial tests being performed with unirradiated, non-radioactive samples of various types of Zircaloy materials that are used in UNF cladding to develop the recovery process and determine the degree of purification that can be obtained. Future tests with actual radioactive UNF cladding are planned.« less

Feasibility of processing the experimental breeder reactor-II driver fuel from the Idaho National Laboratory through Savannah River Site's H-Canyon facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Magoulas, V. E.

Savannah River National Laboratory (SRNL) was requested to evaluate the potential to receive and process the Idaho National Laboratory (INL) uranium (U) recovered from the Experimental Breeder Reactor II (EBR-II) driver fuel through the Savannah River Site’s (SRS) H-Canyon as a way to disposition the material. INL recovers the uranium from the sodium bonded metallic fuel irradiated in the EBR-II reactor using an electrorefining process. There were two compositions of EBR-II driver fuel. The early generation fuel was U-5Fs, which consisted of 95% U metal alloyed with 5% noble metal elements “fissium” (2.5% molybdenum, 2.0% ruthenium, 0.3% rhodium, 0.1% palladium,more » and 0.1% zirconium), while the later generation was U-10Zr which was 90% U metal alloyed with 10% zirconium. A potential concern during the H-Canyon nitric acid dissolution process of the U metal containing zirconium (Zr) is the explosive behavior that has been reported for alloys of these materials. For this reason, this evaluation was focused on the ability to process the lower Zr content materials, the U-5Fs material.« less

Zirconium, calcium, and strontium contents in magnesium based biodegradable alloys modulate the efficiency of implant-induced osseointegration

PubMed Central

Mushahary, Dolly; Sravanthi, Ragamouni; Li, Yuncang; Kumar, Mahesh J; Harishankar, Nemani; Hodgson, Peter D; Wen, Cuie; Pande, Gopal

2013-01-01

Development of new biodegradable implants and devices is necessary to meet the increasing needs of regenerative orthopedic procedures. An important consideration while formulating new implant materials is that they should physicochemically and biologically mimic bone-like properties. In earlier studies, we have developed and characterized magnesium based biodegradable alloys, in particular magnesium-zirconium (Mg-Zr) alloys. Here we have reported the biological properties of four Mg-Zr alloys containing different quantities of strontium or calcium. The alloys were implanted in small cavities made in femur bones of New Zealand White rabbits, and the quantitative and qualitative assessments of newly induced bone tissue were carried out. A total of 30 experimental animals, three for each implant type, were studied, and bone induction was assessed by histological, immunohistochemical and radiological methods; cavities in the femurs with no implants and observed for the same period of time were kept as controls. Our results showed that Mg-Zr alloys containing appropriate quantities of strontium were more efficient in inducing good quality mineralized bone than other alloys. Our results have been discussed in the context of physicochemical and biological properties of the alloys, and they could be very useful in determining the nature of future generations of biodegradable orthopedic implants. PMID:23976848

Process for electroless deposition of metals on zirconium materials

DOEpatents

Donaghy, Robert E.

1978-01-01

A process for the electroless 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 electroless plating solution containing the metal to be deposited on the article upon sufficient contact with the article.

Process for electrolytic deposition of metals on zirconium materials

DOEpatents

Donaghy, Robert E.

1979-01-30

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.

Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Kurt R.; Howard, Richard H.; Daily, Charles R.

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designsmore » allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.« less

High temperature coatings for gas turbines

DOEpatents

Zheng, Xiaoci Maggie

2003-10-21

Coating for high temperature gas turbine components that include a MCrAlX phase, and an aluminum-rich phase, significantly increase oxidation and cracking resistance of the components, thereby increasing their useful life and reducing operating costs. The aluminum-rich phase includes aluminum at a higher concentration than aluminum concentration in the MCrAlX alloy, and an aluminum diffusion-retarding composition, which may include cobalt, nickel, yttrium, zirconium, niobium, molybdenum, rhodium, cadmium, indium, cerium, iron, chromium, tantalum, silicon, boron, carbon, titanium, tungsten, rhenium, platinum, and combinations thereof, and particularly nickel and/or rhenium. The aluminum-rich phase may be derived from a particulate aluminum composite that has a core comprising aluminum and a shell comprising the aluminum diffusion-retarding composition.

Advanced Steels for Accident Tolerant Fuel Cladding in Current Light Water Reactors

NASA Astrophysics Data System (ADS)

Rebak, Raul B.

After the March 2011 Fukushima events, the U.S. Congress directed the Department of Energy (DOE) to focus efforts on the development of fuel cladding materials with enhanced accident tolerance. In comparison with the stand-ard UO2-Zirconium based system, the new fuels need to tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operation conditions. Advanced steels such as iron-chromium-aluminum (FeCrAl) alloys are being investigated for degradation behavior both under normal operation conditions in high temperature water (e.g. 288°C) and under accident conditions for reaction with steam up to 1400°C. Commercial and experimental alloys were tested for several periods of time in 100% superheated steam from 800°C to 1475°C. Results show that FeCrAl alloys significantly outperform the resistance in steam of the current zirconium alloys.

Refractory metal joining for first wall applications

NASA Astrophysics Data System (ADS)

Cadden, C. H.; Odegard, B. C.

2000-12-01

The potential use of high temperature coolant (e.g. 900°C He) in first wall structures would preclude the applicability of copper alloy heat sink materials and refractory metals would be potential replacements. Brazing trials were conducted in order to examine techniques to join tungsten armor to high tungsten (90-95 wt%) or molybdenum TZM heat sink materials. Palladium-, nickel- and zirconium-based filler metals were investigated using brazing temperatures ranging from 1000°C to 1275°C. Palladium-nickel and palladium-cobalt braze alloys were successful in producing generally sound metallurgical joints in tungsten alloy/tungsten couples, although there was an observed tendency for the pure tungsten armor material to exhibit grain boundary cracking after bonding. The zirconium- and nickel-based filler metals produced defect-containing joints, specifically cracking and porosity, respectively. The palladium-nickel braze alloy produced sound joints in the Mo TZM/tungsten couple. Substitution of a lanthanum oxide-containing, fine-grained tungsten material (for the pure tungsten) eliminated the observed tungsten grain boundary cracking.

DIMENSIONALLY STABLE, CORROSION RESISTANT NUCLEAR FUEL

DOEpatents

Kittel, J.H.

1963-10-31

A method of making a uranium alloy of improved corrosion resistance and dimensional stability is described. The alloy contains from 0-9 weight per cent of an additive of zirconium and niobium in the proportions by weight of 5 to 1 1/ 2. The alloy is cold rolled, heated to two different temperatures, air-cooled, heated to a third temperature, and quenched in water. (AEC)

Preparation and properties of chrome-free colored Ti/Zr based conversion coating on aluminum alloy

NASA Astrophysics Data System (ADS)

Yi, AiHua; Li, WenFang; Du, Jun; Mu, SongLin

2012-06-01

A golden conversion coating on the surface of aluminum alloy was prepared by adding tannic acid and coating-forming accelerator in the treatment solution containing titanium and zirconium ions. The growth process, main component and corrosion resistance of the conversion coating were characterized by EDS, SEM, XRD, XPS, FIIR and electrochemical workstation. The results showed that the main components of the conversion coating were Na3AlF6 and the conversion coating owns a double-layer structure. The outer layer consists of metal-organic complex and the inner layer is mainly made up of Na3AlF6. The mechanism of the formation of the golden conversion coating can be deemed as nucleation, growth of Na3AlF6 crystal and formation of metal-organic complex. In potentiodynamic polarization test, the corrosion current density decreases to 0.283 μA cm-2 from 5.894 μA cm-2, which indicates an obvious improvement of corrosion resistance.

Thermomechanical processing of aluminum micro-alloyed with Sc, Zr, Ti, B, and C

NASA Astrophysics Data System (ADS)

McNamara, Cameron T.

Critical exploration of the minimalistic high strength low alloy aluminum (HSLA-Al) paradigm is necessary for the continued development of advanced aluminum alloys. In this study, scandium (Sc) and zirconium (Zr) are examined as the main precipitation strengthening additions, while magnesium (Mg) is added to probe the synergistic effects of solution and precipitation hardening, as well as the grain refinement during solidification afforded by a moderate growth restriction factor. Further, pathways of recrystallization are explored in several potential HSLA-Al syste =ms sans Sc. Aluminum-titanium-boron (Al-Ti-B) and aluminum-titanium-carbon (Al-Ti-C) grain refining master alloys are added to a series of Al-Zr alloys to examine both the reported Zr poisoning effect on grain size reduction and the impact on recrystallization resistance through the use of electron backscattered diffraction (EBSD) imaging. Results include an analysis of active strengthening mechanisms and advisement for both constitution and thermomechanical processing of HSLA-Al alloys for wrought or near-net shape cast components. The mechanisms of recrystallization are discussed for alloys which contain a bimodal distribution of particles, some of which act as nucleation sites for grain formation during annealing and others which restrict the growth of the newly formed grains.

Hardness behavior of binary and ternary niobium alloys at 77 and 300 K

NASA Technical Reports Server (NTRS)

Stephens, J. R.; Witzke, W. R.

1974-01-01

The effects of alloy additions of zirconium, hafnium, molybdenum, tungsten, rhenium, ruthenium, osmium, rhodium, and iridium on the hardness of niobium was determined. Both binary and ternary alloys were investigated by means of hardness tests at 77 K and 300 K. Results showed that atomic size misfit plays a dominant role in controlling hardness of binary niobium alloys. Alloy softening, which occurred at dilute solute additions, is most likely due to an extrinsic mechanism involving interaction between solute elements and interstitial impurities.

Alloys for a liquid metal fast breeder reactor

DOEpatents

Rowcliffe, Arthur F.; Bleiberg, Melvin L.; Diamond, Sidney; Bajaj, Ram

1979-01-01

An essentially gamma-prime precipitation-hardened iron-chromium-nickel alloy has been designed with emphasis on minimum nickel and chromium contents to reduce the swelling tendencies of these alloys when used in liquid metal fast breeder reactors. The precipitation-hardening components have been designed for phase stability and such residual elements as silicon and boron, also have been selected to minimize swelling. Using the properties of these alloys in one design would result in an increased breeding ratio over 20% cold worked stainless steel, a reference material, of 1.239 to 1.310 and a reduced doubling time from 15.8 to 11.4 years. The gross stoichiometry of the alloying composition comprises from about 0.04% to about 0.06% carbon, from about 0.05% to about 1.0% silicon, up to about 0.1% zirconium, up to about 0.5% vanadium, from about 24% to about 31% nickel, from 8% to about 11% chromium, from about 1.7% to about 3.5% titanium, from about 1.0% to about 1.8% aluminum, from about 0.9% to about 3.7% molybdenum, from about 0.04% to about 0.8% boron, and the balance iron with incidental impurities.

40 CFR 421.336 - Pretreatment standards for new sources.

Code of Federal Regulations, 2013 CFR

2013-07-01

....000 499.500 (d) SiCl4 purification wet air pollution control. PSNS for the Primary Zirconium and....262 Lead 4.416 2.050 Nickel 8.674 5.835 Ammonia (as N) 2,102.000 895.800 (q) Leaching rinse water from... Nickel 32.410 21.810 Ammonia (as N) 7,856.000 3,453.000 (r) Leaching rinse water from zirconium alloy...

40 CFR 421.336 - Pretreatment standards for new sources.

Code of Federal Regulations, 2011 CFR

2011-07-01

....000 499.500 (d) SiCl4 purification wet air pollution control. PSNS for the Primary Zirconium and....262 Lead 4.416 2.050 Nickel 8.674 5.835 Ammonia (as N) 2,102.000 895.800 (q) Leaching rinse water from... Nickel 32.410 21.810 Ammonia (as N) 7,856.000 3,453.000 (r) Leaching rinse water from zirconium alloy...

40 CFR 421.336 - Pretreatment standards for new sources.

Code of Federal Regulations, 2010 CFR

2010-07-01

....000 499.500 (d) SiCl4 purification wet air pollution control. PSNS for the Primary Zirconium and....262 Lead 4.416 2.050 Nickel 8.674 5.835 Ammonia (as N) 2,102.000 895.800 (q) Leaching rinse water from... Nickel 32.410 21.810 Ammonia (as N) 7,856.000 3,453.000 (r) Leaching rinse water from zirconium alloy...

40 CFR 421.336 - Pretreatment standards for new sources.

Code of Federal Regulations, 2014 CFR

2014-07-01

....000 499.500 (d) SiCl4 purification wet air pollution control. PSNS for the Primary Zirconium and....262 Lead 4.416 2.050 Nickel 8.674 5.835 Ammonia (as N) 2,102.000 895.800 (q) Leaching rinse water from... Nickel 32.410 21.810 Ammonia (as N) 7,856.000 3,453.000 (r) Leaching rinse water from zirconium alloy...

40 CFR 421.336 - Pretreatment standards for new sources.

Code of Federal Regulations, 2012 CFR

2012-07-01

....000 499.500 (d) SiCl4 purification wet air pollution control. PSNS for the Primary Zirconium and....262 Lead 4.416 2.050 Nickel 8.674 5.835 Ammonia (as N) 2,102.000 895.800 (q) Leaching rinse water from... Nickel 32.410 21.810 Ammonia (as N) 7,856.000 3,453.000 (r) Leaching rinse water from zirconium alloy...

Remanufacture of Zirconium-Based Conversion Coatings on the Surface of Magnesium Alloy

NASA Astrophysics Data System (ADS)

Liu, Zhe; Jin, Guo; Song, Jiahui; Cui, Xiufang; Cai, Zhaobing

2017-04-01

Brush plating provides an effective method for creating a coating on substrates of various shapes. A corroded zirconium-based conversion coating was removed from the surface of a magnesium alloy and then replaced with new coatings prepared via brush plating. The structure and composition of the remanufactured coating were determined via x-ray photoelectron spectroscopy, x-ray diffraction, and Fourier transform infrared spectroscopy. The results revealed that the coatings consist of oxide, fluoride, and tannin-related organics. The composition of the coatings varied with the voltage. Furthermore, as revealed via potentiodynamic polarization spectroscopy, these coatings yielded a significant increase in the corrosion resistance of the magnesium alloy. The friction coefficient remained constant for almost 300s during wear resistance measurements performed under a 1-N load and dry sliding conditions, indicating that the remanufactured coatings provide effective inhibition to corrosion.

Castable hot corrosion resistant alloy

NASA Technical Reports Server (NTRS)

Barrett, Charles A. (Inventor); Holt, William H. (Inventor)

1988-01-01

Some 10 wt percent nickel is added to an Fe-base alloy which has a ferrite microstructure to improve the high temperature castability and crack resistance while about 0.2 wt percent zirconium is added for improved high temperatur cyclic oxidation and corrosion resistance. The basic material is a high temperature FeCrAl heater alloy, and the addition provides a material suitable for burner rig nozzles.

Hydrogenation behavior of Ti-implanted Zr-1Nb alloy with TiN films deposited using filtered vacuum arc and magnetron sputtering

NASA Astrophysics Data System (ADS)

Kashkarov, E. B.; Nikitenkov, N. N.; Sutygina, A. N.; Bezmaternykh, A. O.; Kudiiarov, V. N.; Syrtanov, M. S.; Pryamushko, T. S.

2018-02-01

More than 60 years of operation of water-cooled reactors have shown that local or general critical hydrogen concentration is one of the basic limiting criteria of zirconium-based fuel element claddings. During the coolant radiolysis, released hydrogen penetrates and accumulates in zirconium alloys. Hydrogenation of zirconium alloys leads to degradation of their mechanical properties, hydride cracking and stress corrosion cracking. In this research the effect of titanium nitride (TiN) deposition on hydrogenation behavior of Ti-implanted Zr-1Nb alloy was described. Ti-implanted interlayer was fabricated by plasma immersion ion implantation (PIII) at the pulsed bias voltage of 1500 V to improve the adhesion of TiN and reduce hydrogen penetration into Zr-1Nb alloy. We conducted the comparative analysis on hydrogenation behavior of the Ti-implanted alloy with sputtered and evaporated TiN films by reactive dc magnetron sputtering (dcMS) and filtered cathodic vacuum arc deposition (FVAD), respectively. The crystalline structure and surface morphology were investigated using X-ray diffraction (XRD) and scanning electron microscopy (SEM). The elemental distribution was analyzed using glow-discharge optical emission spectroscopy (GD-OES). Hydrogenation was performed from gas atmosphere at 350 °C and 2 atm hydrogen pressure. The results revealed that TiN films as well as Ti implantation significantly reduce hydrogen absorption rate of Zr-1Nb alloy. The best performance to reduce the rate of hydrogen absorption is Ti-implanted layer with evaporated TiN film. Morphology of the films impacted hydrogen permeation through TiN films: the denser film the lower hydrogen permeation. The Ti-implanted interface plays an important role of hydrogen accumulation layer for trapping the penetrated hydrogen. No deterioration of adhesive properties of TiN films on Zr-1Nb alloy with Ti-implanted interface occurs under high-temperature hydrogen exposure. Thus, the fabrication of Ti-implanted layer with dense TiN films can be an effective way to protect Zr-1Nb alloy from hydrogen embrittlement.

Structure, mechanical properties, and grindability of dental Ti-Zr alloys.

PubMed

Ho, Wen-Fu; Chen, Wei-Kai; Wu, Shih-Ching; Hsu, Hsueh-Chuan

2008-10-01

Structure, mechanical properties and grindability of a series of binary Ti-Zr alloys with zirconium contents ranging from 10 to 40 wt% have been investigated. Commercially pure titanium (c.p. Ti) was used as a control. Experimental results indicated that the diffraction peaks of all the Ti-Zr alloys matched those for alpha Ti. No beta-phase peaks were found. The hardness of the Ti-Zr alloys increased as the Zr contents increased, and ranged from 266 HV (Ti-10Zr) to 350 HV (Ti-40Zr). As the concentration of zirconium in the alloys increased, the strength, elastic recovery angles and hardness increased. Moreover, the elastically recoverable angle of Ti-40Zr was higher than of c.p. Ti by as much as 550%. The grindability of each metal was found to be largely dependent on the grinding conditions. The Ti-40Zr alloy had a higher grinding rate and grinding ratio than c.p. Ti at low speed. The grinding rate of the Ti-40Zr alloy at 500 m/min was about 1.8 times larger than that of c.p. Ti, and the grinding ratio was about 1.6 times larger than that of c.p. Ti. Our research suggested that the Ti-40Zr alloy has better mechanical properties, excellent elastic recovery capability and improved grindability at low grinding speed. The Ti-40Zr alloy has a great potential for use as a dental machining alloy.

Corrosion resistant iron aluminides exhibiting improved mechanical properties and corrosion resistance

DOEpatents

Liu, C.T.; McKamey, C.G.; Tortorelli, P.F.; David, S.A.

1994-06-14

The specification discloses a corrosion-resistant intermetallic alloy comprising, in atomic percent, an FeAl iron aluminide containing from about 30 to about 40% aluminum alloyed with from about 0.01 to 0.4% zirconium and from 0.01 to about 0.8% boron. The alloy exhibits considerably improved room temperature ductility for enhanced usefulness in structural applications. The high temperature strength and fabricability is improved by alloying with molybdenum, carbon, chromium and vanadium. 9 figs.

Corrosion resistant iron aluminides exhibiting improved mechanical properties and corrosion resistance

DOEpatents

Liu, Chain T.; McKamey, Claudette G.; Tortorelli, Peter F.; David, Stan A.

1994-01-01

The specification discloses a corrosion-resistant intermetallic alloy comprising, in atomic percent, an FeAl iron aluminide containing from about 30 to about 40% aluminum alloyed with from about 0.01 to 0.4% zirconium and from 0.01 to about 0.8% boron. The alloy exhibits considerably improved room temperature ductility for enhanced usefulness in structural applications. The high temperature strength and fabricability is improved by alloying with molybdenum, carbon, chromium and vanadium.

40 CFR 421.332 - Effluent limitations guidelines representing the degree of effluent reduction attainable by the...

Code of Federal Regulations, 2013 CFR

2013-07-01

... times. (d) SiCl4 purification wet air pollution control. BPT Limitations for the Primary Zirconium and....600 307.600 pH (1) (1) 1 Within the range of 7.5 to 10.0 at all times. (q) Leaching rinse water from... Within the range of 7.5 to 10.0. (r) Leaching rinse water from zirconium alloy production. BPT...

40 CFR 421.333 - Effluent limitations guidelines representing the degree of effluent reduction attainable by the...

Code of Federal Regulations, 2011 CFR

2011-07-01

... 2.387 1.108 Nickel 4.688 3.154 Ammonia (as N) 1,136.000 499.500 (d) SiCl4 purification wet air....674 5.835 Ammonia (as N) 2,102.000 895.000 (q) Leaching rinse water from zirconium metal production... Ammonia (as N) 7,856.000 3,453.000 (r) Leaching rinse water from zirconium alloy production. BAT...

40 CFR 421.332 - Effluent limitations guidelines representing the degree of effluent reduction attainable by the...

Code of Federal Regulations, 2011 CFR

2011-07-01

... times. (d) SiCl4 purification wet air pollution control. BPT Limitations for the Primary Zirconium and....600 307.600 pH (1) (1) 1 Within the range of 7.5 to 10.0 at all times. (q) Leaching rinse water from... Within the range of 7.5 to 10.0. (r) Leaching rinse water from zirconium alloy production. BPT...

40 CFR 421.332 - Effluent limitations guidelines representing the degree of effluent reduction attainable by the...

Code of Federal Regulations, 2012 CFR

2012-07-01

... times. (d) SiCl4 purification wet air pollution control. BPT Limitations for the Primary Zirconium and....600 307.600 pH (1) (1) 1 Within the range of 7.5 to 10.0 at all times. (q) Leaching rinse water from... Within the range of 7.5 to 10.0. (r) Leaching rinse water from zirconium alloy production. BPT...

40 CFR 421.332 - Effluent limitations guidelines representing the degree of effluent reduction attainable by the...

Code of Federal Regulations, 2014 CFR

2014-07-01

... times. (d) SiCl4 purification wet air pollution control. BPT Limitations for the Primary Zirconium and....600 307.600 pH (1) (1) 1 Within the range of 7.5 to 10.0 at all times. (q) Leaching rinse water from... Within the range of 7.5 to 10.0. (r) Leaching rinse water from zirconium alloy production. BPT...

40 CFR 421.332 - Effluent limitations guidelines representing the degree of effluent reduction attainable by the...

Code of Federal Regulations, 2010 CFR

2010-07-01

... times. (d) SiCl4 purification wet air pollution control. BPT Limitations for the Primary Zirconium and....600 307.600 pH (1) (1) 1 Within the range of 7.5 to 10.0 at all times. (q) Leaching rinse water from... Within the range of 7.5 to 10.0. (r) Leaching rinse water from zirconium alloy production. BPT...

40 CFR 421.333 - Effluent limitations guidelines representing the degree of effluent reduction attainable by the...

Code of Federal Regulations, 2010 CFR

2010-07-01

... 2.387 1.108 Nickel 4.688 3.154 Ammonia (as N) 1,136.000 499.500 (d) SiCl4 purification wet air....674 5.835 Ammonia (as N) 2,102.000 895.000 (q) Leaching rinse water from zirconium metal production... Ammonia (as N) 7,856.000 3,453.000 (r) Leaching rinse water from zirconium alloy production. BAT...

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Montgomery, Robert; Tomé, Carlos; Liu, Wenfeng

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. CASL has endeavored to improve upon this approach by incorporating a microstructurally-based, atomistically-informed, zirconium alloy mechanical deformation analysis capability into the BISON-CASL engineering scale fuel performance code. Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed bymore » Lebensohn and Tome´ [2], has been coupled with BISON-CASL to represent the mechanistic material processes controlling the deformation behavior of the cladding. A critical component of VPSC is the representation of the crystallographic orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON-CASL and provides initial results utilizing the coupled functionality.« less

Preparation and Corrosion Resistance of Trivalent Chromium-Zirconium Composite Coating

NASA Astrophysics Data System (ADS)

Huang, J. Z.

2018-05-01

Aluminum alloys are widely used in the various industries because of its superior advantages. However there will be a thin oxide layer on the surface of the pure aluminum to inhibit corrosion, when adding some other elements, the obtained aluminum alloy is easy to be corroded. Surface protection is an important means to improve the corrosion resistance of aluminum alloys. The formal research had already confirmed that the trivalent chromium conversion coating can significantly improve the corrosion resistance, and the usage of the zirconium solution can also protect the aluminum alloy from corrosion. In this study, we constructed the binary conversion coating with the Cr2(SO4)3 and the K2ZrF6. The optimum reaction conditions are as follows: 10g/L H3PO4, 2g/L K2ZrF6, 28g/L Cr2(SO4)3, pH=2.5∼3.5, temperature 40°C, and reaction time 10 min. Copper sulfate titration experiment confirmed that the corrosion resistance was significantly improved.

Container for reprocessing and permanent storage of spent nuclear fuel assemblies

DOEpatents

Forsberg, Charles W.

1992-01-01

A single canister process container for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining their integrity at temperature necessary to oxide the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container.

Method to predict relative hydriding within a group of zirconium alloys under nuclear irradiation

DOEpatents

Johnson, Jr., A. Burtron; Levy, Ira S.; Trimble, Dennis J.; Lanning, Donald D.; Gerber, Franna S.

1990-01-01

An out-of-reactor method for screening to predict relative in-reactor hydriding behavior of zirconium-bsed materials is disclosed. Samples of zirconium-based materials having different composition and/or fabrication are autoclaved in a relatively concentrated (0.3 to 1.0M) aqueous lithium hydroxide solution at constant temperatures within the water reactor coolant temperature range (280.degree. to 316.degree. C.). Samples tested by this out-of-reactor procedure, when compared on the basis of the ratio of hydrogen weight gain to oxide weight gain, accurately predict the relative rate of hyriding for the same materials when subject to in-reactor (irradiated) corrision.

Ternary cobalt-molybdenum-zirconium coatings for alternative energies

NASA Astrophysics Data System (ADS)

Yar-Mukhamedova, Gulmira; Ved', Maryna; Sakhnenko, Nikolay; Koziar, Maryna

2017-11-01

Consistent patterns for electrodeposition of Co-Mo-Zr coatings from polyligand citrate-pyrophosphate bath were investigated. The effect of both current density amplitude and pulse on/off time on the quality, composition and surface morphology of the galvanic alloys were determined. It was established the coating Co-Mo-Zr enrichment by molybdenum with current density increasing up to 8 A dm-2 as well as the rising of pulse time and pause duration promotes the content of molybdenum because of subsequent chemical reduction of its intermediate oxides by hydrogen ad-atoms. It was found that the content of the alloying metals in the coating Co-Mo-Zr depends on the current density and on/off times extremely and maximum Mo and Zr content corresponds to the current density interval 4-6 A dm-2, on-/off-time 2-10 ms. Chemical resistance of binary and ternary coatings based on cobalt is caused by the increased tendency to passivity and high resistance to pitting corrosion in the presence of molybdenum and zirconium, as well as the acid nature of their oxides. Binary coating with molybdenum content not less than 20 at.% and ternary ones with zirconium content in terms of corrosion deep index are in a group ;very proof;. It was shown that Co-Mo-Zr alloys exhibits the greatest level of catalytic properties as cathode material for hydrogen electrolytic production from acidic media which is not inferior a platinum electrode. The deposits Co-Mo-Zr with zirconium content 2-4 at.% demonstrate high catalytic properties in the carbon(II) oxide conversion. This confirms the efficiency of materials as catalysts for the gaseous wastes purification and gives the reason to recommend them as catalysts for red-ox processes activating by oxygen as well as electrode materials for red-ox batteries.

Vacuum-Induction, Vacuum-Arc, and Air-Induction Melting of a Complex Heat-Resistant Alloy

NASA Technical Reports Server (NTRS)

Decker, R. F.; Rowe, John P.; Freeman, J. W.

1959-01-01

The relative hot-workability and creep-rupture properties at 1600 F of a complex 55Ni-20Cr-15Co-4Mo-3Ti-3Al alloy were evaluated for vacuum-induction, vacuum-arc, and air-induction melting. A limited study of the role of oxygen and nitrogen and the structural effects in the alloy associated with the melting process was carried out. The results showed that the level of boron and/or zirconium was far more influential on properties than the melting method. Vacuum melting did reduce corner cracking and improve surface during hot-rolling. It also resulted in more uniform properties within heats. The creep-rupture properties were slightly superior in vacuum heats at low boron plus zirconium or in heats with zirconium. There was little advantage at high boron levels and air heats were superior at high levels of boron plus zirconium. Vacuum heats also had fewer oxide and carbonitride inclusions although this was a function of the opportunity for separation of the inclusions from high oxygen plus nitrogen heats. The removal of phosphorous by vacuum melting was not found to be related to properties. Oxygen plus nitrogen appeared to increase ductility in creep-rupture tests suggesting that vacuum melting removes unidentified elements detrimental to ductility. Oxides and carbonitrides in themselves did not initiate microcracks. Carbonitrides in the grain boundaries of air heats did initiate microcracks. The role of microcracking from this source and as a function of oxygen and nitrogen content was not clear. Oxygen and nitrogen did intensify corner cracking during hot-rolling but were not responsible for poor surface which resulted from rolling heats melted in air.

Concept Feasibility Report for Electroplating Zirconium onto Uranium Foil - Year 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Coffey, Greg W.; Meinhardt, Kerry D.; Joshi, Vineet V.

2015-03-01

The Fuel Fabrication Capability within the U.S. High Performance Research Reactor Conversion Program is funded through the National Nuclear Security Administration (NNSA) NA-26 (Office of Material Management and Minimization). An investigation was commissioned to determine the feasibility of using electroplating techniques to apply a coating of zirconium onto depleted uranium/molybdenum alloy (U-10Mo). Electroplating would provide an alternative method to the existing process of hot roll-bonding zirconium foil onto the U-10Mo fuel foil during the fabrication of fuel elements for high-performance research reactors. The objective of this research was to develop a reproducible and scalable plating process that will produce amore » uniform, 25 μm thick zirconium metal coating on U-10Mo foil. In previous work, Pacific Northwest National Laboratory (PNNL) established a molten salt electroplating apparatus and protocol to plate zirconium metal onto molybdenum foil (Coffey 2015). During this second year of the research, PNNL furthered this work by moving to the U-10Mo alloy system (90 percent uranium:10 percent molybdenum). The original plating apparatus was disassembled and re-assembled in a laboratory capable of handling low-level radioactive materials. Initially, the work followed the previous year’s approach, and the salt bath composition was targeted at the eutectic composition (LiF:NaF:ZrF4 = 26:37:37 mol%). Early results indicated that the formation of uranium fluoride compounds would be problematic. Other salt bath compositions were investigated in order to eliminate the uranium fluoride production (LiF:NaF = 61:39 mol% and LiF:NaF:KF = 46.5:11.5:42 mol% ). Zirconium metal was used as the crucible for the molten salt. Three plating methods were used—isopotential, galvano static, and pulsed plating. The molten salt method for zirconium metal application provided high-quality plating on molybdenum in PNNL’s previous work. A key advantage of this approach is that plating can be performed under conditions that would greatly reduce the quantity of intermetallics that form at the interface between the zirconium and U-10Mo; unlike roll bonding, the molten salt plating approach would allow for complete coverage of the U-10Mo foil with zirconium. When utilizing the experimental parameters developed for zirconium plating onto molybdenum, a uranium fluoride reaction product was formed at the Zr/U-10Mo interface. By controlling the initial plating potential, the uranium fluoride could be prevented; however, the targeted zirconium thickness (25 ±12.5 μm) could not be achieved while maintaining 100% coverage.« less

STUDY OF THE OXIDATION OF NON-ALLOYED ZIRCONIUM AND OF OXYGEN DIFFUSION IN THE OXIDE FILM AND IN THE METAL (in French)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Debuigne, J.; Lehr, P.

1963-12-01

The oxidation processes of zirconium at 600-850 deg C were studied. A micrographic and radiocrystallographic analysis of the oxide layers formed at the surface of the metal was carried out. The kinetic results, weight gains as function nf time, were completed by the study of oxygen diffusion through the oxide layer formed and in the underlying metal. (auth)

Weight of Polyethylene Wear Particles is Similar in TKAs with Oxidized Zirconium and Cobalt-chrome Prostheses

PubMed Central

Kim, Jun-Shik; Huh, Wansoo; Lee, Kwang-Hoon

2009-01-01

Background The greater lubricity and resistance to scratching of oxidized zirconium femoral components are expected to result in less polyethylene wear than cobalt-chrome femoral components. Questions/purposes We examined polyethylene wear particles in synovial fluid and compared the weight, size (equivalent circle diameter), and shape (aspect ratio) of polyethylene wear particles in knees with an oxidized zirconium femoral component with those in knees with a cobalt-chrome femoral component. Patients and Methods One hundred patients received an oxidized zirconium femoral component in one knee and a cobalt-chrome femoral component in the other. There were 73 women and 27 men with a mean age of 55.6 years (range, 44–60 years). The minimum followup was 5 years (mean, 5.5 years; range, 5–6 years). Polyethylene wear particles were analyzed using thermogravimetric methods and scanning electron microscopy. Results The weight of polyethylene wear particles produced at the bearing surface was 0.0223 ± 0.0054 g in 1 g synovial fluid in patients with an oxidized zirconium femoral component and 0.0228 ± 0.0062 g in patients with a cobalt-chrome femoral component. Size and shape of polyethylene wear particles were 0.59 ± 0.05 μm and 1.21 ± 0.24, respectively, in the patients with an oxidized zirconium femoral component and 0.52 ± 0.03 μm and 1.27 ± 0.31, respectively, in the patients with a cobalt-chrome femoral component. Knee Society knee and function scores, radiographic results, and complication rate were similar between the knees with an oxidized zirconium and cobalt-chrome femoral component. Conclusions The weight, size, and shape of polyethylene wear particles were similar in the knees with an oxidized zirconium and a cobalt-chrome femoral component. We found the theoretical advantages of this surface did not provide the actual advantage. Level of Evidence Level I, therapeutic study. See the guidelines for Authors for a complete description of levels of evidence. PMID:19949906

The 5-year Results of an Oxidized Zirconium Femoral Component for TKA

PubMed Central

Innocenti, Massimo; Carulli, Christian; Matassi, Fabrizio; Villano, Marco

2009-01-01

Osteolysis secondary to polyethylene wear is one of the major factors limiting long-term performance of TKA. Oxidized zirconium is a new material that combines the strength of a metal with the wear properties of a ceramic. It remains unknown whether implants with a zirconium femoral component can be used safely in TKA. To answer that question, we reviewed, at a minimum of 5 years, the clinical outcome and survivorship of a ceramic-surfaced oxidized zirconium femoral component implanted during 98 primary TKAs between April 2001 and December 2003. Survivorship was 98.7% at 7 years postoperatively. No revision was necessary and only one component failed because of aseptic loosening. Mean Knee Society score improved from 36 to 89. No adverse events were observed clinically or radiologically. These results justify pursuing the use of oxidized zirconium as an alternative bearing surface for a femoral component in TKA. Level of Evidence: Level IV, therapeutic study. See Guidelines for Authors for a complete description of levels of evidence. PMID:19798541

Obtaining and Mechanical Properties of Ti-Mo-Zr-Ta Alloys

NASA Astrophysics Data System (ADS)

Bălţatu, M. S.; Vizureanu, P.; Geantă, V.; Nejneru, C.; Țugui, C. A.; Focşăneanu, S. C.

2017-06-01

Ti-based alloys are successfully used in the area of orthopedic biomaterials for their enhanced biocompatibility, good corrosion and mechanical properties. The most suitable metals as an alloying element for orthopedic biomaterials are zirconium, molybdenum and tantalum because are non toxic and have good properties. The paper purpose development of two alloys of Ti-Mo-Zr-Ta (TMZT) prepared by arc-melting with several mechanical properties determined by microindentation. The mechanical properties analyzed was Vickers hardness and dynamic elasticity modulus. The investigated alloys presents a low Young’s modulus, an important condition of biomaterials for preventing stress shielding phenomenon.

Hydrogen in Ti and Zr alloys: industrial perspective, failure modes and mechanistic understanding

NASA Astrophysics Data System (ADS)

Chapman, T. P.; Dye, D.; Rugg, D.

2017-06-01

Titanium is widely used in demanding applications, such as in aerospace. Its strength-to-weight ratio and corrosion resistance make it well suited to highly stressed rotating components. Zirconium has a no less critical application where its low neutron capture cross section and good corrosion resistance in hot water and steam make it well suited to reactor core use, including fuel cladding and structures. The similar metallurgical behaviour of these alloy systems makes it alluring to compare and contrast their behaviour. This is rarely undertaken, mostly because the industrial and academic communities studying these alloys have little overlap. The similarities with respect to hydrogen are remarkable, albeit potentially unsurprising, and so this paper aims to provide an overview of the role hydrogen has to play through the material life cycle. This includes the relationship between alloy design and manufacturing process windows, the role of hydrogen in degradation and failure mechanisms and some of the underpinning metallurgy. The potential role of some advanced experimental and modelling techniques will also be explored to give a tentative view of potential for advances in this field in the next decade or so. This article is part of the themed issue 'The challenges of hydrogen and metals'.

Hydrogen pickup mechanism of zirconium alloys

NASA Astrophysics Data System (ADS)

Couet, Adrien

Although the optimization of zirconium based alloys has led to significant improvements in hydrogen pickup and corrosion resistance, the mechanisms by which such alloy improvements occur are still not well understood. In an effort to understand such mechanisms, a systematic study of the alloy effect on hydrogen pickup is conducted, using advanced characterization techniques to rationalize precise measurements of hydrogen pickup. The hydrogen pick-up fraction is accurately measured for a specially designed set of commercial and model alloys to investigate the effects of alloying elements, microstructure and corrosion kinetics on hydrogen uptake. Two different techniques to measure hydrogen concentrations were used: a destructive technique, Vacuum Hot Extraction, and a non-destructive one, Cold Neutron Prompt Gamma Activation Analysis. The results indicate that hydrogen pickup varies not only from alloy to alloy but also during the corrosion process for a given alloy. For instance Zircaloy type alloys show high hydrogen pickup fraction and sub-parabolic oxidation kinetics whereas ZrNb alloys show lower hydrogen pickup fraction and close to parabolic oxidation kinetics. Hypothesis is made that hydrogen pickup result from the need to balance charge during the corrosion reaction, such that the pickup of hydrogen is directly related to (and indivisible of) the corrosion mechanism and decreases when the rate of electron transport or oxide electronic conductivity sigmao xe through the protective oxide increases. According to this hypothesis, alloying elements (either in solid solution or in precipitates) embedded in the oxide as well as space charge variations in the oxide would impact the hydrogen pick-up fraction by modifying sigmaox e, which drives oxidation and hydriding kinetics. Dedicated experiments and modelling were performed to assess and validate these hypotheses. In-situ electrochemical impedance spectroscopy (EIS) experiments were performed on Zircaloy-4 tubes to directly measure the evolution of sigma oxe as function of exposure time. The results show that sigmao xe decreases as function of exposure time and that its variations are directly correlated to the instantaneous hydrogen pickup fraction variations. The electron transport through the oxide layer is thus altered as the oxide grows, reasons for which are yet to be exactly determined. Preliminary results also show that sigma oxe of ZrNb alloys would be much higher compared with Zircaloy-4. Thus, it is confirmed that sigmaox e is a key parameter in the hydrogen and oxidation mechanism. Because the mechanism whereby alloying elements are incorporated into the oxide layer is critical to changing sigmao xe, the evolution of the oxidation state of two common alloying elements, Fe and Nb, when incorporated into the growing oxide layers is investigated using X-Ray Absorption Near-Edge Spectroscopy (XANES) using micro-beam synchrotron radiation on cross sectional oxide samples. The results show that the oxidation of both Fe and Nb is delayed in the oxide layer compared to that of Zr, and that this oxidation delay is related to the variations of the instantaneous hydrogen pick-up fraction with exposure time. The evolution of Nb oxidation as function of oxide depth is also compatible with space charge compensation in the oxide and with an increase in sigmaox e of ZrNb alloys compared to Zircaloys. Finally, various successively complex models from the well-known Wagner oxidation theory to the more complex effect of space charge on oxidation kinetics have been developed. The general purpose of the modeling effort is to provide a rationale for the sub-parabolic oxidation kinetics and demonstrate the correlation with hydrogen pickup fraction. It is directly demonstrated that parabolic oxidation kinetics is associated with high sigmao xe and low space charges in the oxide whereas sub-parabolic oxidation kinetics is associated with lower sigmaox e and higher space charge in the oxide. All these observations helped us to propose a general corrosion mechanism of zirconium alloys involving both oxidation and hydrogen pickup mechanism to better understand and predict the effect of alloying additions on the behavior of zirconium alloys.

Alloy Selection for Accident Tolerant Fuel Cladding in Commercial Light Water Reactors

NASA Astrophysics Data System (ADS)

Rebak, Raul B.

2015-12-01

As a consequence of the March 2011 events at the Fukushima site, the U.S. congress asked the Department of Energy (DOE) to concentrate efforts on the development of nuclear fuels with enhanced accident tolerance. The new fuels had to maintain or improve the performance of current UO2-zirconium alloy rods during normal operation conditions and tolerate the loss of active cooling in the core for a considerably longer time period than the current system. DOE is funding cost-shared research to investigate the behavior of advanced steels both under normal operation conditions in high-temperature water [ e.g., 561 K (288 °C)] and under accident conditions for reaction with superheated steam. Current results show that, under accident conditions, the advanced ferritic steels (1) have orders of magnitude lower reactivity with steam, (2) would generate less hydrogen and heat than the current zirconium alloys, (3) are resistant to stress corrosion cracking under normal operation conditions, and (4) have low general corrosion in water at 561 K (288 °C).

Iron state in iron nanoparticles with and without zirconium

NASA Astrophysics Data System (ADS)

Filippov, V. P.; Khasanov, A. M.; Lauer, Yu. A.

2017-11-01

Mössbauer and X-ray methods are used for investigations of structure, stability and characteristics of pure-iron grain and two iron-zirconium alloys such as Fe + 5 wt.% Zr and Fe + 10 wt.% Zr. The used powder was ground for 24 h in a SPEX Model 8000 mill shaker. Complex nanoparticles are found, which change their properties under milling. Mössbauer spectral parameters are obtained for investigated materials. Milling results in formation of nanosized particles with two states of iron atoms: one main part is pure α-Fe and another part of iron atoms displaced in grain boundaries or defective zones in which hyperfine magnetic splitting decrease to ˜ 30.0 T. In alloys with Zr three iron states are formed in each alloy, main part of iron is in the form of α-Fe and another two states depend on the concentration of Zr and represent iron in grain boundaries with Zr atoms in nearest neighbor. The changing of iron states is discussed.

Phase composition and corrosion resistance of magnesium alloys

NASA Astrophysics Data System (ADS)

Morozova, G. I.

2008-03-01

The effects of phase composition of castable experimental and commercial alloys based on the Mg-Al, Mg-Al-Mn, Mg-Al-Zn-Mn, and Mg-Zn-Zr systems and of the form of existence of iron and hydrogen admixtures on the rate of corrosion of the alloys in 3% solution of NaCl are studied. The roles of heat treatment in the processes of hydrogen charging and phase formation in alloy ML5pch and of hydrogen in the process of formation of zirconium hydrides and zinc zirconides in alloys of the Mg-Zn-Zr system and their effect on the corrosion and mechanical properties of alloy ML12 are discussed.

Electrochemical Impedance Spectroscopy Of Metal Alloys

NASA Technical Reports Server (NTRS)

Macdowell, L. G.; Calle, L. M.

1993-01-01

Report describes use of electrochemical impedance spectroscopy (EIS) to investigate resistances of 19 alloys to corrosion under conditions similar to those of corrosive, chloride-laden seaside environment of Space Transportation System launch site. Alloys investigated: Hastelloy C-4, C-22, C-276, and B-2; Inconel(R) 600, 625, and 825; Inco(R) G-3; Monel 400; Zirconium 702; Stainless Steel 304L, 304LN, 316L, 317L, and 904L; 20Cb-3; 7Mo+N; ES2205; and Ferralium 255. Results suggest electrochemical impedance spectroscopy used to predict corrosion performances of metal alloys.

Conventionally cast and forged copper alloy for high-heat-flux thrust chambers

NASA Technical Reports Server (NTRS)

Kazaroff, John M.; Repas, George A.

1987-01-01

The combustion chamber liner of the space shuttle main engine is made of NARloy-Z, a copper-silver-zirconium alloy. This alloy was produced by vacuum melting and vacuum centrifugal casting; a production method that is currently now available. Using conventional melting, casting, and forging methods, NASA has produced an alloy of the same composition called NASA-Z. This report compares the composition, microstructure, tensile properties, low-cycle fatigue life, and hot-firing life of these two materials. The results show that the materials have similar characteristics.

Quantifying the stress fields due to a delta-hydride precipitate in alpha-Zr matrix

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tummala, Hareesh; Capolungo, Laurent; Tome, Carlos N.

This report is a preliminary study on δ-hydride precipitate in zirconium alloy performed using 3D discrete dislocation dynamics simulations. The ability of dislocations in modifying the largely anisotropic stress fields developed by the hydride particle in a matrix phase is addressed for a specific dimension of the hydride. The influential role of probable dislocation nucleation at the hydride-matrix interface is reported. Dislocation nucleation around a hydride was found to decrease the shear stress (S 13) and also increase the normal stresses inside the hydride. We derive conclusions on the formation of stacks of hydrides in zirconium alloys. The contribution ofmore » mechanical fields due to dislocations was found to have a non-negligible effect on such process.« less

Directionally solidified eutectic alloy gamma-beta

NASA Technical Reports Server (NTRS)

Tewari, S. N.

1977-01-01

A pseudobinary eutectic alloy composition was determined by a previously developed bleed-out technique. The directionally solidified eutectic alloy with a composition of Ni-37.4Fe-10.0Cr-9.6Al (in wt%) had tensile strengths decreasing from 1,090 MPa at room temperature to 54 MPa at 1,100 C. The low density, excellent microstructural stability, and oxidation resistance of the alloy during thermal cycling suggest that it might have applicability as a gas turbine vane alloy while its relatively low high temperature strength precludes its use as a blade alloy. A zirconium addition increased the 750 C strength, and a tungsten addition was ineffective. The gamma=beta eutectic alloys appeared to obey a normal freezing relation.

Titanium Brazing for Structures and Survivability

DTIC Science & Technology

2007-05-01

materials, such as ceramics. This work focuses on vacuum brazing of titanium (both Ti- 6Al - 4V and commercially pure titanium ) and the effect of...such as ceramics. This work focuses on vacuum brazing of titanium (both Ti- 6Al - 4V and commercially pure titanium ) and the effect of processing...Suzumura, and Onzawa, reported the joining of Ti- 6Al - 4V and CP titanium alloys with zirconium-rich braze alloys.5 They found that these alloys could

BRAZE BONDING OF COLUMBIUM

DOEpatents

Heestand, R.L.; Picklesimer, M.L.

1962-07-31

A method of brazing niobium parts together is described. The surfaces of the parts to be brazed together are placed in abutting relationship with a brazing alloy disposed adjacent. The alloy consists essentially of, by weight, 12 to 25% niobium, 0.5 to 5% molybdenum, and the balance zirconium, The alloy is heated to at least its melting point to braze the parts together. The brazed joint is then cooled. The heating, melting and cooling take place in an inert atmosphere. (AEC)

Effects of Plasma ZrN Metallurgy and Shot Peening Duplex Treatment on Fretting Wear and Fretting Fatigue Behavior of Ti6Al4V Alloy.

PubMed

Tang, Jingang; Liu, Daoxin; Zhang, Xiaohua; Du, Dongxing; Yu, Shouming

2016-03-23

A metallurgical zirconium nitride (ZrN) layer was fabricated using glow metallurgy using nitriding with zirconiuming prior treatment of the Ti6Al4V alloy. The microstructure, composition and microhardness of the corresponding layer were studied. The influence of this treatment on fretting wear (FW) and fretting fatigue (FF) behavior of the Ti6Al4V alloy was studied. The composite layer consisted of an 8-μm-thick ZrN compound layer and a 50-μm-thick nitrogen-rich Zr-Ti solid solution layer. The surface microhardness of the composite layer is 1775 HK 0.1 . A gradient in cross-sectional microhardness distribution exists in the layer. The plasma ZrN metallurgical layer improves the FW resistance of the Ti6Al4V alloy, but reduces the base FF resistance. This occurs because the improvement in surface hardness results in lowering of the toughness and increasing in the notch sensitivity. Compared with shot peening treatment, plasma ZrN metallurgy and shot peening composite treatment improves the FW resistance and enhances the FF resistance of the Ti6Al4V alloy. This is attributed to the introduction of a compressive stress field. The combination of toughness, strength, FW resistance and fatigue resistance enhance the FF resistance for titanium alloy.

Effects of Plasma ZrN Metallurgy and Shot Peening Duplex Treatment on Fretting Wear and Fretting Fatigue Behavior of Ti6Al4V Alloy

PubMed Central

Tang, Jingang; Liu, Daoxin; Zhang, Xiaohua; Du, Dongxing; Yu, Shouming

2016-01-01

A metallurgical zirconium nitride (ZrN) layer was fabricated using glow metallurgy using nitriding with zirconiuming prior treatment of the Ti6Al4V alloy. The microstructure, composition and microhardness of the corresponding layer were studied. The influence of this treatment on fretting wear (FW) and fretting fatigue (FF) behavior of the Ti6Al4V alloy was studied. The composite layer consisted of an 8-μm-thick ZrN compound layer and a 50-μm-thick nitrogen-rich Zr–Ti solid solution layer. The surface microhardness of the composite layer is 1775 HK0.1. A gradient in cross-sectional microhardness distribution exists in the layer. The plasma ZrN metallurgical layer improves the FW resistance of the Ti6Al4V alloy, but reduces the base FF resistance. This occurs because the improvement in surface hardness results in lowering of the toughness and increasing in the notch sensitivity. Compared with shot peening treatment, plasma ZrN metallurgy and shot peening composite treatment improves the FW resistance and enhances the FF resistance of the Ti6Al4V alloy. This is attributed to the introduction of a compressive stress field. The combination of toughness, strength, FW resistance and fatigue resistance enhance the FF resistance for titanium alloy. PMID:28773345

Versatile Oxide Films Protect FeCrAl Alloys Under Normal Operation and Accident Conditions in Light Water Power Reactors

NASA Astrophysics Data System (ADS)

Rebak, Raul B.

2018-02-01

The US has currently a fleet of 99 nuclear power light water reactors which generate approximately 20% of the electricity consumed in the country. Near 90% of the reactors are at least 30 years old. There are incentives to make the existing reactors safer by using accident tolerant fuels (ATF). Compared to the standard UO2-zirconium-based system, ATF need to tolerate loss of active cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. Ferritic iron-chromium-aluminum (FeCrAl) alloys have been identified as an alternative to replace current zirconium alloys. They contain Fe (base) + 10-22 Cr + 4-6 Al and may contain smaller amounts of other elements such as molybdenum and traces of others. FeCrAl alloys offer outstanding resistance to attack by superheated steam by developing an alumina oxide on the surface in case of a loss of coolant accident like at Fukushima. FeCrAl alloys also perform well under normal operation conditions both in boiling water reactors and pressurized water reactors because they are protected by a thin oxide rich in chromium. Under normal operation condition, the key element is Cr and under accident conditions it is Al.

No difference in in vivo polyethylene wear particles between oxidized zirconium and cobalt-chromium femoral component in total knee arthroplasty.

PubMed

Minoda, Yukihide; Hata, Kanako; Iwaki, Hiroyoshi; Ikebuchi, Mitsuhiko; Hashimoto, Yusuke; Inori, Fumiaki; Nakamura, Hiroaki

2014-03-01

Polyethylene wear particle generation is one of the most important factors affecting mid- to long-term results of total knee arthroplasties. Oxidized zirconium was introduced as a material for femoral components to reduce polyethylene wear generation. However, an in vivo advantage of oxidized zirconium on polyethylene wear particle generation is still controversial. The purpose of this study was to compare in vivo polyethylene wear particles between oxidized zirconium total knee prosthesis and conventional cobalt-chromium (Co-Cr) total knee prosthesis. Synovial fluid was obtained from the knees of 6 patients with oxidized zirconium total knee prosthesis and from 6 patients with conventional cobalt-chromium (Co-Cr) total knee prosthesis 12 months after the operation. Polyethylene particles were isolated and examined using a scanning electron microscope and image analyser. Total number of particles in each knee was 3.3 ± 1.3 × 10(7) in the case of oxidized zirconium (mean ± SD) and 3.4 ± 1.2 × 10(7) in that of Co-Cr (n.s.). The particle size (equivalent circle diameter) was 0.8 ± 0.3 μm in the case of oxidized zirconium and 0.6 ± 0.1 μm in that of Co-Cr (n.s.). The particle shape (aspect ratio) was 1.4 ± 0.0 in the case of oxidized zirconium and 1.4 ± 0.0 in that of metal Co-Cr (n.s). Although newly introduced oxidized zirconium femoral component did not reduce the in vivo polyethylene wear particles in early clinical stage, there was no adverse effect of newly introduced material. At this moment, there is no need to abandon oxidized zirconium femoral component. However, further follow-up of polyethylene wear particle generation should be performed to confirm the advantage of the oxidized zirconium femoral component. Therapeutic study, Level III.

Container for reprocessing and permanent storage of spent nuclear fuel assemblies

DOEpatents

Forsberg, C.W.

1992-03-24

A single canister process container is described for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining its integrity at a temperature necessary to oxidize the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container. 10 figs.

Method to predict relative hydriding within a group of zirconium alloys under nuclear irradiation

DOEpatents

Johnson, A.B. Jr.; Levy, I.S.; Trimble, D.J.; Lanning, D.D.; Gerber, F.S.

1990-04-10

An out-of-reactor method for screening to predict relative in-reactor hydriding behavior of zirconium-based materials is disclosed. Samples of zirconium-based materials having different compositions and/or fabrication methods are autoclaved in a relatively concentrated (0.3 to 1.0M) aqueous lithium hydroxide solution at constant temperatures within the water reactor coolant temperature range (280 to 316 C). Samples tested by this out-of-reactor procedure, when compared on the basis of the ratio of hydrogen weight gain to oxide weight gain, accurately predict the relative rate of hydriding for the same materials when subject to in-reactor (irradiated) corrosion. 1 figure.

A review of refractory materials for vapor-anode AMTEC cells

NASA Astrophysics Data System (ADS)

King, Jeffrey C.; El-Genk, M. S.

2000-01-01

Recently, refractory alloys have been considered as structural materials for vapor-anode Alkali Metal Thermal-to-Electric Conversion (AMTEC) cells, for extended (7-15 years) space missions. This paper reviewed the existing database for refractory metals and alloys of potential use as structural materials for vapor-anode sodium AMTEC cells. In addition to requiring that the vapor pressure of the material be below 10-9 torr (133 nPa) at a typical hot side temperature of 1200 K, other screening considerations were: (a) low thermal conductivity, low thermal radiation emissivity, and low linear thermal expansion coefficient; (b) low ductile-to-brittle transition temperature, high yield and rupture strengths and high strength-to-density ratio; and (c) good compatibility with the sodium AMTEC operating environment, including high corrosion resistance to sodium in both the liquid and vapor phases. Nb-1Zr (niobium-1% zirconium) alloy is recommended for the hot end structures of the cell. The niobium alloy C-103, which contains the oxygen gettering elements zirconium and hafnium as well as titanium, is recommended for the colder cell structure. This alloy is stronger and less thermally conductive than Nb-1Zr, and its use in the cell wall reduces parasitic heat losses by conduction to the condenser. The molybdenum alloy Mo-44.5Re (molybdenum-44.5% rhenium) is also recommended as a possible alternative for both structures if known problems with oxygen pick up and embrittlement of the niobium alloys proves to be intractable. .

Ductile tungsten-nickel-alloy and method for manufacturing same

DOEpatents

Ludwig, Robert L.

1978-01-01

The tensile elongation of a tungsten-nickel-iron alloy containing essentially 95 weight percent reprocessed tungsten, 3.5 weight percent nickel, and 1.5 weight percent iron is increased from a value of less than about 1 percent up to about 23 percent by the addition of less than 0.5 weight percent of a reactive metal consisting of niobium and zirconium.

Nickel base alloy. [for gas turbine engine stator vanes

NASA Technical Reports Server (NTRS)

Freche, J. C.; Waters, W. J. (Inventor)

1977-01-01

A nickel base superalloy for use at temperatures of 2000 F (1095 C) to 2200 F (1205 C) was developed for use as stator vane material in advanced gas turbine engines. The alloy has a nominal composition in weight percent of 16 tungsten, 7 aluminum, 1 molybdenum, 2 columbium, 0.3 zirconium, 0.2 carbon and the balance nickel.

Oxidation behaviour of zirconium alloys and their precipitates - A mechanistic study

NASA Astrophysics Data System (ADS)

Proff, C.; Abolhassani, S.; Lemaignan, C.

2013-01-01

The precipitate oxidation behaviour of binary zirconium alloys containing 1 wt.% Fe, Ni, Cr or 0.6 wt.% Nb was characterised in TEM on FIB prepared transverse sections of the oxide and reported in previous studies [1,2]. In the present study the following alloys: Zr1%Cu, Zr0.5%Cu0.5%Mo and pure Zr are analysed to add to the available information. In all cases, the observed precipitate oxidation behaviour in the oxide close to the metal-oxide interface could be described either with delayed oxidation with respect to the matrix or simultaneous oxidation as the surrounding zirconium matrix. Attempt was made to explain these observations, with different parameters such as precipitate size and structure, composition and thermodynamic properties. It was concluded that the thermodynamics with the new approach presented could explain most precisely their behaviour, considering the precipitate stoichiometry and the free energy of oxidation of the constituting elements. The surface topography of the oxidised materials, as well as the microstructure of the oxide presenting microcracks have been examined. A systematic presence of microcracks above the precipitates exhibiting delayed oxidation has been found; the height of these crack calculated using the Pilling-Bedworth ratios of different phases present, can explain their origin. The protrusions at the surface in the case of materials containing large precipitates can be unambiguously correlated to the presence of these latter, and the height can be correlated to the Pilling-Bedworth ratios of the phases present as well as the diffusion of the alloying elements to the surface and their subsequent oxidation. This latter behaviour was much more considerable in the case of Fe and Cu with Fe showing systematically diffusion to the outer surface.

Metal alloy coatings and methods for applying

DOEpatents

Merz, Martin D.; Knoll, Robert W.

1991-01-01

A method of coating a substrate comprises plasma spraying a prealloyed feed powder onto a substrate, where the prealloyed feed powder comprises a significant amount of an alloy of stainless steel and at least one refractory element selected from the group consisting of titanium, zirconium, hafnium, niobium, tantalum, molybdenum, and tungsten. The plasma spraying of such a feed powder is conducted in an oxygen containing atmosphere and forms an adherent, corrosion resistant, and substantially homogenous metallic refractory alloy coating on the substrate.

METHOD OF FORMING A PROTECTIVE COATING ON FERROUS METAL SURFACES

DOEpatents

Schweitzer, D.G.; Weeks, J.R.; Kammerer, O.F.; Gurinsky, D.H.

1960-02-23

A method is described of protecting ferrous metal surfaces from corrosive attack by liquid metals, such as liquid bismuth or lead-bismuth alloys. The nitrogen content of the ferrous metal surface is first reduced by reacting the metal surface with a metal which forms a stable nitride. Thereafter, the surface is contacted with liquid metal containing at least 2 ppm zirconium at a temperature in the range of 550 to 1100 deg C to form an adherent zirconium carbide layer on the ferrous surface.

Fracture toughness of copper-base alloys for ITER applications: A preliminary report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Alexander, D.J.; Zinkle, S.J.; Rowcliffe, A.F.

1997-04-01

Oxide-dispersion strengthened copper alloys and a precipitation-hardened copper-nickel-beryllium alloy showed a significant reduction in toughness at elevated temperature (250{degrees}C). This decrease in toughness was much larger than would be expected from the relatively modest changes in the tensile properties over the same temperature range. However, a copper-chromium-zirconium alloy strengthened by precipitation showed only a small decrease in toughness at the higher temperatures. The embrittled alloys showed a transition in fracture mode, from transgranular microvoid coalescence at room temperature to intergranular with localized ductility at high temperatures. The Cu-Cr-Zr alloy maintained the ductile microvoid coalescence failure mode at all test temperatures.

Electrochemical Behavior Assessment of As-Cast Mg-Y-RE-Zr Alloy in Phosphate Buffer Solutions (X Na3PO4 + Y Na2HPO4) Using Electrochemical Impedance Spectroscopy and Mott-Schottky Techniques

NASA Astrophysics Data System (ADS)

Fattah-alhosseini, Arash; Asgari, Hamed

2018-05-01

In the present study, electrochemical behavior of as-cast Mg-Y-RE-Zr alloy (RE: rare-earth alloying elements) was investigated using electrochemical tests in phosphate buffer solutions (X Na3PO4 + Y Na2HPO4). X-ray diffraction techniques and Scanning electron microscopy equipped with energy dispersive x-ray spectroscopy were used to investigate the microstructure and phases of the experimental alloy. Different electrochemical tests such as potentiodynamic polarization (PDP), electrochemical impedance spectroscopy (EIS) and Mott-Schottky (M-S) analysis were carried out in order to study the electrochemical behavior of the experimental alloy in phosphate buffer solutions. The PDP curves and EIS measurements indicated that the passive behavior of the as-cast Mg-Y-RE-Zr alloy in phosphate buffer solutions was weakened by an increase in the pH, which is related to formation of an imperfect and less protective passive layer on the alloy surface. The presence of the insoluble zirconium particles along with high number of intermetallic phases of RE elements mainly Mg24Y5 in the magnesium matrix can deteriorate the corrosion performance of the alloy by disrupting the protective passive layer that is formed at pH values over 11. These insoluble zirconium particles embedded in the matrix can detrimentally influence the passivation. The M-S analysis revealed that the formed passive layers on Mg-Y-RE-Zr alloy behaved as an n-type semiconductor. An increase in donor concentration accompanying solutions of higher alkalinity is thought to result in the formation of a less resistive passive layer.

Evaluation of dip and spray coating techniques in corrosion inhibition of AA2024 alloy using a silicon/zirconium sol-gel film as coating

NASA Astrophysics Data System (ADS)

Garcia, R. B. R.; Silva, F. S.; Kawachi, E. Y.

2017-02-01

For corrosion protection of aluminum alloy AA2024 -T3 a silicon/zirconium films were obtained via sol-gel process, prepared from tetraethoxysilane and zirconium acetate, in acid medium with a 5 wt% of nonionic surfactant in order to replace the pre-treatment based on chromium conversion coatings. A homogeneous film was obtained and deposited, at low viscosity condition of the sol (˜10cP), by dip and spray coating techniques. The films morphology was evaluated by Scanning Electron Microscopy (SEM), and to know more about the used deposition methodology, the deposited mass and the film thickness were measured. The corrosion protection efficiency of deposited films was evaluated by potentiodynamic polarization. The film deposition by both dip and spray coatings were effective for the deposition of a homogeneous film layer, and the results showed the thickness is directly related with the deposited mass, and the film deposited by spray technique presented the lower value. Potentiodynamic polarization indicated that the film deposited by spray coating apparently has a better inert ceramic film due the polarization resistance increased around 57% against 27 and 14% of dip coating samples (4 and 1 layer, respectively).

Duct and cladding alloy

DOEpatents

Korenko, Michael K.

1983-01-01

An austenitic alloy having good thermal stability and resistance to sodium corrosion at 700.degree. C. consists essentially of 35-45% nickel 7.5-14% chromium 0.8-3.2% molybdenum 0.3-1.0% silicon 0.2-1.0% manganese 0-0.1% zirconium 2.0-3.5% titanium 1.0-2.0% aluminum 0.02-0.1% carbon 0-0.01% boron and the balance iron.

Tensile Properties of Molybdenum and Tungsten from 2500 to 3700 F

NASA Technical Reports Server (NTRS)

Hall, Robert W.; Sikora, Paul F.

1959-01-01

Specimens of commercially pure sintered tungsten, arc-cast unalloyed molybdenum, and two arc-cast molybdenum-base alloys (one with 0.5 percent titanium, the other with 0.46 percent titanium and 0.07 percent zirconium) were fabricated from 1/2-inch-diameter rolled or swaged bars. All specimens were evaluated in short-time tensile tests in the as-received condition, and all except the molybdenum-titanium-zirconium alloy were tested after a 30-minute recrystallization anneal at 3800 F in a vacuum of approximately 0.1 micron. Results showed that the tungsten was considerably stronger than either the arc-cast unalloyed molybdenum or the molybdenum-base alloys over the 2500 to 3700 F temperature range. Recrystallization of swaged tungsten at 3800 F considerably reduced its tensile strength at 2500 F. However, above 3100 F, the as-swaged tungsten specimens recrystallized during testing, and had about the same strength as when recrystallized at 3800 F before evaluation. The ductility of molybdenum-base materials was very high at all test temperatures; the ductility of tungsten decreased sharply above about 3120 F.

Hydrogen in Ti and Zr alloys: industrial perspective, failure modes and mechanistic understanding.

PubMed

Chapman, T P; Dye, D; Rugg, D

2017-07-28

Titanium is widely used in demanding applications, such as in aerospace. Its strength-to-weight ratio and corrosion resistance make it well suited to highly stressed rotating components. Zirconium has a no less critical application where its low neutron capture cross section and good corrosion resistance in hot water and steam make it well suited to reactor core use, including fuel cladding and structures. The similar metallurgical behaviour of these alloy systems makes it alluring to compare and contrast their behaviour. This is rarely undertaken, mostly because the industrial and academic communities studying these alloys have little overlap. The similarities with respect to hydrogen are remarkable, albeit potentially unsurprising, and so this paper aims to provide an overview of the role hydrogen has to play through the material life cycle. This includes the relationship between alloy design and manufacturing process windows, the role of hydrogen in degradation and failure mechanisms and some of the underpinning metallurgy. The potential role of some advanced experimental and modelling techniques will also be explored to give a tentative view of potential for advances in this field in the next decade or so.This article is part of the themed issue 'The challenges of hydrogen and metals'. © 2017 The Author(s).

Preparation of Aluminum-Zirconium Master Alloy by Aluminothermic Reduction in Cryolite Melt

NASA Astrophysics Data System (ADS)

Liu, Fengguo; Ding, Chenliang; Tao, Wenju; Hu, Xianwei; Gao, Bingliang; Shi, Zhongning; Wang, Zhaowen

2017-12-01

Al-Zr master alloy was prepared by aluminothermic reduction in cryolite melt without alumina impurity. The Al-Zr master alloy was characterized by x-ray diffraction, scanning electron microscopy, and energy-dispersive spectroscopy. The composition of the master alloy was analyzed by inductively coupled plasma optical emission spectrometry. The results indicated that Al-Zr master alloy with high purity could be obtained when byproduct Al2O3 was dissolved in the cryolite melt. The Al-Zr alloy was embedded in the Al matrix in the form of Al3Zr phase with long rod or tetragonal morphology due to temperature variation. Finally, we obtained Al-Zr alloy with 7 wt.% Zr by aluminothermic reduction for 90 min in cryolite melt at 980°C.

Biocompatibility Study of Zirconium-Based Bulk Metallic Glasses for Orthopedic Applications

NASA Astrophysics Data System (ADS)

He, Wei; Chuang, Andrew; Cao, Zheng; Liaw, Peter K.

2010-07-01

Bulk metallic glasses (BMGs) represent an emerging class of materials that offer an attractive combination of properties, such as high strength, low modulus, good fatigue limit, and near-net-shape formability. The BMGs have been explored in mechanical, chemical, and magnetic applications. However, little research has been attracted in the biomedical field. In this work, we study the potential of BMGs for the orthopedic repair and replacement. We report the biocompatibility study of zirconium (Zr)-based solid BMGs using mouse osteoblast cells. Cell attachment, proliferation, and differentiation are compared to Ti-6Al-4V, a well-studied alloy biomaterial. Our in-vitro study has demonstrated that cells cultured on the Zr-based BMG substrate showed higher attachment, alkaline phosphatase activity, and bone matrix deposition compared to those grown on the control Ti alloy substrate. Cytotoxicity staining also revealed the remarkable viability of cells growing on the BMG substrates.

In situ Raman spectroscopic investigation of zirconium-niobium alloy corrosion under hydrothermal conditions

NASA Astrophysics Data System (ADS)

Maslar, J. E.; Hurst, W. S.; Bowers, W. J.; Hendricks, J. H.

2001-10-01

In situ Raman spectroscopy was employed to investigate corrosion of a zirconium-niobium alloy in air-saturated water at a pressure of 15.5 MPa and temperatures ranging from 22 to 407 °C in an optically accessible flow cell. Monoclinic ZrO 2 (m-ZrO 2) was identified under all conditions after the coupon was heated to 255 °C for 19 h. Cubic ZrO 2 (c-ZrO 2) was tentatively identified in situ during heating at temperatures between 306 and 407 °C, but was not observed under any other conditions. Species tentatively identified as α-CrOOH and a Cr VI and/or Cr III/Cr VI compound were observed in situ during heating at temperatures between 255 and 407 °C, but were not observed under any other conditions. The chromium compounds were identified as corrosion products released from the optical cell and/or flow system.

Materials Testing for an Accelerator-Driven Subcritical Molten Salt Fission System: A look at the Materials Science of Molten Salt Corrosion

NASA Astrophysics Data System (ADS)

Sooby, Elizabeth; Balachandran, Shreyas; Foley, David; Hartwig, Karl; McIntyre, Peter; Phongikaroon, Supathorn; Pogue, Nathaniel; Simpson, Michael; Tripathy, Prabhat

2011-10-01

For an accelerator-driven subcritical molten salt fission core to survive its 50+ year fuel life, the primary vessel, heat exchanger, and various internal components must be made of materials that resist corrosion and radiation damage in a high-temperature environment, (500-800 C). An experimental study of the corrosion behavior of candidate metals in contact with molten salt is being conducted at the Center for Advanced Energy Studies. Initial experiments have been run on Nb, Ta, Ni, two zirconium alloys, Hastelloy-N, and a series of steel alloys to form a base line for corrosion in both chloride and bromide salt. Metal coupons were immersed in LiCl-KCl or LiBr-KBr at 700 C in an inert-atmosphere. Salt samples were extracted on a time schedule over a 24-hr period. The samples were analyzed using inductively coupled plasma-mass spectrometry to determine concentrations of metals from corrosion. Preliminary results will be presented.

Compatibility of refractory materials for nuclear reactor poison control systems

NASA Technical Reports Server (NTRS)

Sinclair, J. H.

1974-01-01

Metal-clad poison rods have been considered for the control system of an advanced space power reactor concept studied at the NASA Lewis Research Center. Such control rods may be required to operate at temperatures of about 140O C. Selected poison materials (including boron carbide and the diborides of zirconium, hafnium, and tantalum) were subjected to 1000-hour screening tests in contact with candidate refractory metal cladding materials (including tungsten and alloys of tantalum, niobium, and molybdenum) to assess the compatibility of these materials combinations at the temperatures of interest. Zirconium and hafnium diborides were compatible with refractory metals at 1400 C, but boron carbide and tantalum diboride reacted with the refractory metals at this temperature. Zirconium diboride also showed promise as a reaction barrier between boron carbide and tungsten.

Fluorometric determination of zirconium in minerals

USGS Publications Warehouse

Alford, W.C.; Shapiro, L.; White, C.E.

1951-01-01

The increasing use of zirconium in alloys and in the ceramics industry has created renewed interest in methods for its determination. It is a common constituent of many minerals, but is usually present in very small amounts. Published methods tend to be tedious, time-consuming, and uncertain as to accuracy. A new fluorometric procedure, which overcomes these objections to a large extent, is based on the blue fluorescence given by zirconium and flavonol in sulfuric acid solution. Hafnium is the only element that interferes. The sample is fused with borax glass and sodium carbonate and extracted with water. The residue is dissolved in sulfuric acid, made alkaline with sodium hydroxide to separate aluminum, and filtered. The precipitate is dissolved in sulfuric acid and electrolysed in a Melaven cell to remove iron. Flavonol is then added and the fluorescence intensity is measured with a photo-fluorometer. Analysis of seven standard mineral samples shows excellent results. The method is especially useful for minerals containing less than 0.25% zirconium oxide.

In vitro and in vivo biological performance of porous Ti alloys prepared by powder metallurgy.

PubMed

do Prado, Renata Falchete; Esteves, Gabriela Campos; Santos, Evelyn Luzia De Souza; Bueno, Daiane Acácia Griti; Cairo, Carlos Alberto Alves; Vasconcellos, Luis Gustavo Oliveira De; Sagnori, Renata Silveira; Tessarin, Fernanda Bastos Pereira; Oliveira, Felipe Eduardo; Oliveira, Luciane Dias De; Villaça-Carvalho, Maria Fernanda Lima; Henriques, Vinicius André Rodrigues; Carvalho, Yasmin Rodarte; De Vasconcellos, Luana Marotta Reis

2018-01-01

Titanium (Ti) and Ti-6 Aluminium-4 Vanadium alloys are the most common materials in implants composition but β type alloys are promising biomaterials because they present better mechanical properties. Besides the composition of biomaterial, many factors influence the performance of the biomaterial. For example, porous surface may modify the functional cellular response and accelerate osseointegration. This paper presents in vitro and in vivo evaluations of powder metallurgy-processed porous samples composed by different titanium alloys and pure Ti, aiming to show their potential for biomedical applications. The porous surfaces samples were produced with different designs to in vitro and in vivo tests. Samples were characterized with scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS) and elastic modulus analyses. Osteogenic cells from newborn rat calvaria were plated on discs of different materials: G1-commercially pure Ti group (CpTi); G2-Ti-6Al-4V alloy; G3-Ti-13 Niobium-13 Zirconium alloy; G4-Ti-35 Niobium alloy; G5-Ti-35 Niobium-7 Zirconium-5 Tantalum alloy. Cell adhesion and viability, total protein content, alkaline phosphatase activity, mineralization nodules and gene expression (alkaline phosphatase, Runx-2, osteocalcin and osteopontin) were assessed. After 2 and 4 weeks of implantation in rabbit tibia, bone ingrowth was analyzed using micro-computed tomography (μCT). EDS analysis confirmed the material production of each group. Metallographic and SEM analysis revealed interconnected pores, with mean pore size of 99,5μm and mean porosity of 42%, without significant difference among the groups (p>0.05). The elastic modulus values did not exhibit difference among the groups (p>0.05). Experimental alloys demonstrated better results than CpTi and Ti-6Al-4V, in gene expression and cytokines analysis, especially in early experimental periods. In conclusion, our data suggests that the experimental alloys can be used for biomedical application since they contributed to excellent cellular behavior and osseointegration besides presenting lower elastic modulus.

Development of phase analysis methods of impurity elements in alloys based on iron and nickel

NASA Astrophysics Data System (ADS)

Andreeva, N. A.; Anuchkin, S. N.; Volchenkova, V. A.; Kazenas, E. K.; Penkina, T. N.; Fomina, A. A.

2018-04-01

Using the method of AES with ICP, new methods have been developed for quantifying the content of various forms of existence of impurity elements: Al-Al2O3; Zr-ZrO2 in alloys based on iron (Fe-Sn) and nickel (Ni-Sn). Open systems were used to dissolve Al and Zr. To translate difficult-to-open aluminum oxides (corundum) and zirconium oxide (baddeleyite) into the solution, accelerated techniques were developed using the microwave system Mars 5. To confirm the completeness of the dissolution of oxides, a classical scheme of alloy fusion with alkali metal salts was used. Optimal analytical parameters for determining the elements: Al and Zr were chosen. The influence of matrix elements (iron and nickel) and methods of its elimination were studied. This made it possible to determine the elements in a wide concentration range from 1 • 10-3 to n% Al and from 1 • 10-4 to n% Zr without preliminary separation of the matrix with good metrological characteristics. The relative standard deviation (Sr) does not exceed 0,2. The separate determination of the contents of aluminum and aluminium oxide in the model melt of Fe-Sn-Al2O3 and zirconium and zirconium oxide in the Ni-Sn-ZrO2 model melt allowed us to estimate the number of nanoparticles participating in the heterophase interaction with tin and retired to the interface in the form of ensembles and the number of nanoparticles present in the melt and affecting the crystallization process and the structure of the metal.

A Prospective Case-Control Clinical Study of Titanium-Zirconium Alloy Implants with a Hydrophilic Surface in Patients with Type 2 Diabetes Mellitus.

PubMed

Cabrera-Domínguez, José; Castellanos-Cosano, Lizett; Torres-Lagares, Daniel; Machuca-Portillo, Guillermo

To evaluate prospectively the behavior of narrow-diameter (3.3-mm) titanium-zirconium alloy implants with a hydrophilic surface (Straumann Roxolid SLActive) in patients with type 2 diabetes mellitus in single-unit restorations, compared with a healthy control group (assessed using the glycosylated hemoglobin HbA1c test). The patients evaluated in this study required single-unit implant treatment; 15 patients had type 2 diabetes mellitus, and 14 patients were healthy (control group [CG]). Marginal bone level (MBL) change around the implants was evaluated using conventional, sequential periapical digital radiographs. Patient HbA1c was assessed in each check-up. Normality test (Kolmogorov-Smirnov), univariate and multivariate logistic regression, analysis of variance (ANOVA), and Mann-Whitney U test were used for statistical analysis. No differences in MBL change and implant survival and success rates were found between the diabetes mellitus group (DMG) versus the control group, either during the initial recording (DMG, 0.99 ± 0.56 vs CG, 0.68 ± 0.54; P > .05) or 6 months after restoration (DMG, 1.28 ± 0.38 vs CG, 1.11 ± 0.59; P > .05). No significant correlation between HbA1c levels and MBL change was detected in these patients (P > .05). Patients with glycemic control exhibit similar outcomes to healthy individuals with regard to the investigated parameters. In light of these findings, the titanium-zirconium alloy small-diameter implants can be used in the anterior region of the mouth in type 2 diabetic patients.

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

NASA Astrophysics Data System (ADS)

Montgomery, Robert; Tomé, Carlos; Liu, Wenfeng; Alankar, Alankar; Subramanian, Gopinath; Stanek, Christopher

2017-01-01

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical constitutive models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. To improve upon this approach, a microstructurally-based zirconium alloy mechanical deformation analysis capability is being developed within the United States Department of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL). Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed by Lebensohn and Tomé [1], has been coupled with the BISON engineering scale fuel performance code to represent the mechanistic material processes controlling the deformation behavior of light water reactor (LWR) cladding. A critical component of VPSC is the representation of the crystallographic nature (defect and dislocation movement) and orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. A future goal is for VPSC to obtain information on reaction rate kinetics from atomistic calculations to inform the defect and dislocation behavior models described in VPSC. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON and provides initial results utilizing the coupled functionality.

Atomistic modeling of high temperature uranium-zirconium alloy structure and thermodynamics

NASA Astrophysics Data System (ADS)

Moore, A. P.; Beeler, B.; Deo, C.; Baskes, M. I.; Okuniewski, M. A.

2015-12-01

A semi-empirical Modified Embedded Atom Method (MEAM) potential is developed for application to the high temperature body-centered-cubic uranium-zirconium alloy (γ-U-Zr) phase and employed with molecular dynamics (MD) simulations to investigate the high temperature thermo-physical properties of U-Zr alloys. Uranium-rich U-Zr alloys (e.g. U-10Zr) have been tested and qualified for use as metallic nuclear fuel in U.S. fast reactors such as the Integral Fast Reactor and the Experimental Breeder Reactors, and are a common sub-system of ternary metallic alloys like U-Pu-Zr and U-Zr-Nb. The potential was constructed to ensure that basic properties (e.g., elastic constants, bulk modulus, and formation energies) were in agreement with first principles calculations and experimental results. After which, slight adjustments were made to the potential to fit the known thermal properties and thermodynamics of the system. The potentials successfully reproduce the experimental melting point, enthalpy of fusion, volume change upon melting, thermal expansion, and the heat capacity of pure U and Zr. Simulations of the U-Zr system are found to be in good agreement with experimental thermal expansion values, Vegard's law for the lattice constants, and the experimental enthalpy of mixing. This is the first simulation to reproduce the experimental thermodynamics of the high temperature γ-U-Zr metallic alloy system. The MEAM potential is then used to explore thermodynamics properties of the high temperature U-Zr system including the constant volume heat capacity, isothermal compressibility, adiabatic index, and the Grüneisen parameters.

Hot corrosion of the B2 nickel aluminides

NASA Technical Reports Server (NTRS)

Ellis, David L.

1993-01-01

The hot corrosion behavior of the B2 nickel aluminides was studied to determine the inherent hot corrosion resistance of the beta nickel aluminides and to develop a mechanism for the hot corrosion of the beta nickel aluminides. The effects of the prior processing of the material, small additions of zirconium, stoichiometry of the materials, and preoxidation of the samples were also examined. Additions of 2, 5, and 15 w/o chromium were used to determine the effect of chromium on the hot corrosion of the beta nickel aluminides and the minimum amount of chromium necessary for good hot corrosion resistance. The results indicate that the beta nickel aluminides have inferior inherent hot corrosion resistance despite their excellent oxidation resistance. Prior processing and zirconium additions had no discernible effect on the hot corrosion resistance of the alloys. Preoxidation extended the incubation period of the alloys only a few hours and was not considered to be an effective means of stopping hot corrosion. Stoichiometry was a major factor in determining the hot corrosion resistance of the alloys with the higher aluminum alloys having a definitely superior hot corrosion resistance. The addition of chromium to the alloys stopped the hot corrosion attack in the alloys tested. From a variety of experimental results, a complex hot corrosion mechanism was proposed. During the early stages of the hot corrosion of these alloys the corrosion is dominated by a local sulphidation/oxidation form of attack. During the intermediate stages of the hot corrosion, the aluminum depletion at the surface leads to a change in the oxidation mechanism from a protective external alumina layer to a mixed nickel-aluminum spinel and nickel oxide that can occur both externally and internally. The material undergoes extensive cracking during the later portions of the hot corrosion.

Coating for components requiring hydrogen peroxide compatibility

NASA Technical Reports Server (NTRS)

Yousefiani, Ali (Inventor)

2010-01-01

The present invention provides a heretofore-unknown use for zirconium nitride as a hydrogen peroxide compatible protective coating that was discovered to be useful to protect components that catalyze the decomposition of hydrogen peroxide or corrode when exposed to hydrogen peroxide. A zirconium nitride coating of the invention may be applied to a variety of substrates (e.g., metals) using art-recognized techniques, such as plasma vapor deposition. The present invention further provides components and articles of manufacture having hydrogen peroxide compatibility, particularly components for use in aerospace and industrial manufacturing applications. The zirconium nitride barrier coating of the invention provides protection from corrosion by reaction with hydrogen peroxide, as well as prevention of hydrogen peroxide decomposition.

Method of improving fatigue life of cast nickel based superalloys and composition

DOEpatents

Denzine, Allen F.; Kolakowski, Thomas A.; Wallace, John F.

1978-03-14

The invention consists of a method of producing a fine equiaxed grain structure (ASTM 2-4) in cast nickel-base superalloys which increases low cycle fatigue lives without detrimental effects on stress rupture properties to temperatures as high as 1800.degree. F. These superalloys are variations of the basic nickel-chromium matrix, hardened by gamma prime [Ni.sub.3 (Al, Ti)] but with optional additions of cobalt, tungsten, molybdenum, vanadium, columbium, tantalum, boron, zirconium, carbon and hafnium. The invention grain refines these alloys to ASTM 2 to 4 increasing low cycle fatigue life by a factor of 2 to 5 (i.e. life of 700 hours would be increased to 1400 to 3500 hours for a given stress) as a result of the addition of 0.01% to 0.2% of a member of the group consisting of boron, zirconium and mixtures thereof to aid heterogeneous nucleation. The alloy is vacuum melted and heated to 250.degree.-400.degree. F. above the melting temperature, cooled to partial solidification, thus resulting in said heterogeneous nucleation and fine grains, then reheated and cast at about 50.degree.-100.degree. F. of superheat. Additions of 0.1% boron and 0.1% zirconium (optional) are the preferred nucleating agents.

Molybdenum disilicide composites reinforced with zirconia and silicon carbide

DOEpatents

Petrovic, John J.

1995-01-01

Compositions consisting essentially of molybdenum disilicide, silicon carbide, and a zirconium oxide component. The silicon carbide used in the compositions is in whisker or powder form. The zirconium oxide component is pure zirconia or partially stabilized zirconia or fully stabilized zirconia.

Acceptable aluminum additions for minimal environmental effect in iron-aluminum alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sikka, V.K.; Viswanathan, S.; Vyas, S.

A systematic study of iron-aluminum alloys has shown that Fe-16 at. % Al alloys are not very sensitive to environmental embrittlement. The Fe-22 and -28 at. % Al alloys are sensitive to environmental embrittlement, and the effect can be reduced by the addition of chromium and through the control of grain size by additions of zirconium and carbon. The Fe-16 at. % Al binary, and alloys based on it, yielded over 20% room-temperature (RT) elongation even after high-temperature annealing treatments at 1100[degree]C. The best values for the Fe-22 and -28 at. % Al-base alloys after similar annealing treatments were 5more » and 10%, respectively. A multicomponent alloy, FAP, based on Fe- 16 at. % Al was designed, which gave an RT ductility of over 25%.« less

Acceptable aluminum additions for minimal environmental effect in iron-aluminum alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sikka, V.K.; Viswanathan, S.; Vyas, S.

A systematic study of iron-aluminum alloys has shown that Fe-16 at. % Al alloys are not very sensitive to environmental embrittlement. The Fe-22 and -28 at. % Al alloys are sensitive to environmental embrittlement, and the effect can be reduced by the addition of chromium and through the control of grain size by additions of zirconium and carbon. The Fe-16 at. % Al binary, and alloys based on it, yielded over 20% room-temperature (RT) elongation even after high-temperature annealing treatments at 1100{degree}C. The best values for the Fe-22 and -28 at. % Al-base alloys after similar annealing treatments were 5more » and 10%, respectively. A multicomponent alloy, FAP, based on Fe- 16 at. % Al was designed, which gave an RT ductility of over 25%.« less

Iron-aluminum alloys having high room-temperature and method for making same

DOEpatents

Sikka, Vinod K.; McKamey, Claudette G.

1993-01-01

Iron-aluminum alloys having selectable room-temperature ductilities of greater than 20%, high resistance to oxidation and sulfidation, resistant pitting and corrosion in aqueous solutions, and possessing relatively high yield and ultimate tensile strengths are described. These alloys comprise 8 to 9.5% aluminum, up to 7% chromium, up to 4% molybdenum, up to 0.05% carbon, up to 0.5% of a carbide former such as zirconium, up to 0.1 yttrium, and the balance iron. These alloys in wrought form are annealed at a selected temperature in the range of 700.degree. C. to about 1100.degree. C. for providing the alloys with selected room-temperature ductilities in the range of 20 to about 29%.

Ratcheting fatigue behavior of Zircaloy-2 at room temperature

NASA Astrophysics Data System (ADS)

Rajpurohit, R. S.; Sudhakar Rao, G.; Chattopadhyay, K.; Santhi Srinivas, N. C.; Singh, Vakil

2016-08-01

Nuclear core components of zirconium alloys experience asymmetric stress or strain cycling during service which leads to plastic strain accumulation and drastic reduction in fatigue life as well as dimensional instability of the component. Variables like loading rate, mean stress, and stress amplitude affect the influence of asymmetric loading. In the present investigation asymmetric stress controlled fatigue tests were conducted with mean stress from 80 to 150 MPa, stress amplitude from 270 to 340 MPa and stress rate from 30 to 750 MPa/s to study the process of plastic strain accumulation and its effect on fatigue life of Zircaloy-2 at room temperature. It was observed that with increase in mean stress and stress amplitude accumulation of ratcheting strain was increased and fatigue life was reduced. However, increase in stress rate led to improvement in fatigue life due to less accumulation of ratcheting strain.

Molybdenum disilicide composites reinforced with zirconia and silicon carbide

DOEpatents

Petrovic, J.J.

1995-01-17

Compositions are disclosed consisting essentially of molybdenum disilicide, silicon carbide, and a zirconium oxide component. The silicon carbide used in the compositions is in whisker or powder form. The zirconium oxide component is pure zirconia or partially stabilized zirconia or fully stabilized zirconia.

Method for improving the mechanical properties of uranium-1 to 3 wt % zirconium alloy

DOEpatents

Anderson, R.C.

1983-11-22

A uranium-1 to 3 wt % zirconium alloy characterized by high strength, high ductility and stable microstructure is fabricated by an improved thermal mechanical process. A homogenous ingot of the alloy which has been reduced in thickness of at least 50% in the two-step forging operation, rolled into a plate with a 75% reduction and then heated in vacuum at a temperature of about 750 to 850/sup 0/C and then quenched in water, is subjected to further thermal-mechanical operation steps to increase the compressive yield strength approximately 30%, stabilize the microstructure, and decrease the variations in mechanical properties throughout the plate is provided. These thermal-mechanical steps are achieved by cold rolling the quenchd plate to reduce the thickness thereof about 8 to 12%, aging the cold rolled plate at a first temperature of about 325 to 375/sup 0/C for five to six hours and then aging the plate at a higher temperature ranging from 480 to 500/sup 0/C for five to six hours prior to cooling the billet to ambient conditions and sizing the billet or plate into articles provides the desired increase in mechanical properties and phase stability throughout the plate.

Plasma electrolytic oxidation treatment mode influence on corrosion properties of coatings obtained on Zr-1Nb alloy in silicate-phosphate electrolyte

NASA Astrophysics Data System (ADS)

Farrakhov, R. G.; Mukaeva, V. R.; Fatkullin, A. R.; Gorbatkov, M. V.; Tarasov, P. V.; Lazarev, D. M.; Babu, N. Ramesh; Parfenov, E. V.

2018-01-01

This research is aimed at improvement of corrosion properties for Zr-1Nb alloy via plasma electrolytic oxidation (PEO). The coatings obtained in DC, pulsed unipolar and pulsed bipolar modes were assessed using SEM, XRD, PDP and EIS techniques. It was shown that pulsed unipolar mode provides the PEO coatings having promising combination of the coating thickness, surface roughness, porosity, corrosion potential and current density, and charge transfer resistance, all contributing to corrosion protection of the zirconium alloy for advanced fuel cladding applications.

Creep behavior of uranium carbide-based alloys

NASA Technical Reports Server (NTRS)

Seltzer, M. S.; Wright, T. R.; Moak, D. P.

1975-01-01

The present work gives the results of experiments on the influence of zirconium carbide and tungsten on the creep properties of uranium carbide. The creep behavior of high-density UC samples follows the classical time-dependence pattern of (1) an instantaneous deformation, (2) a primary creep region, and (3) a period of steady-state creep. Creep rates for unalloyed UC-1.01 and UC-1.05 are several orders of magnitude greater than those measured for carbide alloys containing a Zr-C and/or W dispersoid. The difference in creep strength between alloyed and unalloyed materials varies with temperature and applied stress.

The effect of stem design on the prevalence of squeaking following ceramic-on-ceramic bearing total hip arthroplasty.

PubMed

Restrepo, Camilo; Post, Zachary D; Kai, Brandon; Hozack, William J

2010-03-01

The ceramic-on-ceramic bearing for total hip arthroplasty has an extremely low wear rate and demonstrates minimal inflammatory response in comparison with other bearing choices. However, acoustic emissions such as squeaking and clicking are being reported as annoying complications related to its use. The cause or causes of this phenomenon have not been determined. The purpose of the present study was to evaluate the possibility that design aspects of the femoral component may be a contributing factor to the etiology of squeaking associated with the ceramic-on-ceramic bearing total hip arthroplasty. We retrospectively reviewed 266 consecutive patients (304 hips) who had undergone total hip arthroplasty with use of ceramic-on-ceramic bearings. The first 131 consecutive patients (152 hips) (Group 1) received a hydroxyapatite-coated stem composed of titanium-aluminum-vanadium alloy with a C-taper neck geometry and robust midsection with an anteroposterior diameter of 13 mm. The second 135 consecutive patients (152 hips) (Group 2) also received a hydroxyapatite-coated stem, but in that group the stem was composed of titanium-molybdenum-zirconium-iron alloy, with a V-40 neck geometry and a midsection with an anteroposterior thickness of only 10 mm. All 304 hips received the same cup, composed of titanium-aluminum-vanadium alloy. Demographic characteristics, such as age, sex, height, weight, and body mass index, were similar in both groups. Data regarding the presence of squeaking were obtained prospectively. Patients who were seen for clinical follow-up either expressed the squeaking phenomenon themselves or were asked about it by the physician. Patients who were not seen at a recent clinical follow-up visit were contacted by telephone and were asked specifically about squeaking that might be associated with the hip replacement. Only patients with confirmed squeaking noise were included in the present study. Postoperative radiographs, the Short Form-36 health survey, the Harris hip score, and office or telephone interviews of the patient were used to determine the overall outcome of the procedure. The prevalence of squeaking was seven times higher for patients who received the titanium-molybdenum-zirconium-iron-alloy stem (twenty-seven patients, twenty-eight hips [18.4%]) than in those who received the titanium-aluminum-vanadium-alloy stem (three patients, four hips [2.6%]); this difference was significant (p < 0.0001). Our study suggests that different stem alloys, stem geometries, or neck geometries can have an impact on the frequency of squeaking following a ceramic-on-ceramic total hip arthroplasty.

Accident tolerant fuel cladding development: Promise, status, and challenges

NASA Astrophysics Data System (ADS)

Terrani, Kurt A.

2018-04-01

The motivation for transitioning away from zirconium-based fuel cladding in light water reactors to significantly more oxidation-resistant materials, thereby enhancing safety margins during severe accidents, is laid out. A review of the development status for three accident tolerant fuel cladding technologies, namely coated zirconium-based cladding, ferritic alumina-forming alloy cladding, and silicon carbide fiber-reinforced silicon carbide matrix composite cladding, is offered. Technical challenges and data gaps for each of these cladding technologies are highlighted. Full development towards commercial deployment of these technologies is identified as a high priority for the nuclear industry.

Dynamic Recrystallization Behavior of Zr-1Sn-0.3Nb Alloy During Hot Rolling Process

NASA Astrophysics Data System (ADS)

Zhao, Siyu; Liu, Huiqun; Lin, Gaoyong; Jiang, Yilan; Xun, Jian

2017-11-01

Zirconium alloys are advanced materials with properties that are greatly affected by their crystalline structure. To investigate this, sheets of Zr-1Sn-0.3Nb alloy were hot rolled with different reductions (10%, 30%, 50%, and 60%) at 1023 K and 1073 K to investigate the alloy's dynamic recrystallization behavior. Recrystallization kinetics was observed via electron backscattering diffraction and transmission electron microscopy, and the results were compared with estimates based on the Johnson-Mehl-Avrami-Kolmogorov (JMAK) equation. The values of the JMAK exponent n and k increased with the rolling temperature. The estimates and microstructural observations of dynamic recrystallization (DRX) kinetics were in good agreement.

A comparative study of zirconium and titanium implants in rat: osseointegration and bone material quality.

PubMed

Hoerth, Rebecca M; Katunar, María R; Gomez Sanchez, Andrea; Orellano, Juan C; Ceré, Silvia M; Wagermaier, Wolfgang; Ballarre, Josefina

2014-02-01

Permanent metal implants are widely used in human medical treatments and orthopedics, for example as hip joint replacements. They are commonly made of titanium alloys and beyond the optimization of this established material, it is also essential to explore alternative implant materials in view of improved osseointegration. The aim of our study was to characterize the implant performance of zirconium in comparison to titanium implants. Zirconium implants have been characterized in a previous study concerning material properties and surface characteristics in vitro, such as oxide layer thickness and surface roughness. In the present study, we compare bone material quality around zirconium and titanium implants in terms of osseointegration and therefore characterized bone material properties in a rat model using a multi-method approach. We used light and electron microscopy, micro Raman spectroscopy, micro X-ray fluorescence and X-ray scattering techniques to investigate the osseointegration in terms of compositional and structural properties of the newly formed bone. Regarding the mineralization level, the mineral composition, and the alignment and order of the mineral particles, our results show that the maturity of the newly formed bone after 8 weeks of implantation is already very high. In conclusion, the bone material quality obtained for zirconium implants is at least as good as for titanium. It seems that the zirconium implants can be a good candidate for using as permanent metal prosthesis for orthopedic treatments.

Brazing process provides high-strength bond between aluminum and stainless steel

NASA Technical Reports Server (NTRS)

Huschke, E. G., Jr.; Nord, D. B.

1966-01-01

Brazing process uses vapor-deposited titanium and an aluminum-zirconium-silicon alloy to prevent formation of brittle intermetallic compounds in stainless steel and aluminum bonding. Joints formed by this process maintain their high strength, corrosion resistance, and hermetic sealing properties.

ANALYTICAL METHOD FOR THE DIRECT ABSORPTIOMETRIC DETERMINATION OF BORON IN ZIRCONIUM METAL AND ITS ALLOYS

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1962-01-01

The sample is dissolved in a mixture of ammonium suifate and sulfuric acid. Rosocyanin is formed in the presence of Zr and, after separation from excess curcumin, it is dissolved in ethanol for absorptiometric measurement. (auth)

Imparting passivity to vapor deposited magnesium alloys

NASA Astrophysics Data System (ADS)

Wolfe, Ryan C.

Magnesium has the lowest density of all structural metals. Utilization of low density materials is advantageous from a design standpoint, because lower weight translates into improved performance of engineered products (i.e., notebook computers are more portable, vehicles achieve better gas mileage, and aircraft can carry more payload). Despite their low density and high strength to weight ratio, however, the widespread implementation of magnesium alloys is currently hindered by their relatively poor corrosion resistance. The objective of this research dissertation is to develop a scientific basis for the creation of a corrosion resistant magnesium alloy. The corrosion resistance of magnesium alloys is affected by several interrelated factors. Among these are alloying, microstructure, impurities, galvanic corrosion effects, and service conditions, among others. Alloying and modification of the microstructure are primary approaches to controlling corrosion. Furthermore, nonequilibrium alloying of magnesium via physical vapor deposition allows for the formation of single-phase magnesium alloys with supersaturated concentrations of passivity-enhancing elements. The microstructure and surface morphology is also modifiable during physical vapor deposition through the variation of evaporation power, pressure, temperature, ion bombardment, and the source-to-substrate distance. Aluminum, titanium, yttrium, and zirconium were initially chosen as candidates likely to impart passivity on vapor deposited magnesium alloys. Prior to this research, alloys of this type have never before been produced, much less studied. All of these metals were observed to afford some degree of corrosion resistance to magnesium. Due to the especially promising results from nonequilibrium alloying of magnesium with yttrium and titanium, the ternary magnesium-yttrium-titanium system was investigated in depth. While all of the alloys are lustrous, surface morphology is observed under the scanning electron microscope. The corrosion rate of the nonequilibrium sputtered alloys, as determined by polarization resistance, is significantly reduced compared to the most corrosion resistant commercial magnesium alloys. The open circuit potentials of the sputter deposited alloys are significantly more noble compared to commercial, equilibrium phase magnesium alloys. Galvanic corrosion susceptibility has also been considerably reduced. Nonequilibrium magnesium-yttrium-titanium alloys have been shown to achieve passivity autonomously by alteration of the composition chemistry of the surface oxide/hydroxide layer. Self-healing properties are also evident, as corrosion propagation can be arrested after initial pitting of the material. A clear relationship exists between the corrosion resistance of sputter vapor deposited magnesium alloys and the amount of ion bombardment incurred by the alloy during deposition. Argon pressure, the distance between the source and the substrate, and alloy morphology play important roles in determining the ability of the alloy to develop a passive film. Thermal effects, both during and after alloy deposition, alter the stress state of the alloys, precipitation of second phases, and the mechanical stability of the passive film. An optimal thermal treatment has been developed in order to maximize the corrosion resistance of the magnesium-yttrium-titanium alloys. The significance of the results includes the acquisition of electrochemical data for these novel materials, as well as expanding the utilization of magnesium alloys by the improvement in their corrosion resistance. The magnesium alloys developed in this work are more corrosion resistant than any commercial magnesium alloy. Structural components comprised of these alloys would therefore exhibit unprecedented corrosion performance. Coatings of these alloys on magnesium components would provide a corrosion resistant yet galvanically-compatible coating. The broad impact of these contributions is that these new low-density, corrosion resistant magnesium alloys can be used to produce engineering components for vehicles that have greater acceleration, longer range, heavier payloads, lower life cycle costs, and longer inspection intervals.

Calcium hydride synthesis of Ti-Nb-based alloy powders

NASA Astrophysics Data System (ADS)

Kasimtsev, A. V.; Shuitsev, A. V.; Yudin, S. N.; Levinskii, Yu. V.; Sviridova, T. A.; Alpatov, A. V.; Novosvetlova, E. E.

2017-09-01

The metallothermic (calcium hydride) synthesis of Ti-Nb alloy powders alloyed with tantalum and zirconium is experimentally studied under various conditions. Chemical, X-ray diffraction, and metallographic analyses of the synthesized products show that initial oxides are completely reduced and a homogeneous β-Ti-based alloy powder forms under the optimum synthesis conditions at a temperature of 1200°C. At a lower synthesis temperature, the end products have a high oxygen content. The experimental results are used to plot the thermokinetic dependences o formation of a bcc solid solution at various times of isothermal holding of Ti-22Nb-6Ta and Ti-22Nb-6Zr (at %) alloys. The physicochemical and technological properties of the Ti-22Nb-6Ta and Ti-22Nb-6Zr alloy powders synthesized by calcium hydride reduction under the optimum conditions are determined.

Microgravity

NASA Image and Video Library

2000-07-29

An entranced youngster watches a demonstration of the enhanced resilience of undercooled metal alloys as compared to conventional alloys. Steel bearings are dropped onto plates made of steel, titanium alloy, and zirconium liquid metal alloy, so-called because its molecular structure is amorphous and not crystalline. The bearing on the liquid metal plate bounces for a minute or more longer than on the other plates. Experiments aboard the Space Shuttle helped scientists refine their understanding of the physical properties of certain metal alloys when undercooled (i.e., kept liquid below their normal solidification temperature). This new knowledge then allowed scientists to modify a terrestrial production method so they can now make limited quantities marketed under the Liquid Metal trademark. The exhibit was a part of the NASA outreach activity at AirVenture 2000 sponsored by the Experimental Aircraft Association in Oshkosh, WI.

Sample environment for in situ synchrotron corrosion studies of materials in extreme environments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Elbakhshwan, Mohamed S.; Gill, Simerjeet K.; Motta, Arthur T.

A new in situ sample environment has been designed and developed to study the interfacial interactions of nuclear cladding alloys with high temperature steam. The sample environment is particularly optimized for synchrotron X-ray diffraction (XRD) studies for in situ structural analysis. The sample environment is highly corrosion resistant and can be readily adapted for steam environments. The in situ sample environment design complies with G2 ASTM standards for studying corrosion in zirconium and its alloys and offers remote temperature and pressure monitoring during the in situ data collection. The use of the in situ sample environment is exemplified by monitoringmore » the oxidation of metallic zirconium during exposure to steam at 350°C. Finally, the in situ sample environment provides a powerful tool for fundamental understanding of corrosion mechanisms by elucidating the substoichiometric oxide phases formed during early stages of corrosion, which can provide a better understanding the oxidation process.« less

Sample environment for in situ synchrotron corrosion studies of materials in extreme environments

DOE PAGES

Elbakhshwan, Mohamed S.; Gill, Simerjeet K.; Motta, Arthur T.; ...

2016-10-25

A new in situ sample environment has been designed and developed to study the interfacial interactions of nuclear cladding alloys with high temperature steam. The sample environment is particularly optimized for synchrotron X-ray diffraction (XRD) studies for in situ structural analysis. The sample environment is highly corrosion resistant and can be readily adapted for steam environments. The in situ sample environment design complies with G2 ASTM standards for studying corrosion in zirconium and its alloys and offers remote temperature and pressure monitoring during the in situ data collection. The use of the in situ sample environment is exemplified by monitoringmore » the oxidation of metallic zirconium during exposure to steam at 350°C. Finally, the in situ sample environment provides a powerful tool for fundamental understanding of corrosion mechanisms by elucidating the substoichiometric oxide phases formed during early stages of corrosion, which can provide a better understanding the oxidation process.« less

Investigation and modeling of Al3(Sc, Zr) precipitation strengthening in the presence of enhanced supersaturation and within Al-Cu binary alloys

NASA Astrophysics Data System (ADS)

Deane, Kyle

Diffuse Al-Sc and Al-Zr alloys have been demonstrated in literature to be relatively coarsening resistant at higher temperatures when compared with commonly used precipitation strengthening alloys (e.g. 2000 series, 6000 series). However, because of a limited strengthening due to the low solubility of scandium and zirconium in aluminum, and owing to the scarcity and therefore sizeable price tag attached to scandium, little research has been done in the way of optimizing these alloys for commercial applications. With this in mind, this dissertation describes research which aims to tackle several important areas of Al-Sc-Zr research that have been yet unresolved. In Chapter 4, rapid solidification was utilized to enhance the achievable supersaturation of the alloy in an effort to increase the achievable precipitate strengthening. In Chapter 5, Additive Friction Stir processing (AFS), a novel method of mechanically combining materials without melting, was employed in an attempt to pass the benefits of supersaturation from melt spun ribbon into a more structurally useful bulk material. In Chapter 6, a Matlab program written to predict precipitate nucleation, growth, and coarsening with a modified Kampmann and Wagner Numerical (KWN) model, was used to predict heat treatment regimens for more efficient strengthening. Those predictions were then tested experimentally to test the validity of the results. And lastly, in Chapter 7, the effect of zirconium on Al-Cu secondary precipitates was studied in an attempt to increase their thermal stability, as much higher phase fractions of Al-Cu precipitates are achievable than Al-Zr precipitates.

NEUTRONIC REACTOR CONTROL ROD AND METHOD OF FABRICATION

DOEpatents

Porembka, S.W. Jr.

1961-06-27

A reactor control rod formed from a compacted powder dispersion is patented. The rod consists of titanium sheathed with a cladding alloy. The cladding alloy contains 1.3% to 1.6% by weight of tin, 0.07% to 0.12% by weight of chromium, 0.04% to 0.08% by weight of nickel, 0.09% to 0.16% by weight of iron, carbon not exceeding 0.05%, less than 0.5% by weight of incidental impurities, and the balance zirconium.

Five year survival analysis of an oxidised zirconium total knee arthroplasty.

PubMed

Holland, Philip; Santini, Alasdair J A; Davidson, John S; Pope, Jill A

2013-12-01

Zirconium total knee arthroplasties theoretically have a low incidence of failure as they are low friction, hard wearing and hypoallergenic. We report the five year survival of 213 Profix zirconium total knee arthroplasties with a conforming all polyethylene tibial component. Data was collected prospectively and multiple strict end points were used. SF12 and WOMAC scores were recorded pre-operatively, at three months, at twelve months, at 3 years and at 5 years. Eight patients died and six were "lost to follow-up". The remaining 199 knees were followed up for five years. The mean WOMAC score improved from 56 to 35 and the mean SF12 physical component score improved from 28 to 34. The five year survival for failure due to implant related reasons was 99.5% (95% CI 97.4-100). This was due to one tibial component becoming loose aseptically in year zero. Our results demonstrate that the Profix zirconium total knee arthroplasty has a low medium term failure rate comparable to the best implants. Further research is needed to establish if the beneficial properties of zirconium improve long term implant survival. Copyright © 2012 Elsevier B.V. All rights reserved.

Self-repairing vanadium-zirconium composite conversion coating for aluminum alloys

NASA Astrophysics Data System (ADS)

Zhong, Xin; Wu, Xiaosong; Jia, Yuyu; Liu, Yali

2013-09-01

In this paper, new self-repairing vanadium-zirconium composite conversion coating was prepared and investigated by Electrochemical impedance spectra (EIS), Scanning electron microscope (SEM) and X-ray photoelectron spectroscopy (XPS), respectively. EIS results showed that V-Zr conversion coating with hydrogen peroxide modified (VZO) revealed an increasing corrosion resistance in corrosive media which meant a certain self-repairing effect. SEM comparison photos also disclosed that VZO treated with scratches was gradually ameliorated from the initial cracked configuration to fewer cracks and more fillers through an immersion of 3.5% NaCl solution. XPS results demonstrated that the content of vanadium on VZO increased and zirconium declined when immersed in the corrosive solution. This explained further that the self-repairing ability could be related to vanadium. From the above results, we inferred possible structures of VZO and proposed that self-repairing effect was achieved through a hydrolysis condensation polymerization process of vanadate in the localized corrosion area.

Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Montgomery, Robert, E-mail: robert.montgomery@pnnl.gov; Tomé, Carlos, E-mail: tome@lanl.gov; Liu, Wenfeng, E-mail: wenfeng.liu@anatech.com

Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical constitutive models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. To improve upon this approach, a microstructurally-based zirconium alloy mechanical deformation analysis capability is being developed within the United States Department of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL). Specifically, the viscoplastic self-consistent (VPSC)more » polycrystal plasticity modeling approach, developed by Lebensohn and Tomé [1], has been coupled with the BISON engineering scale fuel performance code to represent the mechanistic material processes controlling the deformation behavior of light water reactor (LWR) cladding. A critical component of VPSC is the representation of the crystallographic nature (defect and dislocation movement) and orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. A future goal is for VPSC to obtain information on reaction rate kinetics from atomistic calculations to inform the defect and dislocation behavior models described in VPSC. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON and provides initial results utilizing the coupled functionality.« less

In vitro and in vivo biological performance of porous Ti alloys prepared by powder metallurgy

PubMed Central

Vasconcellos, Luis Gustavo Oliveira De; Oliveira, Felipe Eduardo; Oliveira, Luciane Dias De; Henriques, Vinicius André Rodrigues; Carvalho, Yasmin Rodarte; De Vasconcellos, Luana Marotta Reis

2018-01-01

Titanium (Ti) and Ti-6 Aluminium-4 Vanadium alloys are the most common materials in implants composition but β type alloys are promising biomaterials because they present better mechanical properties. Besides the composition of biomaterial, many factors influence the performance of the biomaterial. For example, porous surface may modify the functional cellular response and accelerate osseointegration. This paper presents in vitro and in vivo evaluations of powder metallurgy-processed porous samples composed by different titanium alloys and pure Ti, aiming to show their potential for biomedical applications. The porous surfaces samples were produced with different designs to in vitro and in vivo tests. Samples were characterized with scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS) and elastic modulus analyses. Osteogenic cells from newborn rat calvaria were plated on discs of different materials: G1—commercially pure Ti group (CpTi); G2—Ti-6Al-4V alloy; G3—Ti-13 Niobium-13 Zirconium alloy; G4—Ti-35 Niobium alloy; G5—Ti-35 Niobium-7 Zirconium-5 Tantalum alloy. Cell adhesion and viability, total protein content, alkaline phosphatase activity, mineralization nodules and gene expression (alkaline phosphatase, Runx-2, osteocalcin and osteopontin) were assessed. After 2 and 4 weeks of implantation in rabbit tibia, bone ingrowth was analyzed using micro-computed tomography (μCT). EDS analysis confirmed the material production of each group. Metallographic and SEM analysis revealed interconnected pores, with mean pore size of 99,5μm and mean porosity of 42%, without significant difference among the groups (p>0.05). The elastic modulus values did not exhibit difference among the groups (p>0.05). Experimental alloys demonstrated better results than CpTi and Ti-6Al-4V, in gene expression and cytokines analysis, especially in early experimental periods. In conclusion, our data suggests that the experimental alloys can be used for biomedical application since they contributed to excellent cellular behavior and osseointegration besides presenting lower elastic modulus. PMID:29771925

Iron-nickel-chromium alloy having improved swelling resistance and low neutron absorbence

DOEpatents

Korenko, Michael K.

1986-01-01

An iron-nickel-chromium age-hardenable alloy suitable for use in fast breeder reactor ducts and cladding which utilizes the gamma-double prime strengthening phase and characterized in having a delta or eta phase distributed at or near grain boundaries. The alloy consists essentially of about 33-39.5% nickel, 7.5-16% chromium, 1.5-4% niobium, 0.1-0.7% silicon, 0.01-0.2% zirconium, 1-3% titanium, 0.2-0.6% aluminum, and the remainder essentially all iron. Up to 0.4% manganese and up to 0.010% magnesium can be added to inhibit trace element effects.

Microstructural analysis of biodegradable Mg-0.9Ca-1.2Zr alloy

NASA Astrophysics Data System (ADS)

Istrate, B.; Munteanu, C.; Geanta, V.; Baltatu, S.; Focsaneanu, S.; Earar, K.

2016-08-01

Magnesium alloys have applications in aerospace and medical applications as biodegradable orthopedic implants. Alloying with biocompatible elements, such as calcium or zirconium contribute to refining the the microstructure and improves corrosion resistance with the formation of an eutectic compound - Mg2Ca at boundary alpha-Mg grains. The purpose of this paper is to present the microstructure throw optical and scanning electron methods and phase and constituents identification with X-ray analysis. The results showed the presence of alpha-Mg grains with formation of a mechanical compound - Mg2Ca and appearance of alpha- Zr phase relatively uniformly distributed in nests.

Investigation of welding and brazing of molybdenum and TZM alloy tubes

NASA Technical Reports Server (NTRS)

Lundblad, Wayne E.

1991-01-01

This effort involved investigating the welding and brazing techniques of molybdenum tubes to be used as cartridges in the crystal growth cartridge. Information is given in the form of charts and photomicrographs. It was found that the recrystallization temperature of molybdenum can be increased by alloying it with 0.5 percent titanium and 0.1 percent zirconium. Recrystallization temperatures for this alloy, known as TZM, become significant around 2500 F. A series of microhardness tests were run on samples of virgin and heat soaked TZM. The test results are given in tabular form. It was concluded that powder metallurgy TZM may be an acceptable cartridge material.

Influence of oxide microstructure on corrosion behavior of zirconium-based model alloys

NASA Astrophysics Data System (ADS)

Silva, Marcelo Jose Gomes Da

The extensive utilization of zirconium-based alloys in fuel cladding and other reactor internal components in the nuclear power industry has led to the continuous improvement of these alloys. At the present moment, demands for better performing nuclear fuel cladding materials are increasing. Also, new reactor designs have been proposed that would require the materials to withstand even more rigorous conditions. One of the factors that limit s fuel cladding utilization in nuclear reactors is uniform corrosion and the consequent hydriding of the fuel. In an attempt to develop mechanistic understanding of the role of alloying elements in the growth of a stable protective oxide, a series of model zirconium-based alloys was prepared (Zr-xFe-yCr, Zr-xCu-yMo, Zr-xNb-ySn, for various x and y, pure Zr and Zircaloy-4) and examined with advanced characterization techniques. The alloys were corrosion tested in autoclaves under three different conditions: 360°C water, 500°C steam and 500°C supercritical water in excess of 400 days. These autoclave testing conditions simulate nuclear reactor environment for both current designs (360°C water) and the new supercritical water reactor (500°C steam and 500°C supercritical water) proposed by the generation-IV initiative. The oxide films formed were systematically examined at the Advanced Photon Source using microbeam synchrotron radiation diffraction and fluorescence of cross-sectional samples to determine the oxide phases present and their crystallographic texture as a function of distance from the metal/oxide interface. Also, the overall texture of the oxide layers was investigated using synchrotron radiation diffraction in frontal geometry. The corrosion kinetics is a function of the alloy system and showed a wide range of behaviors, from immediately unstable oxide growth to stable behavior. The corrosion weight gains from testing at high temperature are a factor of five higher than those measured at 360°C but the protectiveness ranking of the alloys is similar. Measured pole figures from different oxides in different corrosion regimes showed that monoclinic oxides grow in slightly distinct directions: protective oxides grow along the (-904)m pole, whether non-protective oxides grow along or close to the (-302)m pole. The angle in between these two directions ((-904)m and (-302)m) is about 6°. Microbeam synchrotron radiation diffraction and fluorescence was performed in the oxide layers and systematic differences are observed in protective and non-protective oxides, both near the oxide/metal interface and in the bulk of the oxide layers. The non-protective oxide interfaces show a smooth transition from metal to oxide with metal diffraction peaks disappearing as the monoclinic oxide peaks appear. In contrast, in a protective oxide, a complex structure near the oxide/metal interface was seen, showing peaks from Zr 3O suboxide and a highly oriented tetragonal oxide phase with specific orientation relationships with the monoclinic oxide and the base metal. The highly oriented tetragonal phase, only present in protective oxides, is believed to be a precursor to the formation of monoclinic oxide found in the bulk of the oxide layer. This plane may promote stable growth by causing the oxide to form in a manner that maximizes occupation of the substrate surface and minimizes stress accumulation, leading to more stable oxide growth. The association seen in this work of the precursor oxide phase with protective oxides and its orientation relationship with the monoclinic oxide, combined with the difference in oxide growth direction seen between protective and non-protective oxides, is interpreted as evidence that this phase allows a more properly oriented oxide to grow, in a way that minimizes stress accumulation and therefore delays the oxide transition to larger oxide thicknesses.

Method of increasing the phase stability and the compressive yield strength of uranium-1 to 3 wt. % zirconium alloy

DOEpatents

Anderson, Robert C.

1986-01-01

A uranium-1 to 3 wt. % zirconium alloy characterized by high strength, high ductility and stable microstructure is fabricated by an improved thermal mechanical process. A homogenous ingot of the alloy which has been reduced in thickness of at least 50% in the two-step forging operation, rolled into a plate with a 75% reduction and then heated in vacuum at a temperature of about 750.degree. to 850.degree. C. and then quenched in water is subjected to further thermal-mechanical operation steps to increase the compressive yield strength approximately 30%, stabilize the microstructure, and decrease the variations in mechanical properties throughout the plate is provided. These thermal-mechanical steps are achieved by cold rolling the quenched plate to reduce the thickness thereof about 8 to 12%, aging the cold rolled plate at a first temperature of about 325.degree. to 375.degree. C. for five to six hours and then aging the plate at a higher temperature ranging from 480.degree. to 500.degree. C. for five to six hours prior to cooling the billet to ambient conditions and sizing the billet or plate into articles provides the desired increase in mechanical properties and phase stability throughout the plate.

Calcium and zirconium as texture modifiers during rolling and annealing of magnesium–zinc alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bohlen, Jan, E-mail: jan.bohlen@hzg.de; Wendt, Joachim; Nienaber, Maria

2015-03-15

Rolling experiments were carried out on a ternary Mg–Zn–Ca alloy and its modification with zirconium. Short time annealing of as-rolled sheets is used to reveal the microstructure and texture development. The texture of the as-rolled sheets can be characterised by basal pole figures with split peak towards the rolling direction (RD) and a broad transverse angular spread of basal planes towards the transverse direction (TD). During annealing the RD split peaks as well as orientations in the sheet plane vanish whereas the distribution of orientations tilted towards the TD remains. It is shown in EBSD measurements that during rolling bandsmore » of twin containing structures form. During subsequent annealing basal orientations close to the sheet plane vanish based on a grain nucleation and growth mechanism of recrystallisation. Orientations with tilt towards the TD remain in grains that do not undergo such a mechanism. The addition of Zr delays texture weakening. - Highlights: • Ca in Mg–Zn-alloys contributes to a significant texture weakening during rolling and annealing. • Grain nucleation and growth in structures consisting of twins explain a texture randomisation during annealing. • Grains with transverse tilt of basal planes preferentially do not undergo a grain nucleation and growth mechanism. • Zr delays the microstructure and texture development.« less

AN ATTEMPT TO LOCATE INTERMETALLIC PARTICLES IN ZIRCONIUM ALLOYS USING A BITTER FIGURE TECHNIQUE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cox, B.; Harder, B.R.

1961-10-01

The compound ZrFe/sub 2/ is known to be ferromagnetic, and an attempt to locate particles of magnetic material in zircaloy-2 and dilute Zr- Fe alloys by a Bitter figure technlque is described. An Fe/sub 3/O/sub 4/ sol in water-soluble plastic was used to prepare Bitter figures of the alloy surfaces in the form of replicas, which were then examined in an electron microscope. No magnetic particles were located in either zircaloy-2 or a Zr-O.3% Fe alloy. Subsequent work on specimens of ZrFe/sub 2/ showed that the failure to detect it in the dilute alloys arose because the size of themore » intermetallic particles in the latter was smaller than the size of the magnetic domains. (auth)« less

Space Research Benefits Demonstrated

NASA Technical Reports Server (NTRS)

2000-01-01

An entranced youngster watches a demonstration of the enhanced resilience of undercooled metal alloys as compared to conventional alloys. Steel bearings are dropped onto plates made of steel, titanium alloy, and zirconium liquid metal alloy, so-called because its molecular structure is amorphous and not crystalline. The bearing on the liquid metal plate bounces for a minute or more longer than on the other plates. Experiments aboard the Space Shuttle helped scientists refine their understanding of the physical properties of certain metal alloys when undercooled (i.e., kept liquid below their normal solidification temperature). This new knowledge then allowed scientists to modify a terrestrial production method so they can now make limited quantities marketed under the Liquid Metal trademark. The exhibit was a part of the NASA outreach activity at AirVenture 2000 sponsored by the Experimental Aircraft Association in Oshkosh, WI.

The neutron texture diffractometer at the China Advanced Research Reactor

NASA Astrophysics Data System (ADS)

Li, Mei-Juan; Liu, Xiao-Long; Liu, Yun-Tao; Tian, Geng-Fang; Gao, Jian-Bo; Yu, Zhou-Xiang; Li, Yu-Qing; Wu, Li-Qi; Yang, Lin-Feng; Sun, Kai; Wang, Hong-Li; Santisteban, J. r.; Chen, Dong-Feng

2016-03-01

The first neutron texture diffractometer in China has been built at the China Advanced Research Reactor, due to strong demand for texture measurement with neutrons from the domestic user community. This neutron texture diffractometer has high neutron intensity, moderate resolution and is mainly applied to study texture in commonly used industrial materials and engineering components. In this paper, the design and characteristics of this instrument are described. The results for calibration with neutrons and quantitative texture analysis of zirconium alloy plate are presented. The comparison of texture measurements with the results obtained in HIPPO at LANSCE and Kowari at ANSTO illustrates the reliability of the texture diffractometer. Supported by National Nature Science Foundation of China (11105231, 11205248, 51327902) and International Atomic Energy Agency-TC program (CPR0012)

Around Marshall

NASA Image and Video Library

2003-01-01

This is a close-up of a sample of titanium-zirconium-nickel alloy inside the Electrostatic Levitator (ESL) vacuum chamber at NASA's Marshall Space Flight Center (MSFC). The ESL uses static electricity to suspend an object (about 3-4 mm in diameter) inside a vacuum chamber allowing scientists to record a wide range of physical properties without the sample contracting the container or any instruments, conditions that would alter the readings. Once inside the chamber, a laser heats the sample until it melts. The laser is then turned off and the sample cools, changing from a liquid drop to a solid sphere. Since 1977, the ESL has been used at MSFC to study the characteristics of new metals, ceramics, and glass compounds. Materials created as a result of these tests include new optical materials, special metallic glasses, and spacecraft components.

Total knee arthroplasty with an oxidised zirconium femoral component: ten-year survivorship analysis.

PubMed

Ahmed, I; Salmon, L J; Waller, A; Watanabe, H; Roe, J P; Pinczewski, L A

2016-01-01

Oxidised zirconium was introduced as a material for femoral components in total knee arthroplasty (TKA) as an attempt to reduce polyethylene wear. However, the long-term survival of this component is not known. We performed a retrospective review of a prospectively collected database to assess the ten year survival and clinical and radiological outcomes of an oxidised zirconium total knee arthroplasty with the Genesis II prosthesis. The Western Ontario and McMaster Universities Osteoarthritis Index (WOMAC), Knee Injury and Osteoarthritis Outcome Score (KOOS) and a patient satisfaction scale were used to assess outcome. A total of 303 consecutive TKAs were performed in 278 patients with a mean age of 68 years (45 to 89). The rate of survival ten years post-operatively as assessed using Kaplan-Meier analysis was 97% (95% confidence interval 94 to 99) with revision for any reason as the endpoint. There were no revisions for loosening, osteolysis or failure of the implant. There was a significant improvement in all components of the WOMAC score at final follow-up (p < 0.001). The mean individual components of the KOOS score for symptoms (82.4 points; 36 to 100), pain (87.5 points; 6 to 100), activities of daily life (84.9 points; 15 to 100) and quality of life (71.4 points; 6 to 100) were all at higher end of the scale. This study provides further supportive evidence that the oxidised zirconium TKA gives comparable rates of survival with other implants and excellent functional outcomes ten years post-operatively. Total knee arthroplasty with an oxidised zirconium femoral component gives comparable long-term rates of survival and functional outcomes with conventional implants. ©2016 The British Editorial Society of Bone & Joint Surgery.

Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

McDeavitt, Sean M

2011-04-29

Overview Fast reactors were evaluated to enable the transmutation of transuranic isotopes generated by nuclear energy systems. The motivation for this was that TRU isotopes have high radiotoxicity and relatively long half-lives, making them unattractive for disposal in a long-term geologic repository. Fast reactors provide an efficient means to utilize the energy content of the TRUs while destroying them. An enabling technology that requires research and development is the fabrication metallic fuel containing TRU isotopes using powder metallurgy methods. This project focused upon developing a powder metallurgical fabrication method to produce U-Zr-transuranic (TRU) alloys at relatively low processing temperatures (500ºCmore » to 600ºC) using either hot extrusion or alpha-phase sintering for charecterization. Researchers quantified the fundamental aspects of both processing methods using surrogate metals to simulate the TRU elements. The process produced novel solutions to some of the issues relating to metallic fuels, such as fuel-cladding chemical interactions, fuel swelling, volatility losses during casting, and casting mold material losses. Workscope There were two primary tasks associated with this project: 1. Hot working fabrication using mechanical alloying and extrusion • Design, fabricate, and assemble extrusion equipment • Extrusion database on DU metal • Extrusion database on U-10Zr alloys • Extrusion database on U-20xx-10Zr alloys • Evaluation and testing of tube sheath metals 2. Low-temperature sintering of U alloys • Design, fabricate, and assemble equipment • Sintering database on DU metal • Sintering database on U-10Zr alloys • Liquid assisted phase sintering on U-20xx-10Zr alloys Appendices Outline Appendix A contains a Fuel Cycle Research & Development (FCR&D) poster and contact presentation where TAMU made primary contributions. Appendix B contains MSNE theses and final defense presentations by David Garnetti and Grant Helmreich outlining the beginning of the materials processing setup. Also included within this section is a thesis proposal by Jeff Hausaman. Appendix C contains the public papers and presentations introduced at the 2010 American Nuclear Society Winter Meeting. Appendix A—MSNE theses of David Garnetti and Grant Helmreich and proposal by Jeff Hausaman A.1 December 2009 Thesis by David Garnetti entitled “Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.2 September 2009 Presentation by David Garnetti (same title as document in Appendix B.1) A.3 December 2010 Thesis by Grant Helmreich entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.4 October 2010 Presentation by Grant Helmreich (same title as document in Appendix B.3) A.5 Thesis Proposal by Jeffrey Hausaman entitled “Hot Extrusion of Alpha Phase Uranium-Zirconium Alloys for TRU Burning Fast Reactors” Appendix B—External presentations introduced at the 2010 ANS Winter Meeting B.1 J.S. Hausaman, D.J. Garnetti, and S.M. McDeavitt, “Powder Metallurgy of Alpha Phase Uranium Alloys for TRU Burning Fast Reactors,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.2 PowerPoint Presentation Slides from C.1 B.3 G.W. Helmreich, W.J. Sames, D.J. Garnetti, and S.M. McDeavitt, “Uranium Powder Production Using a Hydride-Dehydride Process,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.4. PowerPoint Presentation Slides from C.3 B.5 Poster Presentation from C.3 Appendix C—Fuel cycle research and development undergraduate materials and poster presentation C.1 Poster entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys” presented at the Fuel Cycle Technologies Program Annual Meeting C.2 April 2011 Honors Undergraduate Thesis by William Sames, Research Fellow, entitled “Uranium Metal Powder Production, Particle Distribution Analysis, and Reaction Rate Studies of a Hydride-Dehydride Process"« less

Solid solution strengthened duct and cladding alloy D9-B1

DOEpatents

Korenko, Michael K.

1983-01-01

A modified AISI type 316 stainless steel is described for use in an atmosphere where the alloy will be subject to neutron irradiation. The alloy is characterized by its phase stability in both the annealed as well as cold work condition and above all by its superior resistance to radiation induced swelling. Graphical data is included to demonstrate the superior swelling resistance of the alloy which contains from about 0.5% to 2.2% manganese, from about 0.7% to about 1.1% silicon, from about 12.5% to 14% chromium, from about 14.5% to about 16.5% nickel, from about 1.2% to about 1.6% molybdenum, from 0.15% to 0.30% titanium, from 0.02% to 0.08% zirconium, and the balance iron with incidental impurities.

Some observations on uranium carbide alloy/tungsten compatibility

NASA Technical Reports Server (NTRS)

Phillips, W. M.

1972-01-01

Chemical compatibility between both pure and thoriated tungsten and uranium carbide alloys was studied at 1800 C for up to 3300 hours. Alloying with zirconium carbide appeared to widen the homogeneity range of uranium carbide, making additional carbon available for reaction with the tungsten. Reaction layers were formed both by vapor phase reaction and by physical contact, producing either or both UWC2 and W2C, dependent upon the phases present in the starting fuel alloy. Formation of UWC2 results in slow growth of the reaction layer with time, while W2C reaction layers grow rapidly, allowing equilibrium to be reached in less than 2500 hours at 1800 C. The presence of a thermal gradient had no effect on the reactions observed nor did the presence of thoria in the tungsten clad.

Some observations on uranium carbide alloy/tungsten compatibility.

NASA Technical Reports Server (NTRS)

Phillips, W. M.

1972-01-01

Results of chemical compatibility tests between both pure tungsten and thoriated tungsten run at 1800 C for up to 3300 hours with uranium carbide alloys. Alloying with zirconium carbide appeared to widen the homogeneity range of uranium carbide, making additional carbon available for reaction with the tungsten. Reaction layers were formed both by vapor phase reaction and by physical contact, producing either or both UWC2 and W2C, depending upon the phases present in the starting fuel alloy. Formation of UWC2 results in slow growth of the reaction layer with time, while W2C reaction layers grow rapidly, allowing equilibrium to be reached in less than 2500 hours at 1800 C. Neither the presence of a thermal gradient nor the presence of thoria in the tungsten clad affect the reactions observed.

An experimental study of steam explosions involving chemically reactive metal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cho, D.H.; Armstrong, D.R.; Gunther, W.H.

1997-07-01

An experimental study of molten zirconium-water explosions was conducted. A 1-kg mass of zirconium melt was dropped into a column of water. Explosions took place only when an external trigger was used. In the triggered tests, the extent of oxidation of the zirconium melt was very extensive. However, the explosion energetics estimated were found to be very small compared to the potential chemical energy available from the oxidation reaction. Zirconium is of particular interest, since it is a component of the core materials of the current nuclear power reactors. This paper describes the test apparatus and summarizes the results ofmore » four tests conducted using pure zirconium melt.« less

Zirconium nitride precipitation in nominally pure yttria-stabilized zirconia

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gomez-Garcia, D.; Martinez-Fernandez, J.; Dominguez-Rodriguez, A.

Nominally pure yttria-stabilized zirconia alloys are shown to contain unexpectedly large amounts of dissolved nitrogen. Its presence in the lattice was detected through the observation of large precipitates in alloys with three different concentrations of yttria deformed in compression in argon in the temperature range 1,600--1,800 C. Electron diffraction, EDS and PEELS analyses, and Moire imaging were used to identify the precipitates as ZrN. The possible origin of the nitrogen, its likely effects on properties, and the role of annealing atmosphere are briefly discussed.

Analysis of Operating Strategies Using Different Target Designs For 238Pu Production

DOE Office of Scientific and Technical Information (OSTI.GOV)

Thomas, Tomcy; Sherman, Steven R; Sawhney, Dr. Rapinder

2017-01-01

An engineering effort is underway to re-establish capability to produce 238Pu oxide at the kilogram scale in the United States. A multi-step batch process is being developed to produce this important material. Recently, a portion of this process was studied using discrete-event simulation tools to determine whether the conceptual process might achieve its yearly production goal. The study showed the conceptual process can meet the yearly production goal under some circumstances, but process improvements would be needed to ensure greater likelihood of success. This study extends the work performed previously by examining the effects of changing the reactor target designmore » on the yearly process output. Two new reactor target configurations are considered an aluminum-clad reactor target containing 50% greater 237Np oxide content than the original target, and a zirconium alloy-clad target using no aluminum. The results indicate that use of the new aluminum-clad target configuration may allow the process to achieve the same yearly production goal in less time using fewer targets. If the zirconium alloy-clad target is used, then even fewer targets would be needed to reach the production goal, but some process changes would be required to handle the zirconium cladding. The number of days needed to process a target batch to completion, and the steady state 238Pu oxide production rate, for each configuration are compared to the results from the initial simulation study.« less

Ceramic material suitable for repair of a space vehicle component in a microgravity and vacuum environment, method of making same, and method of repairing a space vehicle component

NASA Technical Reports Server (NTRS)

Riedell, James A. (Inventor); Easler, Timothy E. (Inventor)

2009-01-01

A precursor of a ceramic adhesive suitable for use in a vacuum, thermal, and microgravity environment. The precursor of the ceramic adhesive includes a silicon-based, preceramic polymer and at least one ceramic powder selected from the group consisting of aluminum oxide, aluminum nitride, boron carbide, boron oxide, boron nitride, hafnium boride, hafnium carbide, hafnium oxide, lithium aluminate, molybdenum silicide, niobium carbide, niobium nitride, silicon boride, silicon carbide, silicon oxide, silicon nitride, tin oxide, tantalum boride, tantalum carbide, tantalum oxide, tantalum nitride, titanium boride, titanium carbide, titanium oxide, titanium nitride, yttrium oxide, zirconium diboride, zirconium carbide, zirconium oxide, and zirconium silicate. Methods of forming the ceramic adhesive and of repairing a substrate in a vacuum and microgravity environment are also disclosed, as is a substrate repaired with the ceramic adhesive.

Aluminum-Silicon Alloy Having Improved Properties at Elevated Temperatures and Articles Cast Therefrom

NASA Technical Reports Server (NTRS)

Lee, Jonathan A. (Inventor); Chen, Po-Shou (Inventor)

2002-01-01

An aluminum alloy suitable for high temperature applications, such as heavy duty pistons and other internal combustion applications. having the following composition, by weight percent (wt %): Silicon: 11.0-14.0; Copper: 5.6-8.0; Iron: 0-0.8; Magnesium: 0.5-1.5; Nickel: 0.05-0.9; Manganese: 0.5-1.5; Titanium: 0.05-1.2; Zirconium: 0.12-1.2; Vanadium: 0.05-1.2; Zinc: 0.005-0.9; Strontium: 0.001-0.1; Aluminum: balance. In this alloy the ratio of silicon:magnesium is 10-25, and the ratio of copper:magnesium is 4-15. After an article is cast from this alloy, the article is treated in a solutionizing step which dissolves unwanted precipitates and reduces any segregation present in the original alloy. After this solutionizing step, the article is quenched, and is then aged at an elevated temperature for maximum strength.

Cytotoxicity of titanium and titanium alloying elements.

PubMed

Li, Y; Wong, C; Xiong, J; Hodgson, P; Wen, C

2010-05-01

It is commonly accepted that titanium and the titanium alloying elements of tantalum, niobium, zirconium, molybdenum, tin, and silicon are biocompatible. However, our research in the development of new titanium alloys for biomedical applications indicated that some titanium alloys containing molybdenum, niobium, and silicon produced by powder metallurgy show a certain degree of cytotoxicity. We hypothesized that the cytotoxicity is linked to the ion release from the metals. To prove this hypothesis, we assessed the cytotoxicity of titanium and titanium alloying elements in both forms of powder and bulk, using osteoblast-like SaOS(2) cells. Results indicated that the metal powders of titanium, niobium, molybdenum, and silicon are cytotoxic, and the bulk metals of silicon and molybdenum also showed cytotoxicity. Meanwhile, we established that the safe ion concentrations (below which the ion concentration is non-toxic) are 8.5, 15.5, 172.0, and 37,000.0 microg/L for molybdenum, titanium, niobium, and silicon, respectively.

Effects of cobalt, boron, and zirconium on the microstructure of Udimet 738. M.S. Thesis. Final Report

NASA Technical Reports Server (NTRS)

Nakanishi, T. G.

1984-01-01

A structural study was carried out on Co modified Udimet 738 alloys containing 0.04, 0.10, and 0.20 wt % Zr at 0.01 and 0.03 wt % B levels. Samples in the as-cast and solution-treated conditions were exposed at 843 C to study structural stability. The structures produced by the interactions of Co, Zr, and B were studied by SEM, X-ray diffraction, and dispersive analysis techniques. The additions of large amounts of Zr and B were found to increase the solidification range of the U-738. Structural changes involved eutectic gamma prime islands, formation of low melting point compounds, and precipitation of borides and Zr rich phases. Boron and zirconium additions did not show substantial changes in mechanical properties. Removal of Co from the alloys resulted in reduction of the matrix solubility for carbon and increase in the gamma prime solvus. Structural instabilities found were continuous grain boundary M23C6 films, MC breakdown, and plate-like phases. Removal of cobalt resulted in a slight decrease in tensile and stress rupture properties. Detailed structural results presented.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Young-Ho; Byun, Thak Sang

Accident-tolerant fuels are expected to have considerably longer coping time to respond to the loss of active cooling under severe accidents and, at the same time, have comparable or improved fuel performance during normal operation. The wear resistance of accident tolerant fuels, therefore, needs to be examined to determine the applicability of these cladding candidates to the current operating PWRs because the most common failure of nuclear fuel claddings is still caused by grid-to-rod fretting during normal operations. In this study, reciprocating sliding wear tests on three kinds of cladding candidates for accident-tolerant fuels have been performed to investigate themore » tribological compatibilities of selfmated cladding candidates and to determine the direct applicability of conventional Zirconium-based alloys as supporting structural materials. The friction coefficients of the cladding candidates are strongly influenced by the test environments and coupled materials. The wear test results under water lubrication conditions indicate that the supporting structural materials for the cladding candidates of accident-tolerant fuels need to be replaced with the same cladding materials instead of using conventional Zirconium-based alloys.« less

High temperature, low cycle fatigue of copper-base alloys in argon. Part 3: Zirconium-copper; thermal-mechanical strain cycling, hold-time and notch fatigue results

NASA Technical Reports Server (NTRS)

Conway, J. B.; Stentz, R. H.; Berling, J. T.

1973-01-01

The low-cycle fatigue characteristics of smooth bar and notched bar specimens (hourglass shape) of zirconium-copper, 1/2 Hard, material (R-2 Series) were evaluated at room temperature in axial strain control. Over the fatigue life range from about 300 to 3000 cycles the ratio of fatigue life for smooth bar to fatigue life for notched bar remained constant at a value of about 6.0. Some additional hold-time data for the R-2 alloy tested in argon at 538 C are reported. An analysis of the relaxation data obtained in these hold-time tests is also reported and it is shown that these data yield a fairly consistent correlation in terms of instantaneous stress rate divided by instantaneous stress. Two thermal-mechanical strain cycling tests were also performed using a cyclic frequency of 4.5 cycles per hour and a temperature cycling interval from 260 to 538 C. The fatigue life values in these tests were noticeably lower than that observed in isothermal tests at 538 C.

Transport property correlations for the niobium-1% zirconium alloy

NASA Astrophysics Data System (ADS)

Senor, David J.; Thomas, J. Kelly; Peddicord, K. L.

1990-10-01

Correlations were developed for the electrical resistivity (ρ), thermal conductivity ( k), and hemispherical total emittance (ɛ) of niobium-1% zirconium as functions of temperature. All three correlations were developed as empirical fits to experimental data. ρ = 5.571 + 4.160 × 10 -2(T) - 4.192 × 10 -6(T) 2 μΩcm , k = 13.16( T) 0.2149W/ mK, ɛ = 6.39 × 10 -2 + 4.98 × 10 -5( T) + 3.62 × 10 -8( T) 2 - 7.28 × 10 -12( T) 3. The relative standard deviation of the electrical resistivity correlation is 1.72% and it is valid over the temperature range 273 to 2700 K. The thermal conductivity correlation has a relative standard deviation of 3.24% and is valid over the temperature range 379 to 1421 K. The hemispherical total emittance correlation was developed for smooth surface materials only and represents a conservative estimate of the emittance of the alloy for space reactor fuel element modeling applications. It has a relative standard deviation of 9.50% and is valid over the temperature range 755 to 2670 K.

Corrosion Assessment of Candidate Materials for the SHINE Subcritical Assembly Vessel and Components FY14 Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pawel, Steven J.

2014-10-01

Laboratory corrosion testing of candidate alloys—including Zr-4 and Zr-2.5Nb representing the target solution vessel, and 316L, 2304, 304L, and 17-4 PH stainless steels representing process piping and balance-of-plant components—was performed in support of the proposed SHINE process to produce 99Mo from low-enriched uranium. The test solutions used depleted uranyl sulfate in various concentrations and incorporated a range of temperatures, excess sulfuric acid concentrations, nitric acid additions (to simulate radiolysis product generation), and iodine additions. Testing involved static immersion of coupons in solution and in the vapor above the solution, and was extended to include planned-interval tests to examine details associatedmore » with stainless steel corrosion in environments containing iodine species. A large number of galvanic tests featuring couples between a stainless steel and a zirconium-based alloy were performed, and limited vibratory horn testing was incorporated to explore potential erosion/corrosion features of compatibility. In all cases, corrosion of the zirconium alloys was observed to be minimal, with corrosion rates based on weight loss calculated to be less than 0.1 mil/year with no change in surface roughness. The resulting passive film appeared to be ZrO2 with variations in thickness that influence apparent coloration (toward light brown for thicker films). Galvanic coupling with various stainless steels in selected exposures had no discernable effect on appearance, surface roughness, or corrosion rate. Erosion/corrosion behavior was the same for zirconium alloys in uranyl sulfate solutions and in sodium sulfate solutions adjusted to a similar pH, suggesting there was no negative effect of uranium resulting from fluid dynamic conditions aggressive to the passive film. Corrosion of the candidate stainless steels was similarly modest across the entire range of exposures. However, some sensitivity to corrosion of the stainless steels was observed in solutions with 50 wppm iodine (the actual SHINE process expects 0.1–1 wppm) with the highest corrosion rates (up to ~6 mil/year) observed on specimens exposed in the vapor phase. Lower concentrations of iodine species (5 or 28 wppm) proved much less corrosive, and the planned-interval data indicated that metal corrodibility decreased with time for all immersed exposures and, with one minor exception, all vapor exposures. Little change in susceptibility to corrosion was observed as a result of nitric acid additions to the test environment (simulating radiolysis products). The trend toward reduced corrosion (immersion and vapor phase) with decreasing iodine concentration suggests that, at the expected conditions in the SHINE process, it is unlikely that iodine species will generate a general corrosion concern for the candidate stainless steels.« less

The effects of pulsed electromagnetic field (PEMF) on osteoblast-like cells cultured on titanium and titanium-zirconium surfaces.

PubMed

Atalay, Belir; Aybar, Buket; Ergüven, Mine; Emes, Yusuf; Bultan, Özgür; Akça, Kivanç; Yalçin, Serhat; Baysal, Uğur; Işsever, Halim; Çehreli, Murat Cavit; Bilir, Ayhan

2013-11-01

Commercially pure Ti, together with Ti Ni, Ti-6Al-4V, and Ti-6Al-7Nb alloys, are among the materials currently being used for this purpose. Titanium-zirconium (TiZr) has been developed that allows SLActive surface modification and that has comparable or better mechanical strength and improved biocompatibility compared with existing Ti alloys. Furthermore, approaches have targeted making the implant surface more hydrophilic, as with the Straumann SLActive surface, a modification of the SLA surface. The aim of this study is to evaluate the effects of pulsed electromagnetic field (PEMF) to the behavior of neonatal rat calvarial osteoblast-like cells cultured on commercially pure titanium (cpTi) and titanium-zirconium alloy (TiZr) discs with hydrophilic surface properties. Osteoblast cells were cultured on titanium and TiZr discs, and PEMF was applied. Cell proliferation rates, cell numbers, cell viability rates, alkaline phosphatase, and midkine (MK) levels were measured at 24 and 72 hours. At 24 hours, the number of cells was significantly higher in the TiZr group. At 72 hours, TiZr had a significantly higher number of cells when compared to SLActive, SLActive + PEMF, and machine surface + PEMF groups. At 24 hours, cell proliferation was significantly higher in the TiZr group than SLActive and TiZr + PEMF group. At 72 hours, TiZr group had higher proliferation rate than machine surface and TiZr + PEMF. Cell proliferation in the machine surface group was lower than both SLActive + PEMF and machine surface + PEMF. MK levels of PEMF-treated groups were lower than untreated groups for 72 hours. Our findings conclude that TiZr surfaces are similar to cpTi surfaces in terms of biocompatibility. However, PEMF application has a higher stimulative effect on cells cultured on cpTi surfaces when compared to TiZr.

Aluminum-Scandium Alloys: Material Characterization, Friction Stir Welding, and Compatibility With Hydrogen Peroxide (MSFC Center Director's Discretionary Fund Final Report, Proj. No. 04-14)

NASA Technical Reports Server (NTRS)

Lee, J. A.; Chen, P. S.

2004-01-01

This Technical Memorandum describes the development of several high-strength aluminum (Al) alloys that are compatible with hydrogen peroxide (H2O2) propellant for NASA Hypersonic-X (Hyper-X) vehicles fuel tanks and structures. The yield strengths for some of these Al-magnesium-based alloys are more than 3 times stronger than the conventional 5254-H112 Al alloy, while maintaining excellent H2O2 compatibility similar to class 1 5254 alloy. The alloy development strategy is to add scandium, zirconium, and other transitional metals with unique electrochemical properties, which will not act as catalysts, to decompose the highly concentrated 90 percent H2O2. Test coupons are machined from sheet metals for H2O2 long-term exposure testing and mechanical properties testing. In addition, the ability to weld the new alloys using friction stir welding has also been explored. The new high-strength alloys could represent an enabling material technology for Hyper-X vehicles, where flight weight reduction is a critical requirement.

Layer Formation On Metal Surfaces In Lead-Bismuth At High Temperatures In Presence Of Zirconium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Loewen, Eric Paul; Yount, Hannah J.; Volk, Kevin

If the operating temperature lead–bismuth cooled fission reactor could be extended to 800 °C, they could produce hydrogen directly from water. A key issue for the deployment of this technology at these temperatures is the corrosion of the fuel cladding and structural materials by the lead–bismuth. Corrosion studies of several metals were performed to correlate the interaction layer formation rate as a function of time, temperature, and alloy compositions. The interaction layer is defined as the narrow band between the alloy substrate and the solidified lead–bismuth eutectic on the surface. Coupons of HT-9, 410, 316L, and F22 were tested atmore » 550 and 650 °C for 1000 h inside a zirconium corrosion cell. The oxygen potential ranged from approximately 10-22 to 10-19 Pa. Analyses were performed on the coupons to determine the depth of the interaction layer and the composition, at each time step (100, 300, and 1000 h). The thickness of the interaction layer on F22 at 550 °C was 25.3 µm, the highest of all the alloys tested, whereas at 650 °C, the layer thickness was only 5.6 µm, the lowest of all the alloys tested. The growth of the interaction layer on F22 at 650 °C was suppressed, owing to the presence of Zr (at 1500 wppm) in the LBE. In the case of 316L, the interaction layers of 4.9 and 10.6 µm were formed at 550 and 650 °C, respectively.« less

Physical characterization of a new composition of oxidized zirconium-2.5 wt% niobium produced using a two step process for biomedical applications

NASA Astrophysics Data System (ADS)

Pawar, V.; Weaver, C.; Jani, S.

2011-05-01

Zirconium and particularly Zr-2.5 wt%Nb (Zr2.5Nb) alloy are useful for engineering bearing applications because they can be oxidized in air to form a hard surface ceramic. Oxidized zirconium (OxZr) due to its abrasion resistant ceramic surface and biocompatible substrate alloy has been used as a bearing surface in total joint arthroplasty for several years. OxZr is characterized by hard zirconium oxide (oxide) formed on Zr2.5Nb using one step thermal oxidation carried out in air. Because the oxide is only at the surface, the bulk material behaves like a metal, with high toughness. The oxide, furthermore, exhibits high adhesion to the substrate because of an oxygen-rich diffusion hardened zone (DHZ) interposing between the oxide and the substrate. In this study, we demonstrate a two step process that forms a thicker DHZ and thus increased depth of hardening than that can be obtained using a one step oxidation process. The first step is thermal oxidation in air and the second step is a heat treatment in vacuum. The second step drives oxygen from the oxide formed in the first step deeper into the substrate to form a thicker DHZ. During the process only a portion of the oxide is dissolved. This new composition (DHOxZr) has approximately 4-6 μm oxide similar to that of OxZr. The nano-hardness of the oxide is similar but the DHZ is approximately 10 times thicker. The stoichiometry of the oxide is similar and a secondary phase rich in oxygen is present through the entire thickness. Due to the increased depth of hardening, the critical load required for the onset of oxide cracking is approximately 1.6 times more than that of the oxide of OxZr. This new composition has a potential to be used as a bearing surface in applications where greater depth of hardening is required.

Hydrogen transport membranes

DOEpatents

Mundschau, Michael V.

2005-05-31

Composite hydrogen transport membranes, which are used for extraction of hydrogen from gas mixtures are provided. Methods are described for supporting metals and metal alloys which have high hydrogen permeability, but which are either too thin to be self supporting, too weak to resist differential pressures across the membrane, or which become embrittled by hydrogen. Support materials are chosen to be lattice matched to the metals and metal alloys. Preferred metals with high permeability for hydrogen include vanadium, niobium, tantalum, zirconium, palladium, and alloys thereof. Hydrogen-permeable membranes include those in which the pores of a porous support matrix are blocked by hydrogen-permeable metals and metal alloys, those in which the pores of a porous metal matrix are blocked with materials which make the membrane impervious to gases other than hydrogen, and cermets fabricated by sintering powders of metals with powders of lattice-matched ceramic.

Comparison of bio-mineralization behavior of Ti-6Al-4V-1Nb and Zr-1Nb nano-tubes formed by anodization

NASA Astrophysics Data System (ADS)

Choi, Yong; Hong, Sun I.

2014-12-01

Nano-tubes of titanium and zirconium alloys like Ti-6Al-4V-1Nb and Zr-1Nb were prepared by anodization followed by coating with hydroxylapatite (HA) and their bio-mineralization behaviors were compared to develop a bio-compatible material for implants in orthopedics, dentistry and cardiology. Ti-6Al-4V-1Nb weight gain in a simulated body solution increased gradually. The bigger tube diameter was, the heavier HA was deposited. Surface roughness of both alloys increased highly with the increasing diameter of nano-tube. Their surface roughness decreased by HA deposition due to the removal of the empty space of the nano-tubes. Zr-1Nb alloy had faster growth of nano-tubes layers more than Ti-6Al-4V-1Nb alloy.

Mössbauer study of oxide films of Fe-, Sn-, Cr- doped zirconium alloys during corrosion in autoclave

NASA Astrophysics Data System (ADS)

Filippov, V. P.; Bateev, A. B.; Lauer, Yu. A.

2016-12-01

Mössbauer investigations were used to compare iron atom states in oxide films of binary Zr-Fe, ternary Zr-Fe-Cu and quaternary Zr-Fe-Cr-Sn alloys. Oxide films are received in an autoclave at a temperature of 350-360 °C and at pressure of 16.8 MPa. The corrosion process decomposes the intermetallic precipitates in alloys and forms metallic iron with inclusions of chromium atoms α-Fe(Cr), α-Fe(Cu), α-Fe 2O3 and Fe 3O4 compounds. Some iron ions are formed in divalent and in trivalent paramagnetic states. The additional doping influences on corrosion kinetics and concentration of iron compounds and phases formed in oxide films. It was shown the correlation between concentration of iron in different chemical states and corrosion resistance of alloys.

An evaluation of the benefits of utilizing rapid solidification for development of 2XXX (Al-Cu-Mg) alloys

NASA Technical Reports Server (NTRS)

Paris, H. G.; Chellman, D. J.

1986-01-01

The advantages of rapid solidification processing over ingot metallurgy processing in the development of 2XXX aluminum alloy compositions were evaluated using a similarly processed ingot metallurgy (IM) control alloy. The powder metallurgy (PM) alloy extrusions showed a reduced age-hardening response in comparison with similar IM compositions, with higher tensile properties for naturally aged extrusions but lower properties for artificially aged ones. However, the tensile properties of naturally and artificially aged PM alloy extrusions based on a version of IM 2034 alloy, but containing 0.6 weight percent zirconium, were comparable to those of the IM control extrusions and had significantly superior combinations of strength and toughness. The tensile properties of this PM alloy showed even greater advantage in 6.4-mm (0.25-in.) and 1.8-mm (0.070-in.) plate and sheet, the yield strength being about 68 MPa (10 ksi) greater than reported values for the IM 2034 alloy sheet. An artificially aged PM alloy based on 2219 alloy also showed a strength and strength-toughness combination comparable to those of the PM Al-Cu-Mg-Zr alloy, substantially outperforming the IM 2219 alloy. These results show that rapid solidification offers the flexibility needed to modify conventional IM compositions to produce new alloy compositions with superior mechanical properties.

TUNGSTEN INTERFERENCE IN VOLUMETRIC ANALYSIS OF URANIUM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dufour, R.F.; Articolo, O.

1958-08-01

Tungsten was found to have a negligible effect on the determination of uranium in uranium-zirconium alloys by the Jones reductor-dichromate method used at KAPL. The tungstate ion interferred seriously and gave high results. However, the soluble tungsten was precipitated by intensive fuming with sulfuric acid and rendered ineffective in tbe subsequent oxidationreduction reactions of the uranium. (auth)

FUEL ELEMENT AND METHOD OF PREPARATION

DOEpatents

Kingston, W.E.

1961-04-25

A nuclear fuel element in the form of a wire is reported. A bar of uranium is enclosed in a thin layer of aluminum and the composite is sheathed in beryllium, zirconium, or stainnless steel. The sheathed article is then drawn to wire form, heated to alloy the aluminum with both uranium and sheath, and finally cold worked.

The shock and spall response of three industrially important hexagonal close-packed metals: magnesium, titanium and zirconium.

PubMed

Hazell, P J; Appleby-Thomas, G J; Wielewski, E; Escobedo, J P

2014-08-28

Magnesium, titanium and zirconium and their alloys are extensively used in industrial and military applications where they would be subjected to extreme environments of high stress and strain-rate loading. Their hexagonal close-packed (HCP) crystal lattice structures present interesting challenges for optimizing their mechanical response under such loading conditions. In this paper, we review how these materials respond to shock loading via plate-impact experiments. We also discuss the relationship between a heterogeneous and anisotropic microstructure, typical of HCP materials, and the directional dependency of the elastic limit and, in some cases, the strength prior to failure. © 2014 The Author(s) Published by the Royal Society. All rights reserved.

Compatibility of buffered uranium carbides with tungsten.

NASA Technical Reports Server (NTRS)

Phillips, W. M.

1971-01-01

Results of compatibility tests between tungsten and hyperstoichiometric uranium carbide alloys run at 1800 C for 1000 and 2500 hours. These tests compared tungsten-buffered uranium carbide with tungsten-buffered uranium-zirconium carbide. The zirconium carbide addition appeared to widen the homogeneity range of the uranium carbide, making additional carbon available for reaction. Reaction layers could be formed by either of two diffusion paths, one producing UWC2, while the second resulted in the formation of W2C. UWC2 acts as a diffusion barrier for carbon and slows the growth of the reaction layer with time, while carbon diffusion is relatively rapid in W2C, allowing equilibrium to be reached in less than 2500 hours at a temperature of 1800 C.

Hydrogen calibration of GD-spectrometer using Zr-1Nb alloy

NASA Astrophysics Data System (ADS)

Mikhaylov, Andrey A.; Priamushko, Tatiana S.; Babikhina, Maria N.; Kudiiarov, Victor N.; Heller, Rene; Laptev, Roman S.; Lider, Andrey M.

2018-02-01

To study the hydrogen distribution in Zr-1Nb alloy (Э110 alloy) GD-OES was applied in this work. Qualitative analysis needs the standard samples with hydrogen. However, the standard samples with high concentrations of hydrogen in the zirconium alloy which would meet the requirements of the shape, size are absent. In this work method of Zr + H calibration samples production was performed at the first time. Automated Complex Gas Reaction Controller was used for samples hydrogenation. To calculate the parameters of post-hydrogenation incubation of the samples in an inert gas atmosphere the diffusion equations were used. Absolute hydrogen concentrations in the samples were determined by melting in the inert gas atmosphere using RHEN602 analyzer (LECO Company). Hydrogen distribution was studied using nuclear reaction analysis (HZDR, Dresden, Germany). RF GD-OES was used for calibration. The depth of the craters was measured with the help of a Hommel-Etamic profilometer by Jenoptik, Germany.

Possible origin and roles of nano-porosity in ZrO2 scales for hydrogen pick-up in Zr alloys

NASA Astrophysics Data System (ADS)

Lindgren, Mikaela; Geers, Christine; Panas, Itai

2017-08-01

A mechanistic understanding of Wagnerian build-up and subsequent non-Wagnerian break-down of barrier oxide upon oxidation of zirconium alloys by water is reiterated. Hydrogen assisted build-up of nano-porosity is addressed. Growth of sub-nanometer wide stalactitic pores owing to increasing aggregation of neutral oxygen vacancies offering a means to permeate hydrogen into the alloy is explored by density functional theory. The Wagnerian channel utilizes charge separation allowing charged oxygen vacancies and electrons to move separately from nominal anode to nominal cathode. This process becomes increasingly controlled by the charging of the barrier oxide resulting in sub-parabolic rate law for oxide growth. The break-down of the barrier oxide is understood to be preceded by avalanching hydrogen pick-up in the alloy. Pore mediated diffusion allows water to effectively short circuit the barrier oxide.

Using ToF-SIMS and EIS to evaluate green pretreatment reagent: Corrosion protection of aluminum alloy by silica/zirconium/cerium hybrid coating

NASA Astrophysics Data System (ADS)

Chang, Chun-Chao; Wang, Chiung-Chi; Wu, Chia-Wei; Liu, Shou-Ching; Mai, Fu-Der

2008-12-01

Increasing environmental concern has led to the restrictive use of chromate conversion coatings to protect Al-alloys from corrosion. Our research is under way to find environmentally compliant substitute coating such as Si/Zr/Ce hybrid coating. The corrosion protection effect of green pretreatment reagent consisted of Si-containing base solution, Ce- and Zr-containing sealing solutions on the corrosion protection of Al-alloys was studied with a 3.5% NaCl aqueous testing solution. The correlation between the corrosion resistance measured by electrochemical impedance spectroscopy (EIS) and surface chemical composition of the hybrid coating measured by time-of-flight secondary ion mass spectroscopy (ToF-SIMS) was studied. The proposed green pretreatment reagent was found improve the corrosion protection of Al-alloys, presumably due to the formation of protective oxide film acting as an oxygen barrier.

PLUTONIUM METALLIC FUELS FOR FAST REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

STAN, MARIUS; HECKER, SIEGFRIED S.

2007-02-07

Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low-melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuelsmore » suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher-melting U-PuZr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.« less

Protected Nuclear Fuel Element

DOEpatents

Kittel, J. H.; Schumar, J. F.

1962-12-01

A stainless steel-clad actinide metal fuel rod for use in fast reactors is reported. In order to prevert cladding failures due to alloy formation between the actinide metal and the stainless steel, a mesh-like sleeve of expanded metal is interposed between them, the sleeve metal being of niobium, tantalum, molybdenum, tungsten, zirconium, or vanadium. Liquid alkali metal is added as a heat transfer agent. (AEC)

Device and method for shortening reactor process tubes

DOEpatents

Frantz, C.E.; Alexander, W.K.; Lander, W.E.B.

A device and method are described for in situ shortening of nuclear reactor zirconium alloy process tubes which have grown as a result of radiation exposure. An upsetting technique is utilized which involves inductively heating a short band of a process tube with simultaneous application of an axial load sufficient to cause upsetting with an attendant decrease in length of the process tube.

Phase separation of metal-added corium and its effect on a steam explosion

NASA Astrophysics Data System (ADS)

Min, B. T.; Kim, J. H.; Hong, S. W.; Hong, S. H.; Park, I. K.; Song, J. H.; Kim, H. D.

2008-07-01

To simulate a relocation of molten core material and its interaction phenomenon with water during a severe accident in a nuclear reactor, a typical corium of UO 2/ZrO 2/Zr/Stainless steel mixed at a 62 wt%, 15 wt%, 12 wt% and 11 wt%, respectively, was melted and then cooled down to become a solidified ingot. It was shown that the molten corium was separated into two layers, of which the upper layer was oxide mixtures and the lower layer was metal alloys. The upper layer was UO 2 and ZrO 2 and the lower layer mostly consisted of metal mixtures such as uranium, zirconium and stainless steel. Iron content varied with the positions and about a half of it existed as an alloy such as Fe 2U. Uranium metal was produced by reduction of UO 2 by zirconium metal. The average densities of the upper oxide layer and the lower metal layer were 8.802 and 9.411 g/cm 3, respectively. In another test, metal-added molten corium was poured into water and it showed that a steam explosion could occur by applying an external trigger.

Application of hard coatings to substrates at low temperatures

NASA Technical Reports Server (NTRS)

Sproul, William D.

1993-01-01

BIRL, the industrial research laboratory of Northwestern University, has conducted unique and innovative research, under sponsorship from the NASA Marshall Space Flight Center (MSFC), in the application of hard, wear resistant coatings to bearing steels using the high-rate reactive sputtering (HRRS) process that was pioneered by Dr. William Sproul, the principal investigator on this program. Prior to this program, Dr. Sproul had demonstrated that it is possible to apply hard coatings such as titanium nitride (TiN) to alloy steels at low temperatures via the HRRS process without changing the metallurgical properties of the steel. The NASA MSFC program at BIRL had the specific objectives to: apply TiN to 440C stainless steel without changing the metallurgical properties of the steel; prepare rolling contact fatigue (RCF) test samples coated with binary hard coatings of TiN, zirconium nitride (ZrN), hafnium nitride (HfN), chromium nitride (CrN), and molybdenum nitride (MoN), and metal coatings of copper (Cu) and gold (Au); and develop new alloyed hard coatings of titanium aluminum nitride (Ti(0.5)Al(0.5)N), titanium zirconium nitride (Ti(0.5)Zr(0.5)N), and titanium aluminum vanadium nitride.

Materials Coating Techniques

DTIC Science & Technology

1980-03-01

applications from decorative to utilitarian over significant segments of the engineering, chemical, nuclear , microelectronics, and related Industries. PVD...Thermal-control coating. Boron 2430 Cermet component, nuclear shielding and controlrod material; Carbide wear- and temperature-resistant. Calcium...Zirconium Oxide (Hafnia-Pree � Thermal-barrier coatings for nuclear applications. Lime Stabi!Aed) Zirconium 2563 Resistant to high-temperature

Carbide and carbonitride surface treatment method for refractory metals

DOEpatents

Meyer, G.A.; Schildbach, M.A.

1996-12-03

A carbide and carbonitride surface treatment method for refractory metals is provided, in steps including, heating a part formed of boron, chromium, hafnium, molybdenum, niobium, tantalum, titanium, tungsten or zirconium, or alloys thereof, in an evacuated chamber and then introducing reaction gases including nitrogen and hydrogen, either in elemental or water vapor form, which react with a source of elemental carbon to form carbon-containing gaseous reactants which then react with the metal part to form the desired surface layer. Apparatus for practicing the method is also provided, in the form of a carbide and carbonitride surface treatment system including a reaction chamber, a source of elemental carbon, a heating subassembly and a source of reaction gases. Alternative methods of providing the elemental carbon and the reaction gases are provided, as well as methods of supporting the metal part, evacuating the chamber with a vacuum subassembly and heating all of the components to the desired temperature. 5 figs.

Around Marshall

NASA Image and Video Library

2003-01-01

This Photo, which appeared on the July cover of `Physics Today', is of the Electrostatic Levitator (ESL) at NASA's Marshall Space Flight Center (MSFC). The ESL uses static electricity to suspend an object (about 3-4 mm in diameter) inside a vacuum chamber allowing scientists to record a wide range of physical properties without the sample contracting the container or any instruments, conditions that would alter the readings. Once inside the chamber, a laser heats the sample until it melts. The laser is then turned off and the sample cools, changing from a liquid drop to a solid sphere. In this particular shot, the ESL contains a solid metal sample of titanium-zirconium-nickel alloy. Since 1977, the ESL has been used at MSFC to study the characteristics of new metals, ceramics, and glass compounds. Materials created as a result of these tests include new optical materials, special metallic glasses, and spacecraft components.

Effect of power and type of substrate on calcium-phosphate coating morphology and microhardness

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kulyashova, Ksenia, E-mail: kseniya@ispms.tsc.ru; Glushko, Yurii, E-mail: glushko@ispms.tsc.ru; Sharkeev, Yurii, E-mail: sharkeev@ispms.tsc.ru

2015-10-27

As known, the influence of the different sputtering process parameters and type of substrate on structure of the deposited coating is important to identify, because these parameters are significantly affected on structure of coating. The studies of the morphology and microhardness of calcium-phosphate (CaP) coatings formed and obtained on the surface of titanium, zirconium, titanium and niobium alloy for different values of the power of radio frequency discharge are presented. The increase in the radio frequency (rf) magnetron discharge leads to the formation of a larger grain structure of the coating. The critical depths of indentation for coatings determining themore » value of their microhardness have been estimated. Mechanical properties of the composite material on the basis of the bioinert substrate metal and CaP coatings are superior to the properties of the separate components that make up this composite material.« less

The Mechanical Response of Advanced Claddings during Proposed Reactivity Initiated Accident Conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cinbiz, Mahmut N; Brown, Nicholas R; Terrani, Kurt A

2017-01-01

This study investigates the failure mechanisms of advanced nuclear fuel cladding of FeCrAl at high-strain rates, similar to design basis reactivity initiated accidents (RIA). During RIA, the nuclear fuel cladding was subjected to the plane-strain to equibiaxial tension strain states. To achieve those accident conditions, the samples were deformed by the expansion of high strength Inconel alloy tube under pre-specified pressure pulses as occurring RIA. The mechanical response of the advanced claddings was compared to that of hydrided zirconium-based nuclear fuel cladding alloy. The hoop strain evolution during pressure pulses were collected in situ; the permanent diametral strains of bothmore » accident tolerant fuel (ATF) claddings and the current nuclear fuel alloys were determined after rupture.« less

High strength nickel-chromium-iron austenitic alloy

DOEpatents

Gibson, Robert C.; Korenko, Michael K.

1980-01-01

A solid solution strengthened Ni-Cr-Fe alloy capable of retaining its strength at high temperatures and consisting essentially of 42 to 48% nickel, 11 to 13% chromium, 2.6 to 3.4% niobium, 0.2 to 1.2% silicon, 0.5 to 1.5% vanadium, 2.6 to 3.4% molybdenum, 0.1 to 0.3% aluminum, 0.1 to 0.3% titanium, 0.02 to 0.05% carbon, 0.002 to 0.015% boron, up to 0.06 zirconium, and the balance iron. After solution annealing at 1038.degree. C. for one hour, the alloy, when heated to a temperature of 650.degree. C., has a 2% yield strength of 307 MPa, an ultimate tensile strength of 513 MPa and a rupture strength of as high as 400 MPa after 100 hours.

Electrochemical study of pre- and post-transition corrosion of Zr alloys in PWR coolant

NASA Astrophysics Data System (ADS)

Macák, Jan; Novotný, Radek; Sajdl, Petr; Renčiuková, Veronika; Vrtílková, Věra

Corrosion properties of Zr-Sn and Zr-Nb zirconium alloys were studied under simulated PWR conditions (or, more exactly, VVER conditions — boric acid, potassium hydroxide, lithium hydroxide) at temperatures up to 340°C and 15MPa using in-situ electrochemical impedance spectroscopy (EIS) and polarization measurements. EIS spectra were obtained in a wide range of frequencies (typically 100kHz — 100μHz). It enabled to gain information of both dielectric properties of oxide layers developing on the Zr-alloys surface and of the kinetics of the corrosion process and the associated charge and mass transfer phenomena. Experiments were run for more than 380 days; thus, the study of all the corrosion stages (pre-transition, transition, post-transition) was possible.

Dislocation dynamics simulations of interactions between gliding dislocations and radiation induced prismatic loops in zirconium

NASA Astrophysics Data System (ADS)

Drouet, Julie; Dupuy, Laurent; Onimus, Fabien; Mompiou, Frédéric; Perusin, Simon; Ambard, Antoine

2014-06-01

The mechanical behavior of Pressurized Water Reactor fuel cladding tubes made of zirconium alloys is strongly affected by neutron irradiation due to the high density of radiation induced dislocation loops. In order to investigate the interaction mechanisms between gliding dislocations and loops in zirconium, a new nodal dislocation dynamics code, adapted to Hexagonal Close Packed metals, has been used. Various configurations have been systematically computed considering different glide planes, basal or prismatic, and different characters, edge or screw, for gliding dislocations with -type Burgers vectors. Simulations show various interaction mechanisms such as (i) absorption of a loop on an edge dislocation leading to the formation of a double super-jog, (ii) creation of a helical turn, on a screw dislocation, that acts as a strong pinning point or (iii) sweeping of a loop by a gliding dislocation. It is shown that the clearing of loops is more favorable when the dislocation glides in the basal plane than in the prismatic plane explaining the easy dislocation channeling in the basal plane observed after neutron irradiation by transmission electron microscopy.

Structure and Thermodynamical Properties of Zirconium Hydrides from First-Principle

NASA Astrophysics Data System (ADS)

Blomqvist, Jakob; Olofsson, Johan; Alvarez, Anna-Maria; Bjerkén, Christina

Zirconium alloys are used as nuclear fuel cladding material due to their mechanical and corrosion resistant properties together with their favorable cross-section for neutron scattering. At running conditions, however, there will be an increase of hydrogen in the vicinity of the cladding surface at the water side of the fuel. The hydrogen will diffuse into the cladding material and at certain conditions, such as lower temperatures and external load, hydrides will precipitate out in the material and cause well known embrittlement, blistering and other unwanted effects. Using phase-field methods it is now possible to model precipitation buildup in metals, for example as a function of hydrogen concentration, temperature and external load, but the technique relies on input of parameters, such as the formation energy of the hydrides and matrix. To that end, we have computed, using the density functional theory (DFT) code GPAW, the latent heat of fusion as well as solved the crystal structure for three zirconium hydride polymorphs: δ-ZrH1.6, γ-ZrH, and Є-ZrH2.

Effect of solutes in binary columbium /Nb/ alloys on creep strength

NASA Technical Reports Server (NTRS)

Klein, M. J.; Metcalfe, A. G.

1973-01-01

The effect of seven different solutes in binary columbium (Nb) alloys on creep strength was determined from 1400 to 3400 F for solute concentrations to 20 at.%, using a new method of creep-strength measurement. The technique permits rapid determination of approximate creep strength over a large temperature span. All of the elements were found to increase the creep strength of columbium except tantalum. This element did not strengthen columbium until the concentration exceeded 10 at.%. Hafnium, zirconium, and vanadium strengthed columbium most at low temperatures and concentrations, whereas tungsten, molybdenum, and rhenium contributed more to creep strength at high temperatures and concentrations.

General Corrosion Resistance Assessments of AA7085, AA7129, and Other High-Performance Aluminum Alloys for Department of Defense (DOD) Systems UsingLaboratory Based Accelerated Corrosion Methods and Electrochemistry

DTIC Science & Technology

2013-09-01

laboratory should play a role in the final design decision process. Integration factors such as conversion coatings , primers, topcoats, and their...Cyclic Accelerated Corrosion Analysis of Nonchromate Conversion Coatings on Aluminum Alloys 2024, 2219, 5083, and 7075 Using DoD Paint Systems; ARL...Titanium 0.08 0.10 max 0.10 max 0.15 max 0.08 max 0.05 max Zirconium 0.05 – 0.15 0.05 – 0.15 - 0.10 – 0.25 0.05 – 0.15 - Vanadium - - - - - 0.05 max

Characterisation of well-adhered ZrO2 layers produced on structured reactors using the sonochemical sol-gel method

NASA Astrophysics Data System (ADS)

Jodłowski, Przemysław J.; Chlebda, Damian K.; Jędrzejczyk, Roman J.; Dziedzicka, Anna; Kuterasiński, Łukasz; Sitarz, Maciej

2018-01-01

The aim of this study was to obtain thin zirconium dioxide coatings on structured reactors using the sonochemical sol-gel method. The preparation method of metal oxide layers on metallic structures was based on the synergistic combination of three approaches: the application of ultrasonic irradiation during the synthesis of Zr sol-gel based on a precursor solution containing zirconium(IV) n-propoxide, the addition of stabilszing agents, and the deposition of ZrO2 on the metallic structures using the dip-coating method. As a result, dense, uniform zirconium dioxide films were obtained on the FeCrAlloy supports. The structured reactors were characterised by various physicochemical methods, such as BET, AFM, EDX, XRF, XRD, XPS and in situ Raman spectroscopy. The results of the structural analysis by Raman and XPS spectroscopy confirmed that the metallic surface was covered by a ZrO2 layer without any impurities. SEM/EDX mapping revealed that the deposited ZrO2 covered the metallic support uniformly. The mechanical and high temperature tests showed that the developed ultrasound assisted sol-gel method is an efficient way to obtain thin, well-adhered zirconium dioxide layers on the structured reactors. The prepared metallic supports covered with thin ZrO2 layers may be a good alternative to layered structured reactors in several dynamics flow processes, for example for gas exhaust abatement.

Device and method for shortening reactor process tubes

DOEpatents

Frantz, Charles E.; Alexander, William K.; Lander, Walter E. B.

1980-01-01

This disclosure describes a device and method for in situ shortening of nuclear reactor zirconium alloy process tubes which have grown as a result of radiation exposure. An upsetting technique is utilized which involves inductively heating a short band of a process tube with simultaneous application of an axial load sufficient to cause upsetting with an attendant decrease in length of the process tube.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pickman, D.O.

Various aspects of zirconium alloy development for light water reactors in the UK and Scandinavia are reviewed, including the contribution made by some unique nuclear testing facilities. Among the problems encountered were the irradiation enhancement of corrosion and hydrogen pickup, crud deposition, iodine-induced stress-corrosion cracking on power ramping, and severe cladding deformation in loss-of-coolant accident conditions. The causes and behavior of defects, including hydride defects and fretting corrosion, are discussed.

Understanding thermally activated plastic deformation behavior of Zircaloy-4

NASA Astrophysics Data System (ADS)

Kumar, N.; Alomari, A.; Murty, K. L.

2018-06-01

Understanding micromechanics of plastic deformation of existing materials is essential for improving their properties further and/or developing advanced materials for much more severe load bearing applications. The objective of the present work was to understand micromechanics of plastic deformation of Zircaloy-4, a zirconium-based alloy used as fuel cladding and channel (in BWRs) material in nuclear reactors. The Zircaloy-4 in recrystallized (at 973 K for 4 h) condition was subjected to uniaxial tensile testing at a constant cross-head velocity at temperatures in the range 293 K-1073 K and repeated stress relaxation tests at 293 K, 573 K, and 773 K. The minimum in the total elongation was indicative of dynamic strain aging phenomenon in this alloy in the intermediate temperature regime. The yield stress of the alloy was separated into effective and athermal components and the transition from thermally activated dislocation glide to athermal regime took place at around 673 K with the athermal stress estimated to be 115 MPa. The activation volume was found to be in the range of 40 b3 to 160 b3. The activation volume values and the data analyses using the solid-solution models in literature indicated dislocation-solute interaction to be a potential deformation mechanism in thermally activated regime. The activation energy calculated at 573 K was very close to that found for diffusivity of oxygen in α-Zr that was suggestive of dislocations-oxygen interaction during plastic deformation. This type of information may be helpful in alloy design in selecting different elements to control the deformation behavior of the material and impart desired mechanical properties in those materials for specific applications.

The use of precious-metal-modified nickel-based superalloys for thin gage applications

NASA Astrophysics Data System (ADS)

Ballard, Donna L.; Pilchak, Adam L.

2010-10-01

Precious-metal-modified nickel-based superalloys are being investigated for use in thin gage applications, such as thermal protection systems or heat exchangers, due to their strength and inherent oxidation resistance at temperatures in excess of 1,050°C. This overview paper summarizes the Air Force Research Laboratory (AFRL) interest in experimental two-phase γ-Ni + γ'-Ni3Al superalloys. The AFRL is interested in alloys with a based composition of Ni-15Al-5Cr (at. %) with carbon, boron, and zirconium additions for grain-boundary refinement and strengthening. The alloys currently being evaluated also contain 4-5 at.% of platinum-group metals, in this case platinum and iridium. The feasibility of hot rolling these alloys to a final thickness of 0.12-0.25 mm and obtaining a nearly fully recrystallized microstructure was demonstrated.

Improvement of GRCop-84 Through the Addition of Zirconium

NASA Technical Reports Server (NTRS)

Ellis, David L.; Lerch, Bradley A.

2012-01-01

GRCop-84 (Cu-8 at.% Cr-4 at.% Nb) has excellent strength, creep resistance, low cycle fatigue (LCF) life and stability at elevated temperatures. It suffers in comparison to many commercially available precipitation-strengthened alloys below 500 C (932 F). It was observed that the addition of Zr consistently improved the mechanical properties of Cu-based alloys especially below 500 C. In an effort to improve the low temperature properties of GRCop-84, 0.35 wt.% Zr was added to the alloy. Limited tensile, creep, and LCF testing was conducted to determine if improvements occur. The results showed some dramatic increases in the tensile and creep properties at the conditions tested with the probability of additional improvements being possible through cold working. LCF testing at room temperature did not show an improvement, but improvements might occur at elevated temperatures.

10 CFR Appendix I to Part 110 - Illustrative List of Reprocessing Plant Components Under NRC Export Licensing Authority

Code of Federal Regulations, 2011 CFR

2011-01-01

... assurance and quality control techniques) out of low carbon stainless steels, titanium, zirconium or other... materials such as low carbon stainless steels, titanium or zirconium, or other high quality materials... features for control of nuclear criticality: (i) Walls or internal structures with a boron equivalent of at...

Thermal Stabilization and Mechanical Properties of Nanocrystalline Iron-Nickel-Zirconium Alloys

NASA Astrophysics Data System (ADS)

Kotan, Hasan

Ultrafine grained and nanostructured materials are promising for structural applications because of the high strength compared to coarse grained counterparts. However, their widespread application is limited by an inherently high driving force for thermally induced grain growth, even at low temperatures. Accordingly, the understanding of and control over grain growth in nanoscale materials is of great technological and scientific importance as many physical properties (i.e. mechanical properties) are functions of the average grain size and the grain size distribution within the microstructure. Here, we investigate the microstructural evolution and grain growth in Fe-Ni alloys with Zr addition and differentiate the stabilization mechanisms acting on grain boundaries. Fe-Ni alloys are chosen for stability investigations since they are important for understanding the behavior of many steels and other ferrous alloys. Zirconium is proven to be an effective grain size stabilizer in pure Fe and Fe-base systems. In this study, nanocrystalline alloys were prepared by high energy ball milling. In situ and ex situ experiments were utilized to directly follow grain growth and microstructural evolution as a function of temperature and composition. The information obtained from these experiments enables the real time observation of microstructural evolution and phase transformation and provides a unique view of dynamic reactions as they occur. The knowledge of the thermal stability will exploit the potential high temperature applications and the consolidation conditions (i.e. temperature and pressure) to obtain high dense materials for advanced mechanical tests. Our investigations reveal that the grain growth of Fe-Ni alloys is not affected by Ni content but strongly inhibited by the addition of 1 at% Zr up to about 700 °C. The microstructural stability is lost due to the bcc-to-fcc transformation (occurring at 700°C) by the sudden appearance of abnormally grown fcc grains. However it was determined grain growth can be suppressed kinetically at higher temperatures for high Zr containing alloys by precipitation of intermetallic compounds. Eventually at high enough temperatures the retention of nanocrystallinity was lost, leaving behind fine micron grains filled with nanoscale intermetallic precipitates. Despite the loss of stability the in-situ formed precipitates were found to induce an Orowan hardening affect. The results from the mechanical tests show that Orowan particle strengthening can be as significant as Hall Petch hardening is at the smallest grain sizes.

In vitro degradation behavior and cytocompatibility of Mg–Zn–Zr alloys

PubMed Central

Huan, Z. G.; Leeflang, M. A.; Fratila-Apachitei, L. E.; Duszczyk, J.

2010-01-01

Zinc and zirconium were selected as the alloying elements in biodegradable magnesium alloys, considering their strengthening effect and good biocompatibility. The degradation rate, hydrogen evolution, ion release, surface layer and in vitro cytotoxicity of two Mg–Zn–Zr alloys, i.e. ZK30 and ZK60, and a WE-type alloy (Mg–Y–RE–Zr) were investigated by means of long-term static immersion testing in Hank’s solution, non-static immersion testing in Hank’s solution and cell-material interaction analysis. It was found that, among these three magnesium alloys, ZK30 had the lowest degradation rate and the least hydrogen evolution. A magnesium calcium phosphate layer was formed on the surface of ZK30 sample during non-static immersion and its degradation caused minute changes in the ion concentrations and pH value of Hank’s solution. In addition, the ZK30 alloy showed insignificant cytotoxicity against bone marrow stromal cells as compared with biocompatible hydroxyapatite (HA) and the WE-type alloy. After prolonged incubation for 7 days, a stimulatory effect on cell proliferation was observed. The results of the present study suggested that ZK30 could be a promising material for biodegradable orthopedic implants and worth further investigation to evaluate its in vitro and in vivo degradation behavior. PMID:20532960

Overview of the multifaceted activities towards development and deployment of nuclear-grade FeCrAl Alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, Kevin G; Yamamoto, Yukinori; Pint, Bruce A

2016-01-01

A large effort is underway under the leadership of US DOE Fuel Cycle R&D program to develop advanced FeCrAl alloys as accident tolerant fuel (ATF) cladding to replace Zr-based alloys in light water reactors. The primary motivation is the excellent oxidation resistance of these alloys in high-temperature steam environments right up to their melting point (roughly three orders of magnitude slower oxidation kinetics than zirconium). A multifaceted effort is ongoing to rapidly advance FeCrAl alloys as a mature ATF concept. The activities span the broad spectrum of alloy development, environmental testing (high-temperature high-pressure water and elevated temperature steam), detailed mechanicalmore » characterization, material property database development, neutron irradiation, thin tube production, and multiple integral fuel test campaigns. Instead of off-the-shelf commercial alloys that might not prove optimal for the LWR fuel cladding application, a large amount of effort has been placed on the alloy development to identify the most optimum composition and microstructure for this application. The development program is targeting a cladding that offers performance comparable to or better than modern Zr-based alloys under normal operating and off-normal conditions. This paper provides a comprehensive overview of the systematic effort to advance nuclear-grade FeCrAl alloys as an ATF cladding in commercial LWRs.« less

Oxidized zirconium versus cobalt-chromium against the native patella in total knee arthroplasty: Patellofemoral outcomes.

PubMed

Matassi, Fabrizio; Paoli, Tommaso; Civinini, Roberto; Carulli, Christian; Innocenti, Massimo

2017-10-01

Oxidized zirconium (OxZr) has demonstrated excellent mechanical properties in vitro when used against articular cartilage; less coefficient of friction and less chondral damage have been found when compared with cobalt-chromium (CoCr) implants. However, controversy exists as to whether implants with a zirconium femoral component articulate safely with a native patella in total knee arthroplasty (TKA). To answer this question, the clinical and radiographic results were analysed from a group of patients who underwent a TKA with patella retention; the OxZr versus CoCr femoral components were compared. The present study prospectively evaluated 83 knees of 74 patients from 2009 to 2010. Each patient was evaluated clinically (visual analogue scale, Knee Society score, patellar score) and radiographically (long leg standing radiograph, anterior-posterior and latero-lateral projections, axial view of the patella) pre-operatively and postoperatively with a mean follow-up of 4.47years. The patellar tilt and shift, and progression of patellofemoral osteoarthritis were calculated with the axial view. There were no patient reported adverse reactions and none of the evaluated prostheses failed. Both the clinical and radiographic evaluations showed no statistically significant between-group differences. No adverse events were observed clinically or radiologically. These results justify pursuing the use of oxidized zirconium as an alternative bearing surface for a femoral component associated with patellar retention in TKA. Published by Elsevier B.V.

Nuclear fuel element

DOEpatents

Armijo, Joseph S.; Coffin, Jr., Louis F.

1983-01-01

A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has a composite cladding having a substrate and a metal barrier metallurgically bonded on the inside surface of the substrate so that the metal barrier forms a shield between the substrate and the nuclear fuel material held within the cladding. The metal barrier forms about 1 to about 30 percent of the thickness of the cladding and is comprised of a low neutron absorption metal of substantially pure zirconium. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the substrate from contact and reaction with such impurities and fission products. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy. Methods of manufacturing the composite cladding are also disclosed.

Electrochemical studies on zirconium and its biocompatible alloys Ti-50Zr at.% and Zr-2.5Nb wt.% in simulated physiologic media.

PubMed

Oliveira, Nilson T C; Biaggio, Sonia R; Rocha-Filho, Romeu C; Bocchi, Nerilso

2005-09-01

Different electrochemical studies were carried out for Zr and its biocompatible alloys Ti-50Zr at.% and Zr-2.5Nb wt.% in solutions simulating physiologic media, Ringer and PBS (phosphate buffered saline) solutions. The results from rest-potential measurements showed that the three materials are spontaneously passivated in both solutions and that the Ti-50Zr alloy has the greatest tendency for spontaneous oxide formation. Some corrosion parameters (such as the pitting and repassivation potentials) were obtained via cyclic voltammetry in both solutions, revealing that the Ti-50Zr has the best corrosion protection while Zr has the worst. On the other hand, the pre-anodization (up to 8 V vs. SCE) of the alloys in a 0.15 mol/L Na2SO4 solution led to a significant improvement in their protection against pitting corrosion when exposed to the Ringer solution. Elemental analyses by EDX showed that during pitting corrosion, there is no preferential corrosion of any of the alloying elements (Zr, Ti, Nb). Copyright (c) 2005 Wiley Periodicals, Inc.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kohnert, Aaron A.; Dasgupta, Dwaipayan; Wirth, Brian

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, the material response must be demonstrated to provide suitable radiation stability, in order to ensuremore » that there will not be significant dimensional changes (e.g., swelling), as well as quantifying the radiation hardening and radiation creep behavior. In this report, we describe the use of cluster dynamics modeling to evaluate the defect physics and damage accumulation behavior of FeCrAl alloys subjected to neutron irradiation, with a particular focus on irradiation-induced swelling and defect fluxes to dislocations that are required to model irradiation creep behavior.« less

Biocorrosion resistance of coated magnesium alloy by microarc oxidation in electrolyte containing zirconium and calcium salts

NASA Astrophysics Data System (ADS)

Wang, Ya-Ming; Guo, Jun-Wei; Wu, Yun-Feng; Liu, Yan; Cao, Jian-Yun; Zhou, Yu; Jia, De-Chang

2014-09-01

The key to use magnesium alloys as suitable biodegradable implants is how to adjust their degradation rates. We report a strategy to prepare biocompatible ceramic coating with improved biocorrosion resistance property on AZ91D alloy by microarc oxidation (MAO) in a silicate-K2ZrF6 solution with and without Ca(H2PO4)2 additives. The microstructure and biocorrosion of coatings were characterized by XRD and SEM, as well as electrochemical and immersion tests in simulated body fluid (SBF). The results show that the coatings are mainly composed of MgO, Mg2SiO4, m-ZrO2 phases, further Ca containing compounds involve the coating by Ca(H2PO4)2 addition in the silicate-K2ZrF6 solution. The corrosion resistance of coated AZ91D alloy is significantly improved compared with the bare one. After immersing in SBF for 28 d, the Si-Zr5-Ca0 coating indicates a best corrosion resistance performance.

Aluminum Alloy and Article Cast Therefrom

NASA Technical Reports Server (NTRS)

Lee, Jonathan A. (Inventor); Chen, Po-Shou (Inventor)

2003-01-01

A cast article from an aluminum alloy, which has improved mechanical properties at elevated temperatures, has the following composition in weight percent: Silicon 14 - 25.0, Copper 5.5 - 8.0, Iron 0.05 - 1.2, Magnesium 0.5 - 1.5, Nickel 0.05 - 0.9, Manganese 0.05 - 1.0, Titanium 0.05 - 1.2, Zirconium 0.05 - 1.2, Vanadium 0.05 - 1.2, Zinc 0.05 - 0.9, Phosphorus 0.001 - 0.1, and the balance is Aluminum, wherein the silicon-to-magnesium ratio is 10 - 25, and the copper-to-magnesium ratio is 4 - 15. The aluminum alloy contains a simultaneous dispersion of three types of Al3X compound particles (X=Ti, V, Zr) having a LI2, crystal structure, and their lattice parameters are coherent to the aluminum matrix lattice. A process for producing this cast article is also disclosed, as well as a metal matrix composite, which includes the aluminum alloy serving as a matrix and containing up to about 60% by volume of a secondary filler material.

40 CFR 421.334 - Standards of performance for new sources.

Code of Federal Regulations, 2013 CFR

2013-07-01

... range of 7.5 to 10.0 at all times. (d) SiC14 purification wet air pollution control. NSPS for the....600 189.300 pH (1) (1) 1 Within the range of 7.5 to 10.0 at all times. (q) Leaching rinse water from...) 1 Within the range of 7.5 to 10.0 at all times. (r) Leaching rinse water from zirconium alloy...

40 CFR 421.334 - Standards of performance for new sources.

Code of Federal Regulations, 2012 CFR

2012-07-01

... range of 7.5 to 10.0 at all times. (d) SiC14 purification wet air pollution control. NSPS for the....600 189.300 pH (1) (1) 1 Within the range of 7.5 to 10.0 at all times. (q) Leaching rinse water from...) 1 Within the range of 7.5 to 10.0 at all times. (r) Leaching rinse water from zirconium alloy...

40 CFR 421.334 - Standards of performance for new sources.

Code of Federal Regulations, 2014 CFR

2014-07-01

... range of 7.5 to 10.0 at all times. (d) SiC14 purification wet air pollution control. NSPS for the....600 189.300 pH (1) (1) 1 Within the range of 7.5 to 10.0 at all times. (q) Leaching rinse water from...) 1 Within the range of 7.5 to 10.0 at all times. (r) Leaching rinse water from zirconium alloy...

Analytical electron microscope study of the omega phase transformation in a zirconium-niobium alloy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zaluzec, N. J.

1979-01-01

The study of the as-quenched omega phase morphology shows that the domain size of Zr-15% Nb is on the order of 30 A. No alignment of omega domains along <222>..beta.. directions was observed and samples having undergone thermal cycling in thin foil form, did not develop a long-period structure of alternating ..beta.. and ..omega.. phases below the omega transformation temperature. (FS)

In situ monitored in-pile creep testing of zirconium alloys

NASA Astrophysics Data System (ADS)

Kozar, R. W.; Jaworski, A. W.; Webb, T. W.; Smith, R. W.

2014-01-01

The experiments described herein were designed to investigate the detailed irradiation creep behavior of zirconium based alloys in the HALDEN Reactor spectrum. The HALDEN Test Reactor has the unique capability to control both applied stress and temperature independently and externally for each specimen while the specimen is in-reactor and under fast neutron flux. The ability to monitor in situ the creep rates following a stress and temperature change made possible the characterization of creep behavior over a wide stress-strain-rate-temperature design space for two model experimental heats, Zircaloy-2 and Zircaloy-2 + 1 wt%Nb, with only 12 test specimens in a 100-day in-pile creep test program. Zircaloy-2 specimens with and without 1 wt% Nb additions were tested at irradiation temperatures of 561 K and 616 K and stresses ranging from 69 MPa to 455 MPa. Various steady state creep models were evaluated against the experimental results. The irradiation creep model proposed by Nichols that separates creep behavior into low, intermediate, and high stress regimes was the best model for predicting steady-state creep rates. Dislocation-based primary creep, rather than diffusion-based transient irradiation creep, was identified as the mechanism controlling deformation during the transitional period of evolving creep rate following a step change to different test conditions.

A CAD/CAM Zirconium Bar as a Bonded Mandibular Fixed Retainer: A Novel Approach with Two-Year Follow-Up.

PubMed

Zreaqat, Maen; Hassan, Rozita; Hanoun, Abdul Fatah

2017-01-01

Stainless steel alloys containing 8% to 12% nickel and 17% to 22% chromium are generally used in orthodontic appliances. A major concern has been the performance of alloys in the environment in which they are intended to function in the oral cavity. Biodegradation and metal release increase the risk of hypersensitivity and cytotoxicity. This case report describes for the first time a CAD/CAM zirconium bar as a bonded mandibular fixed retainer with 2-year follow-up in a patient who is subjected to long-term treatment with fixed orthodontic appliance and suspected to have metal hypersensitivity as shown by the considerable increase of nickel and chromium concentrations in a sample of patient's unstimulated saliva. The CAD/CAM design included a 1.8 mm thickness bar on the lingual surface of lower teeth from canine to canine with occlusal rests on mesial side of first premolars. For better retention, a thin layer of feldspathic ceramic was added to the inner surface of the bar and cemented with two dual-cured cement types. The patient's complaint subsided 6 weeks after cementation. Clinical evaluation appeared to give good functional value where the marginal fit of digitized CAD/CAM design and glazed surface offered an enhanced approach of fixed retention.

A CAD/CAM Zirconium Bar as a Bonded Mandibular Fixed Retainer: A Novel Approach with Two-Year Follow-Up

PubMed Central

Hassan, Rozita; Hanoun, Abdul Fatah

2017-01-01

Stainless steel alloys containing 8% to 12% nickel and 17% to 22% chromium are generally used in orthodontic appliances. A major concern has been the performance of alloys in the environment in which they are intended to function in the oral cavity. Biodegradation and metal release increase the risk of hypersensitivity and cytotoxicity. This case report describes for the first time a CAD/CAM zirconium bar as a bonded mandibular fixed retainer with 2-year follow-up in a patient who is subjected to long-term treatment with fixed orthodontic appliance and suspected to have metal hypersensitivity as shown by the considerable increase of nickel and chromium concentrations in a sample of patient's unstimulated saliva. The CAD/CAM design included a 1.8 mm thickness bar on the lingual surface of lower teeth from canine to canine with occlusal rests on mesial side of first premolars. For better retention, a thin layer of feldspathic ceramic was added to the inner surface of the bar and cemented with two dual-cured cement types. The patient's complaint subsided 6 weeks after cementation. Clinical evaluation appeared to give good functional value where the marginal fit of digitized CAD/CAM design and glazed surface offered an enhanced approach of fixed retention. PMID:28819572

High weldability nickel-base superalloy

DOEpatents

Gibson, Robert C.; Korenko, Michael K.

1980-01-01

This is a nickel-base superalloy with excellent weldability and high strength. Its composition consists essentially of, by weight percent, 10-20 iron, 57-63 nickel, 7-18 chromium, 4-6 molybdenum, 1-2 niobium, 0.2-0.8 silicon, 0.01-0.05 zirconium, 1.0-2.5 titanium, 1.0-2.5 aluminum, 0.02-0.06 carbon, and 0.002-0.015 boron. The weldability and strength of this alloy give it a variety of applications. The long-time structural stability of this alloy together with its low swelling under nuclear radiation conditions, make it especially suitable for use as a duct material and controlling element cladding for sodium-cooled nuclear reactors.

The study of the modes of Ta-Zr powder mixture non-vacuum electron-beam cladding on the surface of the cp-titanium plates

NASA Astrophysics Data System (ADS)

Samoylenko, V. V.; Lozhkina, E. A.; Polyakov, I. A.; Lenivtseva, O. G.; Ivanchik, I. S.; Matts, O. E.

2016-11-01

The effect of the modes of non-vacuum electron-beam cladding of Ta-Zr powder mixtures on the structure and properties of the layers formed on the surface of cp-titanium were studied. The mode of the electron-beam alloying of titanium with zirconium and tantalum, which ensured the formation of a defect-free layer with a high content of alloying elements was selected. Metallographic examination indicated the presence of a dendritic- and plate-type structure of cladded layers. The microhardness of the layers, formed at the optimum mode, was not changed in the cross section and was equal to 450 HV.

Effect of Microstructure on the Mechanical Properties of Extruded Magnesium and a Magnesium Alloy

NASA Astrophysics Data System (ADS)

McGhee, Paul

The main objective of this research was to investigate the relationship between the fatigue behavior and crystallographic texture evolution of magnesium (Mg) alloys with a range of microalloying element content processed under various extrusion conditions. Several Mg alloys were processed under a range of extrusion temperatures, extrusion ratios, and alloying content and tested under monotonic and cyclic fatigue loading conditions: fully-reversed condition tested at strain amplitudes of 0.15% - 1.00% in strain-control mode. After fatigue testing, Mg microstructural analysis was performed using SEM, TEM, optical microscopy, and X-ray diffraction techniques. Microstructural observations revealed significant grain refinement through a combination of zirconium (Zr) addition and hot-extrusion, producing fine equiaxed grain structure with grain sizes ranging between 1-5 microm. Texture analysis and partial compression testing results showed that the initial texture of the extruded alloy gradually evolved upon compressive loading along the c-axes inducing extension twinning creating a strong basal texture along the extrusion direction. Full tensile and compression testing at room temperature showed that the combination of hot extrusion and Zr addition can further refine the grains of the Mg alloys microstructure and enhance the texture while simultaneously enhancing the mechanical properties.

Radiation tolerance of neutron-irradiated model Fe-Cr-Al alloys

DOE PAGES

Field, Kevin G.; Hu, Xunxiang; Littrell, Kenneth C.; ...

2015-07-14

The Fe Cr Al alloy system has the potential to form an important class of enhanced accident-tolerant cladding materials in the nuclear power industry owing to the alloy system's higher oxidation resistance in high-temperature steam environments compared with traditional zirconium-based alloys. However, radiation tolerance of Fe Cr Al alloys has not been fully established. In this study, a series of Fe Cr Al alloys with 10 18 wt % Cr and 2.9 4.9 wt % Al were neutron irradiated at 382 C to 1.8 dpa to investigate the irradiation-induced microstructural and mechanical property evolution as a function of alloy composition.more » Dislocation loops with Burgers vector of a/2 111 and a 100 were detected and quantified. Results indicate precipitation of Cr-rich is primarily dependent on the bulk chromium composition. Mechanical testing of sub-size-irradiated tensile specimens indicates the hardening response seen after irradiation is dependent on the bulk chromium composition. Furthermore, a structure property relationship was developed; it indicated that the change in yield strength after irradiation is caused by the formation of these radiation-induced defects and is dominated by the large number density of Cr-rich α' precipitates at sufficiently high chromium contents after irradiation.« less

Theoretical stusy of the reaction between 2,2',4' - trihydroxyazobenzene-5-sulfonic acid and zirconium

USGS Publications Warehouse

Fletcher, Mary H.

1960-01-01

Zirconium reacts with 2,2',4'-trihydroxyazobenzene-5-sulfonic acid in acid solutions to Form two complexes in which the ratios of dye to zirconium are 1 to 1 and 2 to 1. Both complexes are true chelates, with zirconium acting as a bridge between the two orthohydroxy dye groups. Apparent equilibrium constants for the reactions to form each of the complexes are determined. The reactions are used as a basis for the determination of the active component in the dye and a graphical method for the determination of reagent purity is described. Four absorption spectra covering the wave length region from 350 to 750 mu are given, which completely define the color system associated with the reactions in solutions where the hydrochloric acid concentration ranges from 0.0064N to about 7N.

Precipitation hardenable iron-nickel-chromium alloy having good swelling resistance and low neutron absorbence

DOEpatents

Korenko, Michael K.; Merrick, Howard F.; Gibson, Robert C.

1980-01-01

An iron-nickel-chromium age-hardenable alloy suitable for use in fast breeder reactor ducts and cladding which utilizes the gamma-double prime strengthening phase and characterized in having a morphology of the gamma-double prime phase enveloping the gamma-prime phase and delta phase distributed at or near the grain boundaries. The alloy consists essentially of about 40-50% nickel, 7.5-14% chromium, 1.5-4% niobium, 0.25-0.75% silicon, 1-3% titanium, 0.1-0.5% aluminum, 0.02-0.1% carbon, 0.002-0.015% boron, and the balance iron. Up to 2% manganese and up to 0.01% magnesium may be added to inhibit trace element effects; up to 0.1% zirconium may be added to increase radiation swelling resistance; and up to 3% molybdenum may be added to increase strength.

Zr-based bulk metallic glass as a cylinder material for high pressure apparatuses

DOE PAGES

Komatsu, Kazuki; Munakata, Koji; Matsubayashi, Kazuyuki; ...

2015-05-12

Zirconium-based bulk metallic glass (Zr-based BMG) has outstanding properties as a cylinder mate- rial for piston-cylinder high pressure apparatuses and is especially useful for neutron scattering. The piston-cylinder consisting of a Zr-based BMG cylinder with outer/inner diameters of 8.8/2.5 mm sustains pressures up to 1.81 GPa and ruptured at 2.0 GPa, with pressure values determined by the superconduct- ing temperature of lead. The neutron attenuation of Zr-based BMG is similar to that of TiZr null-scattering alloy and more transparent than that of CuBe alloy. No contamination of sharp Bragg reflections is observed in the neutron diffraction pattern for Zr-based BMG.more » The magnetic susceptibility of Zr-based BMG is similar to that of CuBe alloy; this leads to a potential application for measurements of magnetic properties under pressure.« less

A promising structure for fabricating high strength and high electrical conductivity copper alloys

PubMed Central

Li, Rengeng; Kang, Huijun; Chen, Zongning; Fan, Guohua; Zou, Cunlei; Wang, Wei; Zhang, Shaojian; Lu, Yiping; Jie, Jinchuan; Cao, Zhiqiang; Li, Tingju; Wang, Tongmin

2016-01-01

To address the trade-off between strength and electrical conductivity, we propose a strategy: introducing precipitated particles into a structure composed of deformation twins. A Cu-0.3%Zr alloy was designed to verify our strategy. Zirconium was dissolved into a copper matrix by solution treatment prior to cryorolling and precipitated in the form of Cu5Zr from copper matrix via a subsequent aging treatment. The microstructure evolutions of the processed samples were investigated by transmission electron microscopy and X-ray diffraction analysis, and the mechanical and physical behaviours were evaluated through tensile and electrical conductivity tests. The results demonstrated that superior tensile strength (602.04 MPa) and electrical conductivity (81.4% IACS) was achieved. This strategy provides a new route for balancing the strength and electrical conductivity of copper alloys, which can be developed for large-scale industrial application. PMID:26856764

Design and fabrication of brayton cycle solar heat receiver

NASA Technical Reports Server (NTRS)

Mendelson, I.

1971-01-01

A detail design and fabrication of a solar heat receiver using lithium fluoride as the heat storage material was completed. A gas flow analysis was performed to achieve uniform flow distribution within overall pressure drop limitations. Structural analyses and allowable design criteria were developed for anticipated environments such as launch, pressure containment, and thermal cycling. A complete heat receiver assembly was fabricated almost entirely from the refractory alloy, niobium-1% zirconium.

Zirconium vanadium chromium alloy

DOEpatents

Mendelsohn, M.H.; Gruen, D.M.

1980-10-14

A ternary intermetallic compound having the formula Zr(V/sub 1-x/Cr/sub x/)/sub 2/ where x is in the range of 0.01 to 0.90 is capable of reversibly sorbing hydrogen at temperatures ranging from room temperature to 200/sup 0/C, at pressures down to 10/sup -6/ torr. The compound is suitable for use as a hydrogen getter in low pressure, high temperature applications such as magnetic confinement fusion devices.

Development in corrosion resistance by microstructural refinement in Zr-16 SS 304 alloy using suction casting technique

DOE Office of Scientific and Technical Information (OSTI.GOV)

Das, N., E-mail: nirupamd@barc.gov.in; Sengupta, P.; Abraham, G.

Highlights: • Grain refinement was made in Zr–16 wt.% SS alloy while prepared by suction casting process. • Distribution of Laves phase, e.g., Zr{sub 2}(Fe, Cr) was raised in suction cast (SC) Zr–16 wt.% SS. • Corrosion resistance was improved in SC alloy compared to that of arc-melt-cast alloy. • Grain refinement in SC alloy assisted for an increase in its corrosion resistance. - Abstract: Zirconium (Zr)-stainless steel (SS) hybrid alloys are being considered as baseline alloys for developing metallic-waste-form (MWF) with the motivation of disposing of Zr and SS base nuclear metallic wastes. Zr–16 wt.% SS, a MWF alloymore » optimized from previous studies, exhibit significant grain refinement and changes in phase assemblages (soft phase: Zr{sub 2}(Fe, Cr)/α-Zr vs. hard phase: Zr{sub 3}(Fe, Ni)) when prepared by suction casting (SC) technique in comparison to arc-cast-melt (AMC) route. Variation in Cr-distribution among different phases are found to be low in suction cast alloy, which along with grain refinement restricted Cr-depletion at the Zr{sub 2}(Fe, Cr)/Zr interfaces, prone to localized attack. Hence, SC alloy, compared to AMC alloy, showed lower current density, higher potential at the breakdown of passivity and higher corrosion potential during polarization experiments (carried out under possible geological repository environments, viz., pH 8, 5 and 1) indicating its superior corrosion resistance.« less

Carbide and carbonitride surface treatment method for refractory metals

DOEpatents

Meyer, Glenn A.; Schildbach, Marcus A.

1996-01-01

A carbide and carbonitride surface treatment method for refractory metals is provided, in steps including, heating a part formed of boron, chromium, hafnium, molybdenum, niobium, tantalum, titanium, tungsten or zirconium, or alloys thereof, in an evacuated chamber and then introducing reaction gases including nitrogen and hydrogen, either in elemental or water vapor form, which react with a source of elemental carbon to form carbon-containing gaseous reactants which then react with the metal part to form the desired surface layer. Apparatus for practicing the method is also provided, in the form of a carbide and carbonitride surface treatment system (10) including a reaction chamber (14), a source of elemental carbon (17), a heating subassembly (20) and a source of reaction gases (23). Alternative methods of providing the elemental carbon (17) and the reaction gases (23) are provided, as well as methods of supporting the metal part (12), evacuating the chamber (14) with a vacuum subassembly (18) and heating all of the components to the desired temperature.

Containerless high temperature property measurements

NASA Technical Reports Server (NTRS)

Nordine, Paul C.; Weber, J. K. Richard; Krishnan, Shankar; Anderson, Collin D.

1991-01-01

Containerless processing in the low gravity environment of space provides the opportunity to increase the temperature at which well controlled processing of and property measurements on materials is possible. This project was directed towards advancing containerless processing and property measurement techniques for application to materials research at high temperatures in space. Containerless high temperature material property studies include measurements of the vapor pressure, melting temperature, optical properties, and spectral emissivities of solid boron. The reaction of boron with nitrogen was also studied by laser polarimetric measurement of boron nitride film growth. The optical properties and spectral emissivities were measured for solid and liquid silicon, niobium, and zirconium; liquid aluminum and titanium; and liquid Ti-Al alloys of 5 to 60 atomic pct. titanium. Alternative means for noncontact temperature measurement in the absence of material emissivity data were evaluated. Also, the application of laser induced fluorescence for component activity measurements in electromagnetic levitated liquids was studied, along with the feasibility of a hybrid aerodynamic electromagnetic levitation technique.

Experimental and theoretical investigation of fatigue life in reusable rocket thrust chambers

NASA Technical Reports Server (NTRS)

Hannum, N. P.; Kasper, H. J.; Pavli, A. J.

1976-01-01

During a test program to investigate low-cycle thermal fatigue, 13 rocket combustion chambers were fabricated and cyclically test fired to failure. Six oxygen-free, high-conductivity (OFHC) copper and seven Amzirc chambers were tested. The failures in the OFHC copper chambers were not typical fatigue failures but are described as creep rupture enhanced by ratcheting. The coolant channels bulged toward the chamber centerline, resulting in progressive thinning of the wall during each cycle. The failures in the Amzirc alloy chambers were caused by low-cycle thermal fatigue. The zirconium in this alloy was not evenly distributed in the chamber materials. The life that was achieved was nominally the same as would have been predicted from OFHC copper isothermal test data.

Quinary metallic glass alloys

DOEpatents

Lin, Xianghong; Johnson, William L.

1998-01-01

At least quinary alloys form metallic glass upon cooling below the glass transition temperature at a rate less than 10.sup.3 K/s. Such alloys comprise zirconium and/or hafnium in the range of 45 to 65 atomic percent, titanium and/or niobium in the range of 4 to 7.5 atomic percent, and aluminum and/or zinc in the range of 5 to 15 atomic percent. The balance of the alloy compositions comprise copper, iron, and cobalt and/or nickel. The composition is constrained such that the atomic percentage of iron is less than 10 percent. Further, the ratio of copper to nickel and/or cobalt is in the range of from 1:2 to 2:1. The alloy composition formula is: (Zr,Hf).sub.a (Al,Zn).sub.b (Ti,Nb).sub.c (Cu.sub.x Fe.sub.y (Ni,Co).sub.z).sub.d wherein the constraints upon the formula are: a ranges from 45 to 65 atomic percent, b ranges from 5 to 15 atomic percent, c ranges from 4 to 7.5 atomic percent, d comprises the balance, d.multidot.y is less than 10 atomic percent, and x/z ranges from 0.5 to 2.

Hydrogen permeation in FeCrAl alloys for LWR cladding application

NASA Astrophysics Data System (ADS)

Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; Snead, Lance L.

2015-06-01

FeCrAl, an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In this study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. The total tritium inventory inside the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.

Quinary metallic glass alloys

DOEpatents

Lin, X.; Johnson, W.L.

1998-04-07

At least quinary alloys form metallic glass upon cooling below the glass transition temperature at a rate less than 10{sup 3}K/s. Such alloys comprise zirconium and/or hafnium in the range of 45 to 65 atomic percent, titanium and/or niobium in the range of 4 to 7.5 atomic percent, and aluminum and/or zinc in the range of 5 to 15 atomic percent. The balance of the alloy compositions comprise copper, iron, and cobalt and/or nickel. The composition is constrained such that the atomic percentage of iron is less than 10 percent. Further, the ratio of copper to nickel and/or cobalt is in the range of from 1:2 to 2:1. The alloy composition formula is: (Zr,Hf){sub a}(Al,Zn){sub b}(Ti,Nb){sub c}(Cu{sub x}Fe{sub y}(Ni,Co){sub z}){sub d} wherein the constraints upon the formula are: a ranges from 45 to 65 atomic percent, b ranges from 5 to 15 atomic percent, c ranges from 4 to 7.5 atomic percent, d comprises the balance, d{hor_ellipsis}y is less than 10 atomic percent, and x/z ranges from 0.5 to 2.

Solution Treatment Effect on Tensile, Impact and Fracture Behaviour of Trace Zr Added Al-12Si-1Mg-1Cu Piston Alloy

NASA Astrophysics Data System (ADS)

Kaiser, Md. Salim

2018-04-01

The effects of T6 solution treatment on tensile, impact and fracture properties of cast Al-12Si-1Mg-1Cu piston alloys with trace of zirconium were investigated. Cast alloys were given precipitation strengthening treatment having a sequence of homogenizing, solutionizing, quenching and ageing. Both cast and solutionized samples are isochronally aged for 90 min at different temperatures up to 300 °C. Tensile and impact properties of the differently processed alloys have been studied to understand the precipitation strengthening of the alloys. Fractograpy of the alloys were observed to understand the mode of fracture. It is observed that the improvement in tensile properties in the aged alloys through heat treatment is mainly attributed to the formation of the Al2Cu and Mg2Si precipitates within the Al matrix. Solution treatment improves the tensile strength for the reason that during solution treatment some alloying elements are re-dissolved to produce a solute-rich solid solution. Impact energy decreases with ageing temperature due to formation of GP zones, β' and β precipitates. The fractography shows large and small dimple structure and broken or cracked primary Si, particles. Microstructure study of alloys revealed that the solution treatment improved distribution of silicon grains. The addition of Zr produces an improvement in the tensile properties as a result of its grain refining action and grain coarsening resistance in the matrix at a higher temperature.

Exploratory Investigation of Advanced-Temperature Nickel-Base Alloys

NASA Technical Reports Server (NTRS)

Freche, John C.; Waters, William J.

1959-01-01

An investigation was conducted to provide an advanced-temperature nickel-base alloy with properties suitable for aircraft turbine blades as well as for possible space vehicle applications. An entire series of alloys that do not require vacuum melting techniques and that generally provide good stress-rupture and impact properties was evolved. The basic-alloy composition of 79 percent nickel, 8 percent molybdenum, 6 percent chromium, 6 percent aluminum, and 1 percent zirconium was modified by a series of element additions such as carbon, titanium, and boron, with the nickel content adjusted to account for the additives. Stress-rupture, impact, and swage tests were made with all the alloys. The strongest composition (basic alloy plus 1.5 percent titanium plus 0.125 percent carbon) displayed 384- and 574-hour stress-rupture lives at 1800 F and 15,000 psi in the as-cast and homogenized conditions, respectively. All the alloys investigated demonstrated good impact resistance. Several could not be broken in a low-capacity Izod impact tester and, on this basis, all compared favorably with several high-strength high-temperature alloys. Swaging cracks were encountered with all the alloys. In several cases, however, these cracks were slight and could be detected only by zyglo examination. Some of these compositions may become amenable to hot working on further development. On the basis of the properties indicated, it appears that several of the alloys evolved, particularly the 1.5 percent titanium plus 0.125 percent carbon basic-alloy modification, could be used for advanced- temperature turbine blades, as well as for possible space vehicle applications.

40 CFR 421.333 - Effluent limitations guidelines representing the degree of effluent reduction attainable by the...

Code of Federal Regulations, 2013 CFR

2013-07-01

... 2.387 1.108 Nickel 4.688 3.154 Ammonia (as N) 1,136.000 499.500 (d) SiCl4 purification wet air....262 Lead 4.416 2.050 Nickel 8.674 5.835 Ammonia (as N) 2,102.000 895.000 (q) Leaching rinse water from... Nickel 32.410 21.810 Ammonia (as N) 7,856.000 3,453.000 (r) Leaching rinse water from zirconium alloy...

40 CFR 421.333 - Effluent limitations guidelines representing the degree of effluent reduction attainable by the...

Code of Federal Regulations, 2012 CFR

2012-07-01

... 2.387 1.108 Nickel 4.688 3.154 Ammonia (as N) 1,136.000 499.500 (d) SiCl4 purification wet air....262 Lead 4.416 2.050 Nickel 8.674 5.835 Ammonia (as N) 2,102.000 895.000 (q) Leaching rinse water from... Nickel 32.410 21.810 Ammonia (as N) 7,856.000 3,453.000 (r) Leaching rinse water from zirconium alloy...

40 CFR 421.333 - Effluent limitations guidelines representing the degree of effluent reduction attainable by the...

Code of Federal Regulations, 2014 CFR

2014-07-01

... 2.387 1.108 Nickel 4.688 3.154 Ammonia (as N) 1,136.000 499.500 (d) SiCl4 purification wet air....262 Lead 4.416 2.050 Nickel 8.674 5.835 Ammonia (as N) 2,102.000 895.000 (q) Leaching rinse water from... Nickel 32.410 21.810 Ammonia (as N) 7,856.000 3,453.000 (r) Leaching rinse water from zirconium alloy...

The interface in tungsten fiber reinforced niobium metal-matrix composites. Final Report Ph.D. Thesis - Case Western Reserve Univ., Cleveland, OH

NASA Technical Reports Server (NTRS)

Grobstein, Toni L.

1989-01-01

The creep resistance of tungsten fiber reinforced niobium metal-matrix composites was evaluated. The interface region between the fiber and matrix was characterized by microhardness and electron probe microanalysis measurements which indicated that its properties were between those of fiber and matrix. However, the measured properties of the composite exceeded those calculated by the rule of mixtures even when the interface zone was assumed to retain all the strength of the fiber. The composite structure appeared to enhance the strengths of both the fibers and the matrix above what they exhibited in stand-alone tests. The effect of fiber orientation and matrix alloy composition on the fiber/matrix interface were also evaluated. Small alloying additions of zirconium and tungsten to the niobium matrix affected the creep resistance of the composites only slightly. A decrease in the creep resistance of the composite with increasing zirconium content in the matrix was ascribed to an increase in the diffusion rate of the fiber/matrix interdiffusion reaction, and a slight increase in the creep resistance of the composite was observed with an addition of 9 w percent tungsten to the matrix. In addition, Kirkendall void formation was observed at the fiber/matrix interface; the void distribution differed depending on the fiber orientation relative to the stress axis.

Results of NDE Technique Evaluation of Clad Hydrides

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kunerth, Dennis C.

2014-09-01

This report fulfills the M4 milestone, M4FT-14IN0805023, Results of NDE Technique Evaluation of Clad Hydrides, under Work Package Number FT-14IN080502. During service, zirconium alloy fuel cladding will degrade via corrosion/oxidation. Hydrogen, a byproduct of the oxidation process, will be absorbed into the cladding and eventually form hydrides due to low hydrogen solubility limits. The hydride phase is detrimental to the mechanical properties of the cladding and therefore it is important to be able to detect and characterize the presence of this constituent within the cladding. Presently, hydrides are evaluated using destructive examination. If nondestructive evaluation techniques can be used tomore » detect and characterize the hydrides, the potential exists to significantly increase test sample coverage while reducing evaluation time and cost. To demonstrate the viability this approach, an initial evaluation of eddy current and ultrasonic techniques were performed to demonstrate the basic ability to these techniques to detect hydrides or their effects on the microstructure. Conventional continuous wave eddy current techniques were applied to zirconium based cladding test samples thermally processed with hydrogen gas to promote the absorption of hydrogen and subsequent formation of hydrides. The results of the evaluation demonstrate that eddy current inspection approaches have the potential to detect both the physical damage induced by hydrides, e.g. blisters and cracking, as well as the combined effects of absorbed hydrogen and hydride precipitates on the electrical properties of the zirconium alloy. Similarly, measurements of ultrasonic wave velocities indicate changes in the elastic properties resulting from the combined effects of absorbed hydrogen and hydride precipitates as well as changes in geometry in regions of severe degradation. However, for both approaches, the signal responses intended to make the desired measurement incorporate a number of contributing parameters. These contributing factors need to be recognized and a means to control them or separate their contributions will be required to obtain the desired information.« less

Enhanced antibacterial properties, biocompatibility, and corrosion resistance of degradable Mg-Nd-Zn-Zr alloy.

PubMed

Qin, Hui; Zhao, Yaochao; An, Zhiquan; Cheng, Mengqi; Wang, Qi; Cheng, Tao; Wang, Qiaojie; Wang, Jiaxing; Jiang, Yao; Zhang, Xianlong; Yuan, Guangyin

2015-06-01

Magnesium (Mg), a potential biodegradable material, has recently received increasing attention due to its unique antibacterial property. However, rapid corrosion in the physiological environment and potential toxicity limit clinical applications. In order to improve the corrosion resistance meanwhile not compromise the antibacterial activity, a novel Mg alloy, Mg-Nd-Zn-Zr (Hereafter, denoted as JDBM), is fabricated by alloying with neodymium (Nd), zinc (Zn), zirconium (Zr). pH value, Mg ion concentration, corrosion rate and electrochemical test show that the corrosion resistance of JDBM is enhanced. A systematic investigation of the in vitro and in vivo antibacterial capability of JDBM is performed. The results of microbiological counting, CLSM, SEM in vitro, and microbiological cultures, histopathology in vivo consistently show JDBM enhanced the antibacterial activity. In addition, the significantly improved cytocompatibility is observed from JDBM. The results suggest that JDBM effectively enhances the corrosion resistance, biocompatibility and antimicrobial properties of Mg by alloying with the proper amount of Zn, Zr and Nd. Copyright © 2015 Elsevier Ltd. All rights reserved.

Effects of GlidArc plasma treatment on metallic surface

NASA Astrophysics Data System (ADS)

Astanei, D.; Ursache, M.; Hnatiuc, E.; Stoica, I.; Hnatiuc, B.; Felea, C.

2016-12-01

This paper presents the GlidArc plasma effects on some metallic surfaces often used in dentistry: zirconium, titanium and nickel - chromium alloy plates. For the experiments performed, a GlidArc reactor with two planar electrodes has been used. During the tests, the gas flow has been kept constant while the treatment time and the distance between the plasma and the sample were modified. The surfaces were analyzed using atomic force microscopy (AFM) in order to determine the surface morphological modifications induced by the plasma treatment.

Biocompatible Zr-Al-Fe bulk metallic glasses with large plasticity

NASA Astrophysics Data System (ADS)

Hua, NengBin; Li, Ran; Wang, JianFeng; Zhang, Tao

2012-09-01

In the present study, high-zirconium ternary Zr-Al-Fe bulk metallic glasses (BMGs) with low Young's modulus and good plasticity were developed. Zr75Al7.5Fe17.5 BMG exhibits a low Young's modulus of 70 GPa and high Poisson's ratio of 0.403. Pronounced plasticity was demonstrated under both compression and bending conditions for the BMGs. Furthermore, the alloys show high corrosion resistance in phosphate buffered solution. The combination of desirable mechanical and chemical properties implies potential for biomedical applications.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tan, Lizhen; Yang, Ying; Tyburska-Puschel, Beata

The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools)more » is an important path to more efficient alloy development and process optimization. Ferritic-martensitic (FM) steels are important structural materials for nuclear reactors due to their advantages over other applicable materials like austenitic stainless steels, notably their resistance to void swelling, low thermal expansion coefficients, and higher thermal conductivity. However, traditional FM steels exhibit a noticeable yield strength reduction at elevated temperatures above ~500°C, which limits their applications in advanced nuclear reactors which target operating temperatures at 650°C or higher. Although oxide-dispersion-strengthened (ODS) ferritic steels have shown excellent high-temperature performance, their extremely high cost, limited size and fabricability of products, as well as the great difficulty with welding and joining, have limited or precluded their commercial applications. Zirconium has shown many benefits to Fe-base alloys such as grain refinement, improved phase stability, and reduced radiation-induced segregation. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of a new generation of Zr-bearing ferritic alloys to be fabricated using conventional steelmaking practices, which have excellent radiation resistance and enhanced high-temperature creep performance greater than Grade 91.« less

Zirconium-nickel crystals—hydrogen accumulators: Dissolution and penetration of hydrogen atoms in alloys

NASA Astrophysics Data System (ADS)

Matysina, Z. A.; Zaginaichenko, S. Yu.; Shchur, D. V.; Gabdullin, M. T.; Kamenetskaya, E. A.

2016-07-01

The calculation of the free energy, thermodynamic equilibrium equations, and kinetic equations of the intermetallic compound Zr2NiH x has been carried out based on molecular-kinetic concepts. The equilibrium hydrogen concentration depending on the temperature, pressure, and energy parameters has been calculated. The absorption-desorption of hydrogen has been studied, and the possibility of the realization of the hysteresis effect has been revealed. The kinetics of the dissolution and permeability of hydrogen is considered, the time dependence of these values has been found, and conditions for the extremum character of their time dependence have been determined. Relaxation times of the dissolution and permeability of hydrogen into the alloy have been calculated. The calculation results are compared with the experimental data available in the literature.

Computational and Experimental Studies of Microstructure-Scale Porosity in Metallic Fuels for Improved Gas Swelling Behavior

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mllett, Paul; McDeavitt, Sean; Deo, Chaitanya

This proposal will investigate the stability of bimodal pore size distributions in metallic uranium and uranium-zirconium alloys during sintering and re-sintering annealing treatments. The project will utilize both computational and experimental approaches. The computational approach includes both Molecular Dynamics simulations to determine the self-diffusion coefficients in pure U and U-Zr alloys in single crystals, grain boundaries, and free surfaces, as well as calculations of grain boundary and free surface interfacial energies. Phase-field simulations using MOOSE will be conducted to study pore and grain structure evolution in microstructures with bimodal pore size distributions. Experiments will also be performed to validate themore » simulations, and measure the time-dependent densification of bimodal porous compacts.« less

Development of Zirconium-based Conversion Coatings for the Pretreatment of AZ91D Magnesium Alloy Prior to Electrocoating

NASA Astrophysics Data System (ADS)

Reck, James; Wang, Yar-Ming; Kuo, Hong-Hsiang Harry

This work examines the use of hexafluorozirconic acid based solutions at concentrations from 0.025 M to 0.100 M and pH values of 2.0 to 4.0 for the creation of a zirconia-based conversion coating less than 1 micron thick to protect magnesium alloy AZ91D. Similar coatings have been found to give excellent protection for steel and aluminum alloys, but little research has been conducted on its application to magnesium. Work was performed to gain an understanding of the film formation mechanisms and related kinetics using x-ray photo-electron spectroscopy, scanning electron microscopy, and open circuit potential monitoring techniques. A design of experiments approach was taken to determine the effects of acid concentration, pH, and soak time on the corrosion properties both as-deposited and with an application of electrocoat. It was found that the application of the zirconia-based coating significantly increased corrosion resistance, and allowed for an acceptable e-coat application with excellent adherence.

Implications of Zircaloy creep and growth to light water reactor performance

NASA Astrophysics Data System (ADS)

Franklin, David G.; Adamson, Ronald B.

1988-10-01

Deformation of zirconium alloy components in nuclear reactors has been a concern since the decision of Admiral Rickover to use them in the US Navy submarine reactors. With the exception of the first few light water reactors (LWRs) most of the core structural materials have been fabricated from either Zircaloy-2 or Zircaloy-4. Performance of these alloys has been extremely good, even though the effects of irradiation on deformation magnitudes and mechanisms were not fully appreciated until extensive service and in-reactor tests were accomplished. Since the reactor components are designed to operate at stress levels well below yield for normal conditions, the only significant deformation is time dependent. Although creep was anticipated, the enhancement by neutron irradiation and the stress-free, nearly constant-volume shape change known as irradiation growth were not known prior to materials testing in reactors under controlled conditions. Both of these phenomena have significant impact on performance and must be accounted for properly in design. Although irradiation creep and growth have resulted in only one significant performance problem (creep collapse of fuel cladding, which has been eliminated), deformation magnitudes and, particularly, differentials in strain magnitudes, are a continuing source of interest. Factors that affect dimensional stability due to both creep and growth include temperature, fluence, residual stress, texture, and microstructure. The first two are reactor variables and the others are related to component fabrication history. This paper includes a review of the applications of Zircaloy creep and growth to LWR fuel designs, a review of the impact of in-reactor creep and growth on fuel rod and fuel assembly performance, and comments on potential improvements. Since the reactor design, fuel design and the core environment in BWRs and PWRs are quite different, appropriate separation of the application of effects are made; of course, the basic phenomena are the same in both systems.

Development of binary and ternary titanium alloys for dental implants.

PubMed

Cordeiro, Jairo M; Beline, Thamara; Ribeiro, Ana Lúcia R; Rangel, Elidiane C; da Cruz, Nilson C; Landers, Richard; Faverani, Leonardo P; Vaz, Luís Geraldo; Fais, Laiza M G; Vicente, Fabio B; Grandini, Carlos R; Mathew, Mathew T; Sukotjo, Cortino; Barão, Valentim A R

2017-11-01

The aim of this study was to develop binary and ternary titanium (Ti) alloys containing zirconium (Zr) and niobium (Nb) and to characterize them in terms of microstructural, mechanical, chemical, electrochemical, and biological properties. The experimental alloys - (in wt%) Ti-5Zr, Ti-10Zr, Ti-35Nb-5Zr, and Ti-35Nb-10Zr - were fabricated from pure metals. Commercially pure titanium (cpTi) and Ti-6Al-4V were used as controls. Microstructural analysis was performed by means of X-ray diffraction and scanning electron microscopy. Vickers microhardness, elastic modulus, dispersive energy spectroscopy, X-ray excited photoelectron spectroscopy, atomic force microscopy, surface roughness, and surface free energy were evaluated. The electrochemical behavior analysis was conducted in a body fluid solution (pH 7.4). The albumin adsorption was measured by the bicinchoninic acid method. Data were evaluated through one-way ANOVA and the Tukey test (α=0.05). The alloying elements proved to modify the alloy microstructure and to enhance the mechanical properties, improving the hardness and decreasing the elastic modulus of the binary and ternary alloys, respectively. Ti-Zr alloys displayed greater electrochemical stability relative to that of controls, presenting higher polarization resistance and lower capacitance. The experimental alloys were not detrimental to albumin adsorption. The experimental alloys are suitable options for dental implant manufacturing, particularly the binary system, which showed a better combination of mechanical and electrochemical properties without the presence of toxic elements. Copyright © 2017 The Academy of Dental Materials. Published by Elsevier Ltd. All rights reserved.

Edge-on dislocation loop in anisotropic hcp zirconium thin foil

NASA Astrophysics Data System (ADS)

Wu, Wenwang; Xia, Re; Qian, Guian; Xu, Shucai; Zhang, Jinhuan

2015-10-01

Edge-on dislocation loops with 〈 a 〉 -type and 〈 c 〉 -type of Burgers vectors can be formed on prismatic or basel habit planes of hexagonal close-packed (hcp) zirconium alloys during in-situ ion irradiation and neutron irradiation experiments. In this work, an anisotropic image stress method was employed to analyze the free surface effects of dislocation loops within hcp Zr thin foils. Calculation results demonstrate that image stress has a remarkable effect on the distortion fields of dislocation loops within infinite medium, and the image energy becomes remarkable when dislocation loops are situated close to the free surfaces. Moreover, image forces of the 1 / 2 〈 0001 〉 (0001) dislocation loop within (0001) thin foil is much stronger than that of the 1 / 3 〈 11 2 bar 0 〉 (11 2 bar 0) dislocation loop within (11 2 bar 0) thin foil of identical geometrical configurations. Finally, image stress effect on the physical behaviors of loops during in-situ ion irradiation experiments is discussed.

Determination of very low concentrations of hydrogen in zirconium alloys by neutron imaging

NASA Astrophysics Data System (ADS)

Buitrago, N. L.; Santisteban, J. R.; Tartaglione, A.; Marín, J.; Barrow, L.; Daymond, M. R.; Schulz, M.; Grosse, M.; Tremsin, A.; Lehmann, E.; Kaestner, A.; Kelleher, J.; Kabra, S.

2018-05-01

Zr-based alloys are used in nuclear power plants because of a unique combination of very low neutron absorption and excellent mechanical properties and corrosion resistance at operating conditions. However, Hydrogen (H) or Deuterium ingress due to waterside corrosion during operation can embrittle these materials. In particular, Zr alloys are affected by Delayed Hydride Cracking (DHC), a stress-corrosion cracking mechanism operating at very low H content (∼100-300 wt ppm), which involves the diffusion of H to the crack tip. H content in Zr alloys is commonly determined by destructive techniques such as inert gas fusion and vacuum extraction. In this work, we have used neutron imaging to non-destructively quantify the spatial distribution of H in Zr alloys specimens with a resolution of ∼5 wt ppm, an accuracy of ∼10 wt ppm and a spatial resolution of ∼25 μm × 5 mm x 10 mm. Non-destructive experiments performed on a comprehensive set of calibrated specimens of Zircaloy-2 and Zr2.5%Nb at four neutron facilities worldwide show the typical precision and repeatability of the technique. We have observed that the microstructure of the alloy plays an important role on the homogeneity of H across a specimen. We propose several strategies for performing H determinations without calibrated specimens, with the most precise results for neutrons having wavelengths longer than 5.7 Å.

Nanoparticle Addition to Enhance the Mechanical Response of Magnesium Alloys Including Nanoscale Deformation Mechanisms

NASA Astrophysics Data System (ADS)

Paramsothy, Muralidharan; Gupta, Manoj

In this study, various magnesium alloy nanocomposites derived from AZ (Aluminium-Zinc) or ZK (Zinc-Zirconium) series matrices and containing Al2O3, Si3N4, TiC or carbon nanotube (CNT) nanoparticle reinforcement (representative oxide, nitride, carbide or carbon nanoparticle reinforcement, respectively) were fabricated using solidification processing followed by hot extrusion. The main aim here was to simultaneously enhance tensile strength and ductility of each alloy using nanoparticles. The magnesium-oxygen strong affinity and magnesium-carbon weak affinity (comparison of extremes in affinity) are both well known in the context of magnesium composite processing. However, an approach to possibly quantify this affinity in magnesium nanocomposite processing is not clear. In this study accordingly, Nanoscale Electro Negative Interface Density or NENID quantifies the nanoparticle-alloy matrix interfacial area per unit volume in the magnesium alloy nanocomposite taking into consideration the electronegativity of the nanoparticle reinforcement. The beneficial (as well as comparative) effect of the nanoparticles on each alloy is discussed in this article. Regarding the mechanical performance of the nanocomposites, it is important to understand the experimentally observed nanoparticle-matrix interactions during plastic deformation (nanoscale deformation mechanisms). Little is known in this area based on direct observations for metal matrix nanocomposites. Here, relevant multiple nanoscale phenomena includes the emanation of high strain zones (HSZs) from nanoparticle surfaces.

Hydrogen permeation in FeCrAl alloys for LWR cladding application

DOE PAGES

Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; ...

2015-03-19

FeCrAl is an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In our study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. Also, the total tritium inventory insidemore » the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.« less

Impact of neutron irradiation on mechanical performance of FeCrAl alloy laser-beam weldments

NASA Astrophysics Data System (ADS)

Gussev, M. N.; Cakmak, E.; Field, K. G.

2018-06-01

Oxidation-resistant iron-chromium-aluminum (FeCrAl) alloys demonstrate better performance in Loss-of-Coolant Accidents, compared with austenitic- and zirconium-based alloys. However, further deployment of FeCrAl-based materials requires detailed characterization of their performance under irradiation; moreover, since welding is one of the key operations in fabrication of light water reactor fuel cladding, FeCrAl alloy weldment performance and properties also should be determined prior to and after irradiation. Here, advanced C35M alloy (Fe-13%Cr-5%Al) and variants with aluminum (+2%) or titanium carbide (+1%) additions were characterized after neutron irradiation in Oak Ridge National Laboratory's High Flux Isotope Reactor at 1.8-1.9 dpa in a temperature range of 195-559 °C. Specimen sets included as-received (AR) materials and specimens after controlled laser-beam welding. Tensile tests with digital image correlation (DIC), scanning electron microscopy-electron back scatter diffraction analysis, fractography, and x-ray tomography analysis were performed. DIC allowed for investigating local yield stress in the weldments, deformation hardening behavior, and plastic anisotropy. Both AR and welded material revealed a high degree of radiation-induced hardening for low-temperature irradiation; however, irradiation at high-temperatures (i.e., 559 °C) had little overall effect on the mechanical performance.

DEVELOPMENT OF PLASTICITY MODEL USING NON ASSOCIATED FLOW RULE FOR HCP MATERIALS INCLUDING ZIRCONIUM FOR NUCLEAR APPLICATIONS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael V. Glazoff; Jeong-Whan Yoon

2013-08-01

In this report (prepared in collaboration with Prof. Jeong Whan Yoon, Deakin University, Melbourne, Australia) a research effort was made to develop a non associated flow rule for zirconium. Since Zr is a hexagonally close packed (hcp) material, it is impossible to describe its plastic response under arbitrary loading conditions with any associated flow rule (e.g. von Mises). As a result of strong tension compression asymmetry of the yield stress and anisotropy, zirconium displays plastic behavior that requires a more sophisticated approach. Consequently, a new general asymmetric yield function has been developed which accommodates mathematically the four directional anisotropies alongmore » 0 degrees, 45 degrees, 90 degrees, and biaxial, under tension and compression. Stress anisotropy has been completely decoupled from the r value by using non associated flow plasticity, where yield function and plastic potential have been treated separately to take care of stress and r value directionalities, respectively. This theoretical development has been verified using Zr alloys at room temperature as an example as these materials have very strong SD (Strength Differential) effect. The proposed yield function reasonably well models the evolution of yield surfaces for a zirconium clock rolled plate during in plane and through thickness compression. It has been found that this function can predict both tension and compression asymmetry mathematically without any numerical tolerance and shows the significant improvement compared to any reported functions. Finally, in the end of the report, a program of further research is outlined aimed at constructing tensorial relationships for the temperature and fluence dependent creep surfaces for Zr, Zircaloy 2, and Zircaloy 4.« less

Trunnion Failure of the Recalled Low Friction Ion Treatment Cobalt Chromium Alloy Femoral Head.

PubMed

Urish, Kenneth L; Hamlin, Brian R; Plakseychuk, Anton Y; Levison, Timothy J; Higgs, Genymphas B; Kurtz, Steven M; DiGioia, Anthony M

2017-09-01

Gross trunnion failure (GTF) is a rare complication in total hip arthroplasty (THA) reported across a range of manufacturers. Specific lots of the Stryker low friction ion treatment (LFIT) anatomic cobalt chromium alloy (CoCr) V40 femoral head were recalled in August 2016. In part, the recall was based out of concerns for disassociation of the femoral head from the stem and GTF. We report on 28 patients (30 implants) with either GTF (n = 18) or head-neck taper corrosion (n = 12) of the LFIT CoCr femoral head and the Accolade titanium-molybdenum-zirconium-iron alloy femoral stems. All these cases were associated with adverse local tissue reactions requiring revision of the THA. In our series, a conservative estimate of the incidence of failure was 4.7% (n = 636 total implanted) at 8.0 ± 1.4 years from the index procedure. Failures were associated with a high-offset 127° femoral stem neck angle and increased neck lengths; 43.3% (13 of 30) of the observed failures included implant sizes outside the voluntary recall (27.8% [5 of 18] of the GTF and 75.0% [8 of 12] of the taper corrosion cases). Serum cobalt and chromium levels were elevated (cobalt: 8.4 ± 7.0 μg/mL; chromium: 3.4 ± 3.3 μ/L; cobalt/chromium ratio: 3.7). The metal artifact reduction sequence magnetic resonance imaging demonstrated large cystic fluid collections typical with adverse local tissue reactions. During revision, a pseudotumor was observed in all cases. Pathology suggested a chronic inflammatory response. Impending GTF could be diagnosed based on aspiration of black synovial fluid and an oblique femoral head as compared with the neck taper on radiographs. In our series of the recalled LFIT CoCr femoral head, the risk of impending GTF or head-neck taper corrosion should be considered as a potential diagnosis in a painful LFIT femoral head and Accolade titanium-molybdenum-zirconium-iron alloy THA with unknown etiology. Almost half of the failures we observed included sizes outside of the voluntary recall. Copyright © 2017 Elsevier Inc. All rights reserved.

Effects of Zr, Ti, and Al Additions on Nonmetallic Inclusions and Impact Toughness of Cast Low-Alloy Steel

NASA Astrophysics Data System (ADS)

Bizyukov, Pavel V.; Giese, Scott R.

2017-04-01

A microalloying of the low-carbon and low-alloy cast steel was conducted with Zr, Ti, and Al that were added to the steel in four combinations. After heat treatment, the samples were tested for impact toughness at room temperature using the Charpy method. The highest values of impact toughness were obtained in the group treated with Zr, while Zr-Ti and Zr-Ti-Al groups showed moderate toughness values; the lowest values were observed in the Zr-Al group. Difference among the treatment groups was observed in the fracture mechanisms, morphology, and area distribution of the inclusions. High toughness values achieved in the trials treated with zirconium corresponded with smooth ductile fracture. The metal treated with a combination of zirconium and titanium had a relatively small area occupied by inclusions, but its toughness was also moderate. Lowest impact toughness values corresponded with the larger area occupied by the inclusions in the trials treated with aluminum. Also, a connection between the solubility product [Al][N] and impact toughness was established. The study also provides a qualitative description and quantitative analysis of the nonmetallic inclusions formation as a result of microalloying treatment. The precipitation sequence of the inclusions was described based on the thermochemical calculations for the nonmetallic compounds discovered in the experimental steel. A description of the size distribution, morphology, and composition was conducted for the oxides, nitrides, sulfides, and multiphase particles.

Hydrogen isotope separation utilizing bulk getters

DOEpatents

Knize, R.J.; Cecchi, J.L.

1991-08-20

Tritium and deuterium are separated from a gaseous mixture thereof, derived from a nuclear fusion reactor or some other source, by providing a casing with a bulk getter therein for absorbing the gaseous mixture to produce an initial loading of the getter, partially desorbing the getter to produce a desorbed mixture which is tritium-enriched, pumping the desorbed mixture into a separate container, the remaining gaseous loading in the getter being deuterium-enriched, desorbing the getter to a substantially greater extent to produce a deuterium-enriched gaseous mixture, and removing the deuterium-enriched mixture into another container. The bulk getter may comprise a zirconium-aluminum alloy, or a zirconium-vanadium-iron alloy. The partial desorption may reduce the loading by approximately fifty percent. The basic procedure may be extended to produce a multistage isotope separator, including at least one additional bulk getter into which the tritium-enriched mixture is absorbed. The second getter is then partially desorbed to produce a desorbed mixture which is further tritium-enriched. The last-mentioned mixture is then removed from the container for the second getter, which is then desorbed to a substantially greater extent to produce a desorbed mixture which is deuterium-enriched. The last-mentioned mixture is then removed so that the cycle can be continued and repeated. The method of isotope separation is also applicable to other hydrogen isotopes, in that the method can be employed for separating either deuterium or tritium from normal hydrogen. 4 figures.

Hydrogen isotope separation utilizing bulk getters

DOEpatents

Knize, Randall J.; Cecchi, Joseph L.

1991-01-01

Tritium and deuterium are separated from a gaseous mixture thereof, derived from a nuclear fusion reactor or some other source, by providing a casing with a bulk getter therein for absorbing the gaseous mixture to produce an initial loading of the getter, partially desorbing the getter to produce a desorbed mixture which is tritium-enriched, pumping the desorbed mixture into a separate container, the remaining gaseous loading in the getter being deuterium-enriched, desorbing the getter to a substantially greater extent to produce a deuterium-enriched gaseous mixture, and removing the deuterium-enriched mixture into another container. The bulk getter may comprise a zirconium-aluminum alloy, or a zirconium-vanadium-iron alloy. The partial desorption may reduce the loading by approximately fifty percent. The basic procedure may be extended to produce a multistage isotope separator, including at least one additional bulk getter into which the tritium-enriched mixture is absorbed. The second getter is then partially desorbed to produce a desorbed mixture which is further tritium-enriched. The last-mentioned mixture is then removed from the container for the second getter, which is then desorbed to a substantially greater extent to produce a desorbed mixture which is deuterium-enriched. The last-mentioned mixture is then removed so that the cycle can be continued and repeated. The method of isotope separation is also applicable to other hydrogen isotopes, in that the method can be employed for separating either deuterium or tritium from normal hydrogen.

Hydrogen isotope separation utilizing bulk getters

DOEpatents

Knize, Randall J.; Cecchi, Joseph L.

1990-01-01

Tritium and deuterium are separated from a gaseous mixture thereof, derived from a nuclear fusion reactor or some other source, by providing a casing with a bulk getter therein for absorbing the gaseous mixture to produce an initial loading of the getter, partially desorbing the getter to produce a desorbed mixture which is tritium-enriched, pumping the desorbed mixture into a separate container, the remaining gaseous loading in the getter being deuterium-enriched, desorbing the getter to a substantially greater extent to produce a deuterium-enriched gaseous mixture, and removing the deuterium-enriched mixture into another container. The bulk getter may comprise a zirconium-aluminum alloy, or a zirconium-vanadium-iron alloy. The partial desorption may reduce the loading by approximately fifty percent. The basic procedure may be extended to produce a multistage isotope separator, including at least one additional bulk getter into which the tritium-enriched mixture is absorbed. The second getter is then partially desorbed to produce a desorbed mixture which is further tritium-enriched. The last-mentioned mixture is then removed from the container for the second getter, which is then desorbed to a substantially greater extent to produce a desorbed mixture which is deuterium-enriched. The last-mentioned mixture is then removed so that the cycle can be continued and repeated. The method of isotope separation is also applicable to other hydrogen isotopes, in that the method can be employed for separating either deuterium or tritium from normal hydrogen.

Clinical evidence on titanium-zirconium dental implants: a systematic review and meta-analysis.

PubMed

Altuna, P; Lucas-Taulé, E; Gargallo-Albiol, J; Figueras-Álvarez, O; Hernández-Alfaro, F; Nart, J

2016-07-01

The use of titanium implants is well documented and they have high survival and success rates. However, when used as reduced-diameter implants, the risk of fracture is increased. Narrow diameter implants (NDIs) of titanium-zirconium (Ti-Zr) alloy have recently been developed (Roxolid; Institut Straumann AG). Ti-Zr alloys (two highly biocompatible materials) demonstrate higher tensile strength than commercially pure titanium. The aim of this systematic review was to summarize the existing clinical evidence on dental NDIs made from Ti-Zr. A systematic literature search was performed using the Medline database to find relevant articles on clinical studies published in the English language up to December 2014. Nine clinical studies using Ti-Zr implants were identified. Overall, 607 patients received 922 implants. The mean marginal bone loss was 0.36±0.06mm after 1 year and 0.41±0.09mm after 2 years. The follow-up period ranged from 3 to 36 months. Mean survival and success rates were 98.4% and 97.8% at 1 year after implant placement and 97.7% and 97.3% at 2 years. Narrow diameter Ti-Zr dental implants show survival and success rates comparable to regular diameter titanium implants (>95%) in the short term. Long-term follow-up clinical data are needed to confirm the excellent clinical performance of these implants. Copyright © 2016 International Association of Oral and Maxillofacial Surgeons. Published by Elsevier Ltd. All rights reserved.

Quantitative in vivo biocompatibility of new ultralow-nickel cobalt-chromium-molybdenum alloys.

PubMed

Sonofuchi, Kazuaki; Hagiwara, Yoshihiro; Koizumi, Yuichiro; Chiba, Akihiko; Kawano, Mitsuko; Nakayama, Masafumi; Ogasawara, Kouetsu; Yabe, Yutaka; Itoi, Eiji

2016-09-01

Nickel (Ni) eluted from metallic biomaterials is widely accepted as a major cause of allergies and inflammation. To improve the safety of cobalt-chromium-molybdenum (Co-Cr-Mo) alloy implants, new ultralow-Ni Co-Cr-Mo alloys with and without zirconium (Zr) have been developed, with Ni contents of less than 0.01%. In the present study, we investigated the biocompatibility of these new alloys in vivo by subcutaneously implanting pure Ni, conventional Co-Cr-Mo, ultralow-Ni Co-Cr-Mo, and ultralow-Ni Co-Cr-Mo with Zr wires into the dorsal sides of mice. After 3 and 7 days, tissues around the wire were excised, and inflammation; the expression of IL-1β, IL-6, and TNF-α; and Ni, Co, Cr, and Mo ion release were analyzed using histological analyses, qRT-PCR, and inductively coupled plasma mass spectrometry (ICP-MS), respectively. Significantly larger amounts of Ni eluted from pure Ni wires than from the other wires, and the degree of inflammation depended on the amount of eluted Ni. Although no significant differences in inflammatory reactions were identified among new alloys and conventional Co-Cr-Mo alloys in histological and qRT-PCR analyses, ICP-MS analysis revealed that Ni ion elution from ultralow-Ni Co-Cr-Mo alloys with and without Zr was significantly lower than from conventional Co-Cr-Mo alloys. Our study, suggests that the present ultralow-Ni Co-Cr-Mo alloys with and without Zr have greater safety and utility than conventional Co-Cr-Mo alloys. © 2016 Orthopaedic Research Society. Published by Wiley Periodicals, Inc. J Orthop Res 34:1505-1513, 2016. © 2016 Orthopaedic Research Society. Published by Wiley Periodicals, Inc.

Modification and intercalation of layered zirconium phosphates: a solid-state NMR monitoring.

PubMed

Bakhmutov, Vladimir I; Kan, Yuwei; Sheikh, Javeed Ahmad; González-Villegas, Julissa; Colón, Jorge L; Clearfield, Abraham

2017-07-01

Several layered zirconium phosphates treated with Zr(IV) ions, modified by monomethoxy-polyethyleneglycol-monophosphate and intercalated with doxorubicin hydrochloride have been studied by solid-state MAS NMR techniques. The organic components of the phosphates have been characterized by the 13 C{ 1 H} CP MAS NMR spectra compared with those of initial compounds. The multinuclear NMR monitoring has provided to establish structure and covalent attachment of organic/inorganic moieties to the surface and interlayer spaces of the phosphates. The MAS NMR experiments including kinetics of proton-phosphorus cross polarization have resulted in an unusual structure of zirconium phosphate 6 combining decoration of the phosphate surface by polymer units and their partial intercalation into the interlayer space. Copyright © 2016 John Wiley & Sons, Ltd. Copyright © 2016 John Wiley & Sons, Ltd.

Deformation mechanism study of a hot rolled Zr-2.5Nb alloy by transmission electron microscopy. II. In situ transmission electron microscopy study of deformation mechanism change of a Zr-2.5Nb alloy upon heavy ion irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Long, Fei; Daymond, Mark R., E-mail: mark.daymond@queensu.ca; Yao, Zhongwen

2015-03-14

The effect of heavy-ion irradiation on deformation mechanisms of a Zr-2.5Nb alloy was investigated by using the in situ transmission electron microscopy deformation technique. The gliding behavior of prismatic 〈a〉 dislocations has been dynamically observed before and after irradiation at room temperature and 300 °C. Irradiation induced loops were shown to strongly pin the gliding dislocations. Unpinning occurred while loops were incorporated into or eliminated by 〈a〉 dislocations. In the irradiated sample, loop depleted areas with a boundary parallel to the basal plane trace were found by post-mortem observation after room temperature deformation, supporting the possibility of basal channel formation inmore » bulk neutron irradiated samples. Strong activity of pyramidal slip was also observed at both temperatures, which might be another important mechanism to induce plastic instability in irradiated zirconium alloys. Finally, (011{sup ¯}1)〈01{sup ¯}12〉 twinning was identified in the irradiated sample deformed at 300 °C.« less

Deformation mechanism study of a hot rolled Zr-2.5Nb alloy by transmission electron microscopy. II. In situ transmission electron microscopy study of deformation mechanism change of a Zr-2.5Nb alloy upon heavy ion irradiation.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Long, Fei; Daymond, Mark R.; Yao, Zhongwen

2015-03-14

The effect of heavy-ion irradiation on deformation mechanisms of a Zr-2.5Nb alloy was investigated by using the in situ transmission electron microscopy deformation technique. The gliding behavior of prismatic < a > dislocations has been dynamically observed before and after irradiation at room temperature and 300 degrees C. Irradiation induced loops were shown to strongly pin the gliding dislocations. Unpinning occurred while loops were incorporated into or eliminated by < a > dislocations. In the irradiated sample, loop depleted areas with a boundary parallel to the basal plane trace were found by post-mortem observation after room temperature deformation, supporting themore » possibility of basal channel formation in bulk neutron irradiated samples. Strong activity of pyramidal slip was also observed at both temperatures, which might be another important mechanism to induce plastic instability in irradiated zirconium alloys. Finally, {01 (1) over bar1}< 0 (1) over bar 12 > twinning was identified in the irradiated sample deformed at 300 degrees C.« less

Hot Corrosion Behavior of Bare, Cr3C2-(NiCr) and Cr3C2-(NiCr) + 0.2wt.%Zr Coated SuperNi 718 at 900 °C

NASA Astrophysics Data System (ADS)

Mudgal, Deepa; Singh, Surendra; Prakash, Satya

2015-01-01

Corrosion in incinerators, power plants, and chemical industries are frequently encountered due to the presence of salts containing sodium, sulphur, and chlorine. To obviate this problem, bare and coated alloys were tested under environments simulating the conditions present inside incinerators and power plants. 0.2 wt.% zirconium powder was incorporated in the Cr3C2-(NiCr) coating powder. The original powder and Zr containing powder was sprayed on Superni 718 alloy by D-gun technique. The bare and coated alloys were tested under Na2SO4 + K2SO4 + NaCl + KCl and Na2SO4 + NaCl environment. The corrosion rate of specimens was monitored using weight change measurements. Characterization of the corrosion products has been done using FE-SEM/EDS and XRD techniques. Bare and coated alloys showed very good corrosion resistance under given molten salt environments. Addition of 0.2wt.%Zr in Cr3C2-25%(NiCr) coating further greatly reduced the oxidation rate as well as improved the adherence of oxide scale to the coating surface during the time of corrosion.

Surface modification techniques for increased corrosion tolerance of zirconium fuel cladding

NASA Astrophysics Data System (ADS)

Carr, James Patrick, IV

Corrosion is a major issue in applications involving materials in normal and severe environments, especially when it involves corrosive fluids, high temperatures, and radiation. Left unaddressed, corrosion can lead to catastrophic failures, resulting in economic and environmental liabilities. In nuclear applications, where metals and alloys, such as steel and zirconium, are extensively employed inside and outside of the nuclear reactor, corrosion accelerated by high temperatures, neutron radiation, and corrosive atmospheres, corrosion becomes even more concerning. The objectives of this research are to study and develop surface modification techniques to protect zirconium cladding by the incorporation of a specific barrier coating, and to understand the issues related to the compatibility of the coatings examined in this work. The final goal of this study is to recommend a coating and process that can be scaled-up for the consideration of manufacturing and economic limits. This dissertation study builds on previous accident tolerant fuel cladding research, but is unique in that advanced corrosion methods are tested and considerations for implementation by industry are practiced and discussed. This work will introduce unique studies involving the materials and methods for accident tolerant fuel cladding research by developing, demonstrating, and considering materials and processes for modifying the surface of zircaloy fuel cladding. This innovative research suggests that improvements in the technique to modify the surface of zirconium fuel cladding are likely. Three elements selected for the investigation of their compatibility on zircaloy fuel cladding are aluminum, silicon, and chromium. These materials are also currently being investigated at other labs as alternate alloys and coatings for accident tolerant fuel cladding. This dissertation also investigates the compatibility of these three elements as surface modifiers, by comparing their microstructural and mechanical properties. To test their application for use in corrosive atmospheres, the corrosion behaviors are also compared in steam, water, and boric-acid environments. Various methods of surface modification were attempted in this investigation, including dip coating, diffusion bonding, casting, sputtering, and evaporation. The benefits and drawbacks of each method are discussed with respect to manufacturing and economic limits. Characterization techniques utilized in this work include optical microscopy, scanning electron microscopy, energy-dispersive spectroscopy, X-ray diffraction, nanoindentation, adhesion testing, and atomic force microscopy. The composition, microstructure, hardness, modulus, and coating adhesion were studied to provide encompassing properties to determine suitable comparisons and to choose an ideal method to scale to industrial applications. The experiments, results, and detailed discussions are presented in the following chapters of this dissertation research.

Method of making metal matrix composites reinforced with ceramic particulates

DOEpatents

Cornie, James A.; Kattamis, Theodoulos; Chambers, Brent V.; Bond, Bruce E.; Varela, Raul H.

1989-01-01

Composite materials and methods for making such materials are disclosed in which dispersed ceramic particles are at chemical equilibrium with a base metal matrix, thereby permitting such materials to be remelted and subsequently cast or otherwise processed to form net weight parts and other finished (or semi-finished) articles while maintaining the microstructure and mechanical properties (e.g. wear resistance or hardness) of the original composite. The composite materials of the present invention are composed of ceramic particles in a base metal matrix. The ceramics are preferably carbides of titanium, zirconium, tungsten, molybdenum or other refractory metals. The base metal can be iron, nickel, cobalt, chromium or other high temperature metal and alloys thereof. For ferrous matrices, alloys suitable for use as the base metal include cast iron, carbon steels, stainless steels and iron-based superalloys.

Method of making metal matrix composites reinforced with ceramic particulates

DOEpatents

Cornie, J.A.; Kattamis, T.; Chambers, B.V.; Bond, B.E.; Varela, R.H.

1989-08-01

Composite materials and methods for making such materials are disclosed in which dispersed ceramic particles are at chemical equilibrium with a base metal matrix, thereby permitting such materials to be remelted and subsequently cast or otherwise processed to form net weight parts and other finished (or semi-finished) articles while maintaining the microstructure and mechanical properties (e.g. wear resistance or hardness) of the original composite. The composite materials of the present invention are composed of ceramic particles in a base metal matrix. The ceramics are preferably carbides of titanium, zirconium, tungsten, molybdenum or other refractory metals. The base metal can be iron, nickel, cobalt, chromium or other high temperature metal and alloys thereof. For ferrous matrices, alloys suitable for use as the base metal include cast iron, carbon steels, stainless steels and iron-based superalloys. 2 figs.

In Vivo Wear Performance of Cobalt-Chromium Versus Oxidized Zirconium Femoral Total Knee Replacements.

PubMed

Gascoyne, Trevor C; Teeter, Matthew G; Guenther, Leah E; Burnell, Colin D; Bohm, Eric R; Naudie, Douglas R

2016-01-01

This study examines the damage and wear on the polyethylene (PE) inserts from 52 retrieved Genesis II total knee replacements to identify differences in tribological performance between matched pairs of cobalt-chromium (CoCr) and oxidized zirconium (OxZr) femoral components. Observer damage scoring and microcomputed tomography were used to quantify PE damage and wear, respectively. No significant differences were found between CoCr and OxZr groups in terms of PE insert damage, surface penetration, or wear. No severe damage such as cracking or delamination was noted on any of the 52 PE inserts. Observer damage scoring did not correlate with penetrative or volumetric PE wear. The more costly OxZr femoral component does not demonstrate clear tribological benefit over the standard CoCr component in the short term with this total knee replacement design. Copyright © 2016 Elsevier Inc. All rights reserved.

Basic principles of creating a new generation of high- temperature brazing filler alloys

NASA Astrophysics Data System (ADS)

Kalin, B. A.; Suchkov, A. N.

2016-04-01

The development of new materials is based on the formation of a structural-phase state providing the desired properties by selecting the base and the complex of alloying elements. The development of amorphous filler alloys for a high-temperature brazing has its own features that are due to the limited life cycle and the production method of brazing filler alloys. The work presents a cycle of analytical and experimental materials science investigations including justification of the composition of a new amorphous filler alloy for brazing the products from zirconium alloys at the temperature of no more than 800 °C and at the unbrazing temperature of permanent joints of more than 1200 °C. The experimental alloys have been used for manufacture of amorphous ribbons by rapid quenching, of which the certification has been made by X-ray investigations and a differential-thermal analysis. These ribbons were used to obtain permanent joints from the spacer grid cells (made from the alloy Zr-1% Nb) of fuel assemblies of the thermal nuclear reactor VVER-440. The brazed samples in the form of a pair of cells have been exposed to corrosion tests in autoclaves in superheated water at a temperature of 350 °C, a pressure of 160 MPa and duration of up to 6,000 h. They have been also exposed to destructive tests using a tensile machine. The experimental results obtained have made it possible to propose and patent a brazing filler alloy of the following composition: Zr-5.5Fe-(2.5-3.5)Be-1Nb-(5-8)Cu-2Sn-0.4Cr-(0.5-1.0)Ge. Its melting point is 780 °C and the recommended brazing temperature is 800°C.

Observations of earth eigen vibrations possibly excited by low frequency gravity waves

NASA Technical Reports Server (NTRS)

Tuman, V. S.

1971-01-01

A cryogenic gravity meter made of two parts, a magnetic suspension unit and a detection module, was used to monitor earth eigen vibrations. The magnetic field and field gradient are generated by energizing a set of superconducting coils made of niobium-zirconium alloy wire. The detection module is a double Josephson junction magnetometer. The output is printed on a chart recorder and later digitized using a computer; a Fourier transformation is performed on the accumulated data. The measurements of eigen vibrations are summarized in tabular and graphical representations.

Controlled atmosphere for fabrication of cermet electrodes

DOEpatents

Ray, Siba P.; Woods, Robert W.

1998-01-01

A process for making an inert electrode composite wherein a metal oxide and a metal are reacted in a gaseous atmosphere at an elevated temperature of at least about 750.degree. C. The metal oxide is at least one of the nickel, iron, tin, zinc and zirconium oxides and the metal is copper, silver, a mixture of copper and silver or a copper-silver alloy. The gaseous atmosphere has an oxygen content that is controlled at about 5-3000 ppm in order to obtain a desired composition in the resulting composite.

Controlled atmosphere for fabrication of cermet electrodes

DOEpatents

Ray, S.P.; Woods, R.W.

1998-08-11

A process is disclosed for making an inert electrode composite wherein a metal oxide and a metal are reacted in a gaseous atmosphere at an elevated temperature of at least about 750 C. The metal oxide is at least one of the nickel, iron, tin, zinc and zirconium oxides and the metal is copper, silver, a mixture of copper and silver or a copper-silver alloy. The gaseous atmosphere has an oxygen content that is controlled at about 5--3000 ppm in order to obtain a desired composition in the resulting composite. 2 figs.

Solution treatment-delayed zirconium-strengthening behavior in Ti-7.5Mo-xZr alloy system

NASA Astrophysics Data System (ADS)

Chern Lin, Jiin-Huey; Fu, Yen-Han; Chen, Yen-Chun; Peng, Yu-Po; Ju, Chien-Ping

2018-01-01

The present study was devoted to investigate and compare the Zr-strengthening behavior in as-cast (AC) and solution-treated (ST) Ti-7.5Mo-xZr alloys. The experimental results indicated that AC Ti-7.5Mo and AC Ti-7.5Mo-1Zr alloys substantially had an orthorhombic {α }\\prime\\prime phase with a fine, acicular morphology. The content of equi-axed β phase continued to increase with increased Zr content at the expense of {α }\\prime\\prime phase. The threshold Zr content for the formation of β phase in the ST Ti-7.5Mo-xZr alloys was apparently higher than that in the AC Ti-7.5Mo-xZr alloys. The β granular structure was revealed in ST Ti-7.5Mo-5Zr alloy, which increased with increased Zr content. Unlike AC Ti-7.5Mo-9Zr alloy, within each grain of ST Ti-7.5Mo-9Zr alloy were still observed a significant portion of {α }\\prime\\prime morphology. AC Ti-7.5Mo alloy had the lowest YS, lowest tensile modulus and highest elongation among all AC Ti-7.5Mo-xZr alloys. When Zr content increased, both YS and modulus significantly increased while the elongation significantly decreased. Compared to AC Ti-7.5Mo alloy, AC Ti-7.5Mo-9Zr alloy had almost double YS, indicating the effectiveness of Zr-induced strengthening in the AC Ti-7.5Mo-xZr alloys. Compared to AC Ti-7.5Mo, ST Ti-7.5Mo alloys had lower YS, UTS and tensile modulus with almost the same elongation. All the XRD, metallography and tensile test results consistently indicated that the presence of Zr could accelerate the formation of β phase and effectively strengthen the AC Ti-7.5Mo-xZr alloys. A phenomenon of delayed β formation and delayed strengthening was noted in the ST Ti-7.5Mo-xZr alloys, compared to the AC Ti-7.5Mo-xZr alloys.

Evaluation of short-term effects of rare earth and other elements used in magnesium alloys on primary cells and cell lines.

PubMed

Feyerabend, Frank; Fischer, Janine; Holtz, Jakob; Witte, Frank; Willumeit, Regine; Drücker, Heiko; Vogt, Carla; Hort, Norbert

2010-05-01

Degradable magnesium alloys for biomedical application are on the verge of being used clinically. Rare earth elements (REEs) are used to improve the mechanical properties of the alloys, but in more or less undefined mixtures. For some elements of this group, data on toxicity and influence on cells are sparse. Therefore in this study the in vitro cytotoxicity of the elements yttrium (Y), neodymium (Nd), dysprosium (Dy), praseodymium (Pr), gadolinium (Gd), lanthanum (La), cerium (Ce), europium (Eu), lithium (Li) and zirconium (Zr) was evaluated by incubation with the chlorides (10-2000 microM); magnesium (Mg) and calcium (Ca) were tested at higher concentrations (200 and 50mM, respectively). The influence on viability of human osteosarcoma cell line MG63, human umbilical cord perivascular (HUCPV) cells and mouse macrophages (RAW 264.7) was determined, as well as the induction of apoptosis and the expression of inflammatory factors (TNF-alpha, IL-1alpha). Significant differences between the applied cells could be observed. RAW exhibited the highest and HUCPV the lowest sensitivity. La and Ce showed the highest cytotoxicity of the analysed elements. Of the elements with high solubility in magnesium alloys, Gd and Dy seem to be more suitable than Y. The focus of magnesium alloy development for biomedical applications should include most defined alloy compositions with well-known tissue-specific and systemic effects. Copyright (c) 2009 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

Isoelectronic substitutions and aluminium alloying in the Ta-Nb-Hf-Zr-Ti high-entropy alloy superconductor

NASA Astrophysics Data System (ADS)

von Rohr, Fabian O.; Cava, Robert J.

2018-03-01

High-entropy alloys (HEAs) are a new class of materials constructed from multiple principal elements statistically arranged on simple crystallographic lattices. Due to the large amount of disorder present, they are excellent model systems for investigating the properties of materials intermediate between crystalline and amorphous states. Here we report the effects of systematic isoelectronic replacements, using Mo-Y, Mo-Sc, and Cr-Sc mixtures, for the valence electron count 4 and 5 elements in the body-centered cubic (BCC) Ta-Nb-Zr-Hf-Ti high-entropy alloy (HEA) superconductor. We find that the superconducting transition temperature Tc strongly depends on the elemental makeup of the alloy, and not exclusively its electron count. The replacement of niobium or tantalum by an isoelectronic mixture lowers the transition temperature by more than 60%, while the isoelectronic replacement of hafnium, zirconium, or titanium has a limited impact on Tc. We further explore the alloying of aluminium into the nearly optimal electron count [TaNb] 0.67(ZrHfTi) 0.33 HEA superconductor. The electron count dependence of the superconducting Tc for (HEA)Al x is found to be more crystallinelike than for the [TaNb] 1 -x(ZrHfTi) x HEA solid solution. For an aluminum content of x =0.4 the high-entropy stabilization of the simple BCC lattice breaks down. This material crystallizes in the tetragonal β -uranium structure type and superconductivity is not observed above 1.8 K.

Thin-Film Phase Plates for Transmission Electron Microscopy Fabricated from Metallic Glasses.

PubMed

Dries, Manuel; Hettler, Simon; Schulze, Tina; Send, Winfried; Müller, Erich; Schneider, Reinhard; Gerthsen, Dagmar; Luo, Yuansu; Samwer, Konrad

2016-10-01

Thin-film phase plates (PPs) have become an interesting tool to enhance the contrast of weak-phase objects in transmission electron microscopy (TEM). The thin film usually consists of amorphous carbon, which suffers from quick degeneration under the intense electron-beam illumination. Recent investigations have focused on the search for alternative materials with an improved material stability. This work presents thin-film PPs fabricated from metallic glass alloys, which are characterized by a high electrical conductivity and an amorphous structure. Thin films of the zirconium-based alloy Zr65.0Al7.5Cu27.5 (ZAC) were fabricated and their phase-shifting properties were evaluated. The ZAC film was investigated by different TEM techniques, which reveal beneficial properties compared with amorphous carbon PPs. Particularly favorable is the small probability for inelastic plasmon scattering, which results from the combined effect of a moderate inelastic mean free path and a reduced film thickness due to a high mean inner potential. Small probability plasmon scattering improves contrast transfer at high spatial frequencies, which makes the ZAC alloy a promising material for PP fabrication.

Electrothermal atomic absorption spectrometric determination of copper in nickel-base alloys with various chemical modifiers*1

NASA Astrophysics Data System (ADS)

Tsai, Suh-Jen Jane; Shiue, Chia-Chann; Chang, Shiow-Ing

1997-07-01

The analytical characteristics of copper in nickel-base alloys have been investigated with electrothermal atomic absorption spectrometry. Deuterium background correction was employed. The effects of various chemical modifiers on the analysis of copper were investigated. Organic modifiers which included 2-(5-bromo-2-pyridylazo)-5-(diethylamino-phenol) (Br-PADAP), ammonium citrate, 1-(2-pyridylazo)-naphthol, 4-(2-pyridylazo)resorcinol, ethylenediaminetetraacetic acid and Triton X-100 were studied. Inorganic modifiers palladium nitrate, magnesium nitrate, aluminum chloride, ammonium dihydrogen phosphate, hydrogen peroxide and potassium nitrate were also applied in this work. In addition, zirconium hydroxide and ammonium hydroxide precipitation methods have also been studied. Interference effects were effectively reduced with Br-PADAP modifier. Aqueous standards were used to construct the calibration curves. The detection limit was 1.9 pg. Standard reference materials of nickel-base alloys were used to evaluate the accuracy of the proposed method. The copper contents determined with the proposed method agreed closely with the certified values of the reference materials. The recoveries were within the range 90-100% with relative standard deviation of less than 10%. Good precision was obtained.

Study of twin-roll cast Aluminium alloys subjected to severe plastic deformation by equal channel angular pressing

NASA Astrophysics Data System (ADS)

Poková, M.; Cieslar, M.

2014-08-01

Aluminium alloys prepared by twin-roll casting method become widely used in industry applications. Their high solid solution supersaturation and finer grains ensure better mechanical properties when compared with the direct-chill cast ones. One of the possibilities how to enhance their thermal stability is the addition of zirconium. After heat treatment Al3Zr precipitates form and these pin moving grain boundaries when the material is exposed to higher temperatures. In the present work twin-roll cast aluminium alloys based on AA3003 with and without Zr addition were annealed for 8 hours at 450 °C to enable precipitation of Al3Zr phase. Afterwards they were subjected to severe plastic deformation by equal channel angular pressing, which led to the reduction of average grain size under 1 μm. During subsequent isochronal annealing recovery and recrystallization took place. These processes were monitored by microhardness measurements, light optical microscopy and in-situ transmission electron microscopy. The addition of Zr stabilizes the grain size and increases the recrystallization temperature by 100 °C.

Comparison of magnesium alloys and poly-l-lactide screws as degradable implants in a canine fracture model.

PubMed

Marukawa, Eriko; Tamai, Masato; Takahashi, Yukinobu; Hatakeyama, Ichiro; Sato, Masaru; Higuchi, Yusuke; Kakidachi, Hiroshi; Taniguchi, Hirofumi; Sakamoto, Takamitsu; Honda, Jun; Omura, Ken; Harada, Hiroyuki

2016-10-01

The aims of this study were to evaluate in vivo the biological responses to implants composed of biodegradable anodized WE43 (containing magnesium yttrium, rare earth elements and zirconium; Elektron SynerMag®) magnesium alloy, monolithic WE43 magnesium alloy and poly-l-lactic acid (PLLA), which are commonly used materials in clinic settings, and to evaluate the effectiveness of the materials as bone screws. The effectiveness of the magnesium alloy implants in osteosynthesis was evaluated using a bone fracture model involving the tibia of beagle dogs. For the monolithic WE43 implants, radiological, and histological evaluation revealed that bone trabeculae around the implanted monolithic WE43 decreased because of an inflammatory response. However, there was no damage due to hydrogen gas or inflammatory response in the bone tissue around the anodized WE43 implants. After 4 weeks, all the PLLA implants (n = 3) had broken but the WE43 implants had not (n = 6). These results suggest that the WE43 implants had sufficient strength to fix bone fractures at load-bearing sites in orthopedic and oral maxillofacial surgery. Therefore, these biodegradable magnesium alloys are good candidates for replacing biodegradable polymers. © 2015 Wiley Periodicals, Inc. J Biomed Mater Res Part B: Appl Biomater, 104B: 1282-1289, 2016. © 2015 Wiley Periodicals, Inc.

Mechanical behavior of aluminum-bearing ferritic alloys for accident-tolerant fuel cladding applications

NASA Astrophysics Data System (ADS)

Guria, Ankan

Nuclear power currently provides about 13% of electrical power worldwide. Nuclear reactors generating this power traditionally use Zirconium (Zr) based alloys as the fuel cladding material. Exothermic reaction of Zr with steam under accident conditions may lead to production of hydrogen with the possibility of catastrophic consequences. Following the Fukushima-Daiichi incident, the exploration of accident-tolerant fuel cladding materials accelerated. Aluminum-rich (around 5 wt. %) ferritic steels such as Fecralloy, APMT(TM) and APM(TM) are considered as potential materials for accident-tolerant fuel cladding applications. These materials create an aluminum-based oxide scale protecting the alloy at elevated temperatures. Tensile deformation behavior of the above alloys was studied at different temperatures (25-500 °C) at a strain rate of 10-3 s-1 and correlated with microstructural characteristics. Higher strength and decent ductility of APMT(TM) led to further investigation of the alloy at various combination of strain rates and temperatures followed by fractography and detailed microscopic analyses. Serrations appeared in the stress-strain curves of APMT(TM) and Fecralloy steel tested in a limited temperature range (250-400 °C). The appearance of serrations is explained on the basis of dynamic strain aging (DSA) effect due to solute-dislocation interactions. The research in this study is being performed using the funds received from the US DOE Office of Nuclear Energy's Nuclear Energy University Programs (NEUP).

Effects of surface treatment of aluminium alloy 1050 on the adhesion and anticorrosion properties of the epoxy coating

NASA Astrophysics Data System (ADS)

Sharifi Golru, S.; Attar, M. M.; Ramezanzadeh, B.

2015-08-01

The objective of this work is to investigate the effects of zirconium-based (Zr) conversion coating on the adhesion properties and corrosion resistance of an epoxy/polyamide coating applied on the aluminium alloy 1050 (AA1050). Field emission scanning electron microscope (FE-SEM), energy dispersive X-ray spectrum (EDS), atomic force microscope (AFM) and contact angle measuring device were employed in order to characterize the surface characteristics of the Zr treated AA1050 samples. The epoxy/polyamide coating was applied on the untreated and Zr treated samples. The epoxy coating adhesion to the aluminium substrate was evaluated by pull-off test before and after 30 days immersion in 3.5% w/w NaCl solution. In addition, the electrochemical impedance spectroscopy (EIS) and salt spray tests were employed to characterize the corrosion protection properties of the epoxy coating applied on the AA1050 samples. Results revealed that the surface treatment of AA1050 by zirconium conversion coating resulted in the increase of surface free energy and surface roughness. The dry and recovery (adhesion strength after 30 days immersion in the 3.5 wt% NaCl solution) adhesion strengths of the coatings applied on the Zr treated aluminium samples were greater than untreated sample. In addition, the adhesion loss of the coating applied on the Zr treated aluminium substrate was lower than other samples. Also, the results obtained from EIS and salt spray test clearly revealed that the Zr conversion coating could enhance the corrosion protective performance of the epoxy coating significantly.

Investigation of hydrogen interaction with defects in zirconia

NASA Astrophysics Data System (ADS)

Melikhova, O.; Kuriplach, J.; Čížek, J.; Procházka, I.; Brauer, G.; Anwand, W.

2010-04-01

Defect studies of a ZrO2 + 9 mol. % Y2O3 single crystal were performed in this work using a high resolution positron lifetime spectroscopy combined with slow positron implantation spectroscopy. In order to elucidate the nature of positron trapping sites observed experimentally, the structural relaxations of several types of vacancy-like defects in zirconia were performed and positron characteristics for them were calculated. Relaxed atomic configurations of studied defects were obtained by means of ab initio pseudopotential method within the supercell approach. Theoretical calculations indicated that neither oxygen vacancies nor their neutral complexes with substitute yttrium atoms are capable of positron trapping. On the other hand, zirconium vacancies are deep positron traps and are most probably responsible for the saturated positron trapping observed in yttria stabilized zirconia single crystals. However, the calculated positron lifetime for zirconium vacancy is apparently longer than the experimental value corresponding to a single-component spectrum measured for the cubic ZrO2 + 9 mol. % Y2O3 single crystal. It was demonstrated that this effect can be explained by hydrogen trapped in zirconium vacancies. On the basis of structure relaxations, we found that zirconium vacancy - hydrogen complexes represent deep positron traps with the calculated lifetime close to the experimental one. In zirconium vacancy - hydrogen complexes the hydrogen atom forms an O-H bond with one of the nearest neighbour oxygen atoms. The calculated bond length is close to 1 Å.

U-Mo Monolithic Fuel for Nuclear Research and Test Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Prabhakaran, Ramprashad

The metallic fuel selected to replace the current HEU fuels in the research and test reactors is the LEU-10 weight % Mo alloy in the form of a thin sheet or foil encapsulated in AA6061 aluminum alloy with a zirconium interlayer. In order to effectively lead this pursuit, new developments in processing and fabrication of the fuel elements have been initiated, along with a better understanding of material behavior before and after irradiation as a result of these new developments. This editorial note gives an introduction about research and test reactors, need for HEU to LEU conversion, fuel requirements, highmore » uranium density monolithic fuel development and an overview of the four articles published in the December 2017 issue of JOM under a special topic titled “U-Mo Monolithic Fuel for Nuclear Research and Test Reactors”.« less

Understanding mechanisms of solid-state phase transformations by probing nuclear materials

NASA Astrophysics Data System (ADS)

Banerjee, Srikumar; Donthula, Harish

2018-04-01

In this review a few examples will be cited to illustrate that a study on a specific nuclear material sometimes lead to a better understanding of scientific phenomena of broader interests. Zirconium alloys offer some unique opportunities in addressing fundamental issues such as (i) distinctive features between displacive and diffusional transformations, (ii) characteristics of shuffle and shear dominated displacive transformations and (iii) nature of mixed-mode transformations. Whether a transformation is of first or higher order?" is often raised while classifying it. There are rare examples, such as Ni-Mo alloys, in which during early stages of ordering the system experiences tendencies for both first order and second order transitions. Studies on the order-disorder transitions under a radiation environment have established the pathway for the evolution of ordering. These studies have also identified the temperature range over which the chemically ordered state remains stable in steady state under radiation.

Applying NASA's explosive seam welding

NASA Technical Reports Server (NTRS)

Bement, Laurence J.

1991-01-01

The status of an explosive seam welding process, which was developed and evaluated for a wide range of metal joining opportunities, is summarized. The process employs very small quantities of explosive in a ribbon configuration to accelerate a long-length, narrow area of sheet stock into a high-velocity, angular impact against a second sheet. At impact, the oxide films of both surface are broken up and ejected by the closing angle to allow atoms to bond through the sharing of valence electrons. This cold-working process produces joints having parent metal properties, allowing a variety of joints to be fabricated that achieve full strength of the metals employed. Successful joining was accomplished in all aluminum alloys, a wide variety of iron and steel alloys, copper, brass, titanium, tantalum, zirconium, niobium, telerium, and columbium. Safety issues were addressed and are as manageable as many currently accepted joining processes.

Corrosion performance of Cr3C2-NiCr+0.2%Zr coated super alloys under actual medical waste incinerator environment

NASA Astrophysics Data System (ADS)

Ahuja, Lalit; Mudgal, Deepa; Singh, Surendra; Prakash, Satya

2018-03-01

Incineration techniques are widely used to dispose of various types of waste which lead to formation of very corrosive environment. Such corrosive environment leads to the degradation of the alloys used in these areas. To obviate this problem, zirconium modified Cr3C2-(NiCr) coating powder has been deposited on three superalloys namely Superni 718, Superni 600 and Superco 605 using Detonation gun technique. Corrosion test was conducted in actual medical waste incinerator environment. The samples were hung inside the secondary chamber operated at 1050°C for 1000h under cyclic condition. Corrosion kinetics was monitored using the weight gain measurements and thickness loss. Corrosion products were characterized using scanning electron microscopy, energy dispersive spectroscopy and X-ray diffraction technique. It was observed that coating is found to be successful in impeding the corrosion problem in superalloys.

Effect of He implantation on the microstructure of zircaloy-4 studied using in situ TEM

NASA Astrophysics Data System (ADS)

Tunes, M. A.; Harrison, R. W.; Greaves, G.; Hinks, J. A.; Donnelly, S. E.

2017-09-01

Zirconium alloys are of great importance to the nuclear industry as they have been widely used as cladding materials in light-water reactors since the 1960s. This work examines the behaviour of these alloys under He ion implantation for the purposes of developing understanding of the fundamental processes behind their response to irradiation. Characterization of zircaloy-4 samples using TEM with in situ 6 keV He irradiation up to a fluence of 2.7 ×1017ions ·cm-2 in the temperature range of 298 to 1148 K has been performed. Ordered arrays of He bubbles were observed at 473 and 1148 K at a fluence of 1.7 ×1017ions ·cm-2 in αZr, the hexagonal compact (HCP) and in βZr, the body centred cubic (BCC) phases, respectively. In addition, the dissolution behaviour of cubic Zr hydrides under He irradiation has been investigated.

Effect of the chemical composition and the structural and phases states of materials on hydrogen retention

NASA Astrophysics Data System (ADS)

Chernov, I. I.; Stal'tsov, M. S.; Kalin, B. A.; Bogachev, I. A.; Guseva, L. Yu.; Korshunov, S. N.

2017-07-01

The results of investigation of the effect of chemical composition and structural and phase states of reactor steels and vanadium alloys on their capture and retention of hydrogen introduced into the materials in various ways are presented. It is shown that, in the case of identical conditions of hydrogen introduction, the amount of hydrogen captured by austenitic steels is substantially higher than that captured by ferritic/ martensitic steels. At the same time, the EP450 ODS ferritic/martensitic steel dispersion-strengthened with nanosized yttrium oxide particles retains a substantially higher amount of hydrogen as compared to that retained in the EP450 matrix steel. The alloying of vanadium with tungsten, zirconium, and titanium leads to an increase in the amount of retained hydrogen. The effect of titanium content on hydrogen retention is found to be nonmonotonic; the phenomenon is explained from a physical view point.

Establishment of the roadmap for chlorination process development for zirconium recovery and recycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collins, E.D.; Del Cul, G.D.; Spencer, B.B.

Process development studies are being conducted to recover, purify, and reuse the zirconium (about 98.5% by mass) in used nuclear fuel (UNF) zirconium alloy cladding. Feasibility studies began in FY 2010 using empty cladding hulls that were left after fuel dissolution or after oxidation to a finely divided oxide powder (voloxidation). In FY 2012, two industrial teams (AREVA and Shaw-Westinghouse) were contracted by the Department of Energy Office of Nuclear Energy (NE) to provide technical assistance to the project. In FY 2013, the NE Fuel Cycle Research and Development Program requested development of a technology development roadmap to guide futuremore » work. The first step in the roadmap development was to assess the starting point, that is, the current state of the technology and the end goal. Based on previous test results, future work is to be focused on first using chlorine as the chlorinating agent and secondly on the use of a process design that utilizes a chlorination reactor and dual ZrCl{sub 4} product salt condensers. The likely need for a secondary purification step was recognized. Completion of feasibility testing required an experiment on the chemical decladding flowsheet option. This was done during April 2013. The roadmap for process development will continue through process chemistry optimization studies, the chlorinated reactor design configuration, product salt condensers, and the off-gas trapping of tritium or other volatile fission products from the off-gas stream. (authors)« less

Microstructure evolution and tensile properties of Zr-2.5 wt.% Nb pressure tubes processed from billets with different microstructures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kapoor, K.; Saratchandran, N.; Muralidharan, K.

1999-02-01

Pressurized heavy water reactors (PHWR) use zirconium-base alloys for their low neutron-absorption cross section, good mechanical strength, low irradiation creep, and high corrosion resistance in reactor atmospheres. Starting with identical ingots, billets having different microstructures were obtained by three different processing methods for fabrication of Zr-2.5 wt%Nb pressure tubes., The billets were further processed by hot extrusion and cold Pilger tube reducing to the finished product. Microstructural characterization was done at each stage of processing. The effects of the initial billet microstructure on the intermediate and final microstructure and mechanical property results were determined. It was found that the structuremore » at each stage and the final mechanical properties depend strongly on the initial billet microstructure. The structure at the final stage consists of elongated alpha zirconium grains with a network of metastable beta zirconium phase. Some of this metastable phase transforms into stable beta niobium during thermomechanical processing. Billets with quenched structure resulted in less beta niobium at the final stage. The air cooled billets resulted in a large amount of beta niobium. The tensile properties, especially the percentage elongation, were found to vary for the different methods. Higher percentage elongation was observed for billets having quenched structure. Extrusion and forging did not produce any characteristic differences in the properties. The results were used to select a process flow sheet which yields the desired mechanical properties with suitable microstructure in the final product.« less

Zirconium Ions Up-Regulate the BMP/SMAD Signaling Pathway and Promote the Proliferation and Differentiation of Human Osteoblasts

PubMed Central

Chen, Yongjuan; Roohani-Esfahani, Seyed-Iman; Lu, ZuFu; Zreiqat, Hala; Dunstan, Colin R.

2015-01-01

Zirconium (Zr) is an element commonly used in dental and orthopedic implants either as zirconia (ZrO2) or in metal alloys. It can also be incorporated into calcium silicate-based ceramics. However, the effects of in vitro culture of human osteoblasts (HOBs) with soluble ionic forms of Zr have not been determined. In this study, primary culture of human osteoblasts was conducted in the presence of medium containing either ZrCl4 or Zirconium (IV) oxynitrate (ZrO(NO3)2) at concentrations of 0, 5, 50 and 500 µM, and osteoblast proliferation, differentiation and calcium deposition were assessed. Incubation of human osteoblast cultures with Zr ions increased the proliferation of human osteoblasts and also gene expression of genetic markers of osteoblast differentiation. In 21 and 28 day cultures, Zr ions at concentrations of 50 and 500 µM increased the deposition of calcium phosphate. In addition, the gene expression of BMP2 and BMP receptors was increased in response to culture with Zr ions and this was associated with increased phosphorylation of SMAD1/5. Moreover, Noggin suppressed osteogenic gene expression in HOBs co-treated with Zr ions. In conclusion, Zr ions appear able to induce both the proliferation and the differentiation of primary human osteoblasts. This is associated with up-regulation of BMP2 expression and activation of BMP signaling suggesting this action is, at least in part, mediated by BMP signaling. PMID:25602473

Explosive Welding of Aluminum, Titanium and Zirconium to Copper Sheet Metal

NASA Technical Reports Server (NTRS)

Hegazy, A. A.; Mote, J. D.

1985-01-01

The main material properties affecting the explosive weldability of a certain metal combination are the yield strength, the ductility, the density and the sonic velocity of the two metals. Successful welding of the metal combination depends mainly on the correct choice of the explosive welding parameters; i.e., the stand off distance, the weight of the explosive charge relative to the weight of the flyer plate and the detonation velocity of the explosive. Based on the measured and the handbook values of the properties of interest, the explosive welding parameters were calculated and the arrangements for the explosive welding of the Al alloy 6061-T6, titanium and zirconium to OFHC copper were determined. The relatively small sheet metal thickness (1/8") and the fact that the thickness of the explosive layer must exceed a certain minimum value were considered during the determination of the explosive welding conditions. The results of the metallographic investigations and the measurements of the shear strength at the interface demonstrate the usefulness of these calculations to minimize the number of experimental trials.

The Chemical Composition and Structure of Supported Sulfated Zirconia with Regulated Size Nanoparticles

NASA Astrophysics Data System (ADS)

Kanazhevskiy, V. V.; Shmachkova, V. P.; Kotsarenko, N. S.; Kochubey, D. I.; Vedrine, J. C.

2007-02-01

A set of model skeletal isomerization catalysts — sulfated zirconia nanoparticles of controlled thickness anchored on different supports — was prepared using colloidal solutions of Zr salt on titania as support. The nanoparticles of zirconia (1-5 nm) are epitaxially connected to the support surface, with S/Zr ratio equals to 1.3-1.5. It was shown by EXAFS that nanoparticles of non-stoichiometric zirconium sulfate Zr(SO4)1+x, where x<0.5, are formed on the support surface. Its structure looks like half-period shifted counterdirected chains built-up by zirconium atoms linked by triangle pyramids of sulfate groups. Considering catalytic data of skeletal n-butane isomerisation at 150°C, one can suggest that these species behave as the active component of sulfated zirconia. They are formed in subsurface layers as zirconium hydroxide undergoes sulfation followed by thermal treatment.

The effect of 0.1 atomic percent zirconium on the cyclic oxidation behavior of beta-NiAl for 300 hours at 1200 C

NASA Technical Reports Server (NTRS)

Barrett, C. A.

1988-01-01

The long time effect of 0.1 at percent Zr (0.2 wt percent Zr) on the cyclic oxidation behavior of hipped beta-NiAl was studied. Oxidation testing was performed in static air at 1200 C for up to 3000 one-hour exposure cycles. Specific weight change versus time data was modeled with the COSP computer program to analyze cyclic oxidation behavior. The Zr-free stoichiometric alloy oxidized and spalled randomly to bare metal between cycles at a rate high enough to deplete Al to a low enough level that oxidation breakaway took place as nonprotective NiO replaced the alpha-Al2O3/NiAl2O4 scale as the controlling oxide. The Zr minimized this severe type of spalling maintaining the protective alpha-Al2O3 scale even out to 3000 hours for the stoichiometric alloy with no significant Al depletion. A third beta-NiAl alloy containing 0.1 at percent Zr but with 10 percent less Al than the stoichiometric alloy was also tested and showed some depletion of Al, but the protective Al2O3/NiAl2O4 was still maintained to close to 2700 hours.

Evolution of interphase and intergranular strain in zirconium-niobium alloys during deformation at room temperature

NASA Astrophysics Data System (ADS)

Cai, Song

Zr-2.5Nb is currently used for pressure tubes in the CANDU (CANada Deuterium Uranium) reactor. A complete understanding of the deformation mechanism of Zr-2.5Nb is important if we are to accurately predict the in-reactor performance of pressure tubes and guarantee normal operation of the reactors. This thesis is a first step in gaining such an understanding; the deformation mechanism of ZrNb alloys at room temperature has been evaluated through studying the effect of texture and microstructure on deformation. In-situ neutron diffraction was used to monitor the evolution of the lattice strain of individual grain families along both the loading and Poisson's directions and to track the development of interphase and intergranular strains during deformation. The following experiments were carried out with data interpreted using elasto-plastic modeling techniques: (1) Compression tests of a 100%betaZr material at room temperature. (2) Tension and compression tests of hot rolled Zr-2.5Nb plate material. (3) Compression of annealed Zr-2.5Nb. (4) Cyclic loading of the hot rolled Zr-2.5Nb. (5) Compression tests of ZrNb alloys with different Nb and oxygen contents. The experimental results were interpreted using a combination of finite element (FE) and elasto-plastic self-consistent (EPSC) models. The phase properties and phase interactions well represented by the FE model, the EPSC model successfully captured the evolution of intergranular constraint during deformation and provided reasonable estimates of the critical resolved shear stress and hardening parameters of different slip systems under different conditions. The consistency of the material parameters obtained by the EPSC model allows the deformation mechanism at room temperature and the effect of textures and microstructures of ZrNb alloys to be understood. This work provides useful information towards manufacturing of Zr-2.5Nb components and helps in producing ideal microstructures and material properties for pressure tubes. Also it is helpful in guiding the development of new materials for the next generation of nuclear reactors. Furthermore, the large data set obtained from this study can be used in evaluation and improving current and future polycrystalline deformation models.

Metal Hypersensitivity and Total Knee Arthroplasty

PubMed Central

Lachiewicz, Paul F.; Watters, Tyler Steven; Jacobs, Joshua J.

2015-01-01

Metal hypersensitivity in patients with a total knee arthroplasty (TKA) is a controversial topic. The diagnosis is difficult, given the lack of robust clinical validation of the utility of cutaneous and in vitro testing. Metal hypersensitivity after TKA is quite rare and should be considered after eliminating other causes of pain and swelling, such as low-grade infection, instability, component loosening or malrotation, referred pain, and chronic regional pain syndrome. Anecdotal observations suggest that two clinical presentations of metal hypersensitivity may occur after TKA: dermatitis or a persistent painful synovitis of the knee. Patients may or may not have a history of intolerance to metal jewelry. Laboratory studies, including erythrocyte sedimentation rate, C-reactive protein level, and knee joint aspiration, are usually negative. Cutaneous and in vitro testing have been reported to be positive, but the sensitivity and specificity of such testing has not been defined. Anecdotal reports suggest that, if metal hypersensitivity is suspected and nonsurgical measures have failed, then revision to components fabricated of titanium alloy or zirconium coating can be successful in relieving symptoms. Revision should be considered as a last resort, however, and patients should be informed that no evidence-based medicine is available to guide the management of these conditions, particularly for decisions regarding revision. Given the limitations of current testing methods, the widespread screening of patients for metal allergies before TKA is not warranted. PMID:26752739

Scanning Electron Microscopy Analysis of the Adaptation of Single-Unit Screw-Retained Computer-Aided Design/Computer-Aided Manufacture Abutments After Mechanical Cycling.

PubMed

Markarian, Roberto Adrian; Galles, Deborah Pedroso; Gomes França, Fabiana Mantovani

To measure the microgap between dental implants and custom abutments fabricated using different computer-aided design/computer-aided manufacture (CAD/CAM) methods before and after mechanical cycling. CAD software (Dental System, 3Shape) was used to design a custom abutment for a single-unit, screw-retained crown compatible with a 4.1-mm external hexagon dental implant. The resulting stereolithography file was sent for manufacturing using four CAD/CAM methods (n = 40): milling and sintering of zirconium dioxide (ZO group), cobalt-chromium (Co-Cr) sintered via selective laser melting (SLM group), fully sintered machined Co-Cr alloy (MM group), and machined and sintered agglutinated Co-Cr alloy powder (AM group). Prefabricated titanium abutments (TI group) were used as controls. Each abutment was placed on a dental implant measuring 4.1× 11 mm (SA411, SIN) inserted into an aluminum block. Measurements were taken using scanning electron microscopy (SEM) (×4,000) on four regions of the implant-abutment interface (IAI) and at a relative distance of 90 degrees from each other. The specimens were mechanically aged (1 million cycles, 2 Hz, 100 N, 37°C) and the IAI width was measured again using the same approach. Data were analyzed using two-way analysis of variance, followed by the Tukey test. After mechanical cycling, the best adaptation results were obtained from the TI (2.29 ± 1.13 μm), AM (3.58 ± 1.80 μm), and MM (1.89 ± 0.98 μm) groups. A significantly worse adaptation outcome was observed for the SLM (18.40 ± 20.78 μm) and ZO (10.42 ± 0.80 μm) groups. Mechanical cycling had a marked effect only on the AM specimens, which significantly increased the microgap at the IAI. Custom abutments fabricated using fully sintered machined Co-Cr alloy and machined and sintered agglutinated Co-Cr alloy powder demonstrated the best adaptation results at the IAI, similar to those obtained with commercial prefabricated titanium abutments after mechanical cycling. The adaptation of custom abutments made by means of SLM or milling and sintering of zirconium dioxide were worse both before and after mechanical cycling.

Cherenkov and Scintillation Properties of Cubic Zirconium

NASA Technical Reports Server (NTRS)

Christl, M.J.; Adams, J.H.; Parnell, T.A.; Kuznetsov, E.N.

2008-01-01

Cubic zirconium (CZ) is a high index of refraction (n =2.17) material that we have investigated for Cherenkov counter applications. Laboratory and proton accelerator tests of an 18cc sample of CZ show that the expected fast Cherenkov response is accompanied by a longer scintillation component that can be separated by pulse shaping. This presents the possibility of novel particle spectrometers which exploits both properties of CZ. Other high index materials being examined for Cherenkov applications will be discussed. Results from laboratory tests and an accelerator exposure will be presented and a potential application in solar energetic particle instruments will be discussed

Structure and Properties of Titanium Tantalum Alloys for Biocompatibility

NASA Astrophysics Data System (ADS)

Huber, Daniel E.

In this thesis, the phase stability and elastic modulus of Ti-Ta simple binary alloys as well as alloys with small additions of ternary elements have been studied. The binary alloy from a nominal 8 to 28 wt.% Ta was first explored using a combinatorial approach. This approach included Laser Engineered Net Shape (LENSTM) processing of materials and subsequent characterization by instrumented indentation and site specific Transmission Electron Microscopy (TEM). The composition range of 15 to 75 wt.% Ta was further explored by more traditional methods that included vacuum arc melting high purity elements, X-Ray Diffraction (XRD) and modulus measurements made by ultrasonic methods. Beyond the simple binary, alloys with low levels of ternary elements, oxygen, aluminum, zirconium and small additions of rare earth oxides were investigated. The crystal structure with space group Cmcm was chosen for it applicability with P63/mmc and Im-3¯m sub group / super group symmetry. This provides a consistent crystal structure framework for the purpose of studying the alpha to beta transformation pathway and associated alpha' and alpha'' martensitic phases. In this case, the pathway is defined by both the lattice parameters and the value of the parameter "y", where the parameter "y" describes the atomic positions of the [002]alpha plane. It was found that the lattice parameter changes in the Ti-Ta binary alloys are similar to structures reported for compositions in the Ti-Nb system of similar atomic percentages. Although samples produced by the LENSTM; process and characterized by instrumented indentation demonstrated the correct trends in modulus behavior, absolute agreement was not seen with modulus values published in literature. Alloys of the binary Ti-Ta system produced from high purity materials do indeed show close agreement with literature where there exist two minima of modulus near the compositions of Ti-28Ta wt.% and Ti-68Ta wt.%. These two minima occur at the discreet boundary between alpha' / alpha'' and alpha'' / beta respectively. The role of oxygen as an alloying addition was studied as it relates to the stability of alpha' and alpha'' martensite, here it was found that oxygen will stabilize alpha' yet cause an increase in the Young's modulus. Rare earth additions to getter interstitial oxygen in the high purity materials show no further reduction in modulus. Conversely, additions of another alpha stabilizer, Al, proved to lower the alpha' stability, with one composition exhibiting a modulus as low as 53 GPa. Zirconium being a neutral element regarding alpha and beta stability slightly changed the structure and lattice parameter, while making a little or no difference in the observed modulus. Observations by TEM of quenched specimens indicate the rise in modulus observed between the two minima is not caused the appearance of o. Rather weak o reflections were observed in Ti-65Ta wt.% in the as arc-melted condition and on annealing for 450°C for 24 hours. Precipitates of o were not clearly identified by dark-field TEM imaging. High Resolution Scanning Transmission Electron Microscopy (HRSTEM) of the aged specimen indicated that o might exist as 3-5nm particles.

Oxygen stabilized zirconium-vanadium-iron alloy

DOEpatents

Mendelsohn, Marshall H.; Gruen, Dieter M.

1982-01-01

An oxygen stabilized intermetallic compound having the formula (Zr.sub.1-x Ti.sub.x).sub.2-u (V.sub.1-y Fe.sub.y)O.sub.z where x=0.0 to 0.9, y=0.01 to 0.9, z=0.25 to 0.5 and u=0 to 1. The compound is capable of reversibly sorbing hydrogen at temperatures from -196.degree. C. to 200.degree. C. at pressures down to 10.sup.-6 torr. The compound is suitable for use as a hydrogen getter in low pressure, high temperature applications such as magnetic confinement fusion devices.

High-resolution neutron-diffraction measurements to 8 kbar

NASA Astrophysics Data System (ADS)

Bull, C. L.; Fortes, A. D.; Ridley, C. J.; Wood, I. G.; Dobson, D. P.; Funnell, N. P.; Gibbs, A. S.; Goodway, C. M.; Sadykov, R.; Knight, K. S.

2017-10-01

We describe the capability to measure high-resolution neutron powder diffraction data to a pressure of at least 8 kbar. We have used the HRPD instrument at the ISIS neutron source and a piston-cylinder design of pressure cell machined from a null-scattering titanium zirconium alloy. Data were collected under hydrostatic conditions from an elpasolite perovskite La?NiMnO?; by virtue of a thinner cell wall on the incident-beam side of the cell, it was possible to obtain data in the instrument's highest resolution back-scattering detector banks up to a maximum pressure of 8.5 kbar.

Oxygen-stabilized zirconium-vanadium-iron alloy

DOEpatents

Mendelsohn, M.H.; Gruen, D.M.

1981-06-16

An oxygen stabilized intermetallic compound is described which has the formula (Zr/sub 1-x/Ti/sub x/)/sub 2-u/(V/sub 1-y/Fe/sub y/)O/sub z/ where x = 0.0 to 0.9, y = 0.01 to 0.9, z = 0.25 to 0.5 and u = 0 to 1. The compound is capable of reversibly sorbing hydrogen at temperatures from -196/sup 0/C to 200/sup 0/C at pressures down to 10/sup -6/ torr. The compound is suitable for use as a hydrogen getter in low pressure, high temperature applications such as magnetic confinement fusion devices.

Development of LWR Fuels with Enhanced Accident Tolerance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lahoda, Edward J.; Boylan, Frank A.

2015-10-30

Significant progress was made on the technical, licensing, and business aspects of the Westinghouse Electric Company’s Enhanced Accident Tolerant Fuel (ATF) by the Westinghouse ATF team. The fuel pellet options included waterproofed U 15N and U 3Si 2 and the cladding options SiC composites and zirconium alloys with surface treatments. Technology was developed that resulted in U 3Si 2 pellets with densities of >94% being achieved at the Idaho National Laboratory (INL). The use of U 3Si 2 will represent a 15% increase in U235 loadings over those in UO₂ fuel pellets. This technology was then applied to manufacture pelletsmore » for 6 test rodlets which were inserted in the Advanced Test Reactor (ATR) in early 2015 in zirconium alloy cladding. The first of these rodlets are expected to be removed in about 2017. Key characteristics to be determined include verification of the centerline temperature calculations, thermal conductivity, fission gas release, swelling and degree of amorphization. Waterproofed UN pellets have achieved >94% density for a 32% U 3Si 2/68% UN composite pellet at Texas A&M University. This represents a U235 increase of about 31% over current UO 2 pellets. Pellets and powders of UO 2, UN, and U 3Si 2the were tested by Westinghouse and Los Alamos National Laboratory (LANL) using differential scanning calorimetry to determine what their steam and 20% oxygen corrosion temperatures were as compared to UO 2. Cold spray application of either the amorphous steel or the Ti 2AlC was successful in forming an adherent ~20 micron coating that remained after testing at 420°C in a steam autoclave. Tests at 1200°C in 100% steam on coatings for Zr alloy have not been successful, possibly due to the low density of the coatings which allowed steam transport to the base zirconium metal. Significant modeling and testing has been carried out for the SiC/SiC composite/SiC monolith structures. A structure with the monolith on the outside and composite on the inside was developed which is the current baseline structure and a SiC to SiC tube closure approach. Permeability tests and mechanical tests were developed to verify the operation of the SiC cladding. Steam autoclave (420°C), high temperature (1200°C) flowing steam tests and quench tests were carried out with minimal corrosion, mechanical or hermeticity degradation effect on the SiC cladding or end plug closure. However, in-reactor loop tests carried out in the MIT reactor indicated an unacceptable degree of corrosion, likely due to the corrosive effect of radiolysis products which attacked the SiC.« less

Burst Ductility of Zirconium Clads: The Defining Role of Residual Stress

NASA Astrophysics Data System (ADS)

Kumar, Gulshan; Kanjarla, A. K.; Lodh, Arijit; Singh, Jaiveer; Singh, Ramesh; Srivastava, D.; Dey, G. K.; Saibaba, N.; Doherty, R. D.; Samajdar, Indradev

2016-08-01

Closed end burst tests, using room temperature water as pressurizing medium, were performed on a number of industrially produced zirconium (Zr) clads. A total of 31 samples were selected based on observed differences in burst ductility. The latter was represented as total circumferential elongation or TCE. The selected samples, with a range of TCE values (5 to 35 pct), did not show any correlation with mechanical properties along axial direction, microstructural parameters, crystallographic textures, and outer tube-surface normal ( σ 11) and shear ( τ 13) components of the residual stress matrix. TCEs, however, had a clear correlation with hydrostatic residual stress ( P h), as estimated from tri-axial stress analysis on the outer tube surface. Estimated P h also scaled with measured normal stress ( σ 33) at the tube cross section. An elastic-plastic finite element model with ductile damage failure criterion was developed to understand the burst mechanism of zirconium clads. Experimentally measured P h gradients were imposed on a solid element continuum finite element (FE) simulation to mimic the residual stresses present prior to pressurization. Trends in experimental TCEs were also brought out with computationally efficient shell element-based FE simulations imposing the outer tube-surface P h values. Suitable components of the residual stress matrix thus determined the burst performance of the Zr clads.

Metallic glass alloys of Zr, Ti, Cu and Ni

DOEpatents

Lin, X.; Peker, A.; Johnson, W.L.

1997-04-08

At least quaternary alloys form metallic glass upon cooling below the glass transition temperature at a rate less than 10{sup 3} K/s. Such alloys comprise titanium from 19 to 41 atomic percent, an early transition metal (ETM) from 4 to 21 atomic percent and copper plus a late transition metal (LTM) from 49 to 64 atomic percent. The ETM comprises zirconium and/or hafnium. The LTM comprises cobalt and/or nickel. The composition is further constrained such that the product of the copper plus LTM times the atomic proportion of LTM relative to the copper is from 2 to 14. The atomic percentage of ETM is less than 10 when the atomic percentage of titanium is as high as 41, and may be as large as 21 when the atomic percentage of titanium is as low as 24. Furthermore, when the total of copper and LTM are low, the amount of LTM present must be further limited. Another group of glass forming alloys has the formula (ETM{sub 1{minus}x}Ti{sub x}){sub a} Cu{sub b} (Ni{sub 1{minus}y}Co{sub y}){sub c} wherein x is from 0.1 to 0.3, y{center_dot}c is from 0 to 18, a is from 47 to 67, b is from 8 to 42, and c is from 4 to 37. This definition of the alloys has additional constraints on the range of copper content, b. 2 figs.

Metallic glass alloys of Zr, Ti, Cu and Ni

DOEpatents

Lin, Xianghong; Peker, Atakan; Johnson, William L.

1997-01-01

At least quaternary alloys form metallic glass upon cooling below the glass transition temperature at a rate less than 10.sup.3 K/s. Such alloys comprise titanium from 19 to 41 atomic percent, an early transition metal (ETM) from 4 to 21 atomic percent and copper plus a late transition metal (LTM) from 49 to 64 atomic percent. The ETM comprises zirconium and/or hafnium. The LTM comprises cobalt and/or nickel. The composition is further constrained such that the product of the copper plus LTM times the atomic proportion of LTM relative to the copper is from 2 to 14. The atomic percentage of ETM is less than 10 when the atomic percentage of titanium is as high as 41, and may be as large as 21 when the atomic percentage of titanium is as low as 24. Furthermore, when the total of copper and LTM are low, the amount of LTM present must be further limited. Another group of glass forming alloys has the formula (ETM.sub.1-x Ti.sub.x).sub.a Cu.sub.b (Ni.sub.1-y Co.sub.y).sub.c wherein x is from 0.1 to 0.3, y.cndot.c is from 0 to 18, a is from 47 to 67, b is from 8 to 42, and c is from 4 to 37. This definition of the alloys has additional constraints on the range of copper content, b.

The John Charnley Award Paper. The role of joint fluid in the tribology of total joint arthroplasty.

PubMed

Mazzucco, Daniel; Spector, Myron

2004-12-01

The effect of joint fluid on the tribology (ie, lubrication, friction, and wear) of total hip arthroplasty has not yet been investigated adequately. In the current study, a friction assay was used to assess four hypotheses relating to the effect of human joint fluid and its principal components on the articulation of metal-on-polyethylene. First, joint fluid was found to produce a widely varying amount of friction between cobalt-chromium and polyethylene; this range exceeded the range produced when the articulation was lubricated by water or bovine serum. Second, it was shown that hyaluronic acid, phospholipid, albumin, and gamma-globulin were not acting as boundary lubricants, but that one or more other proteins (as yet unidentified) were responsible for reducing friction in this couple. Third, lower friction was found when oxidized zirconium alloy replaced cobalt-chromium as a bearing surface on polyethylene. Finally, a pilot study suggested that lubricin, which contributes to cartilage-on-cartilage lubrication, is not a protein responsible for the tribological variabiation found among joint fluid samples. The current study showed that joint fluid is a patient factor that influences the tribology of metal-on-polyethylene arthroplasty.

Intermetallic alloy welding wires and method for fabricating the same

DOEpatents

Santella, M.L.; Sikka, V.K.

1996-06-11

Welding wires for welding together intermetallic alloys of nickel aluminides, nickel-iron aluminides, iron aluminides, or titanium aluminides, and preferably including additional alloying constituents are fabricated as two-component, clad structures in which one component contains the primary alloying constituent(s) except for aluminum and the other component contains the aluminum constituent. This two-component approach for fabricating the welding wire overcomes the difficulties associated with mechanically forming welding wires from intermetallic alloys which possess high strength and limited ductilities at elevated temperatures normally employed in conventional metal working processes. The composition of the clad welding wires is readily tailored so that the welding wire composition when melted will form an alloy defined by the weld deposit which substantially corresponds to the composition of the intermetallic alloy being joined. 4 figs.

Intermetallic alloy welding wires and method for fabricating the same

DOEpatents

Santella, Michael L.; Sikka, Vinod K.

1996-01-01

Welding wires for welding together intermetallic alloys of nickel aluminides, nickel-iron aluminides, iron aluminides, or titanium aluminides, and preferably including additional alloying constituents are fabricated as two-component, clad structures in which one component contains the primary alloying constituent(s) except for aluminum and the other component contains the aluminum constituent. This two-component approach for fabricating the welding wire overcomes the difficulties associated with mechanically forming welding wires from intermetallic alloys which possess high strength and limited ductilities at elevated temperatures normally employed in conventional metal working processes. The composition of the clad welding wires is readily tailored so that the welding wire composition when melted will form an alloy defined by the weld deposit which substantially corresponds to the composition of the intermetallic alloy being joined.

Effects of heat treatment on U-Mo fuel foils with a zirconium diffusion barrier

NASA Astrophysics Data System (ADS)

Jue, Jan-Fong; Trowbridge, Tammy L.; Breckenridge, Cynthia R.; Moore, Glenn A.; Meyer, Mitchell K.; Keiser, Dennis D.

2015-05-01

A monolith fuel design based on U-Mo alloy has been selected as the fuel type for conversion of the United States' high performance research reactors (HPRRs) from highly enriched uranium (HEU) to low-enriched uranium (LEU). In this fuel design, a thin layer of zirconium is used to eliminate the direct interaction between the U-Mo fuel meat and the aluminum-alloy cladding during irradiation. The co-rolling process used to bond the Zr barrier layer to the U-Mo foil during fabrication alters the microstructure of both the U-10Mo fuel meat and the U-Mo/Zr interface. This work studied the effects of post-rolling annealing treatment on the microstructure of the co-rolled U-Mo fuel meat and the U-Mo/Zr interaction layer. Microscopic characterization shows that the grain size of U-Mo fuel meat increases with the annealing temperature, as expected. The grain sizes were ∼9, ∼13, and ∼20 μm for annealing temperature of 650, 750, and 850 °C, respectively. No abnormal grain growth was observed. The U-Mo/Zr interaction-layer thickness increased with the annealing temperature with an Arrhenius constant for growth of 184 kJ/mole, consistent with a previous diffusion-couple study. The interaction layer thickness was 3.2 ± 0.5 μm, 11.1 ± 2.1 μm, 27.1 ± 0.9 μm for annealing temperature of 650, 750, to 850 °C, respectively. The homogeneity of Mo improves with post rolling annealing temperature and with U-Mo coupon homogenization. The phases in the Zr/U-Mo interaction layer produced by co-rolling, however, differ from those reported in the previous diffusion couple studies.

Effects of heat treatment on U–Mo fuel foils with a zirconium diffusion barrier

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jue, Jan-Fong; Trowbridge, Tammy L.; Breckenridge, Cynthia R.

A monolith fuel design based on U–Mo alloy has been selected as the fuel type for conversion of the United States’ high performance research reactors (HPRRs) from highly enriched uranium (HEU) to low-enriched uranium (LEU). In this fuel design, a thin layer of zirconium is used to eliminate the direct interaction between the U–Mo fuel meat and the aluminum-alloy cladding during irradiation. The co-rolling process used to bond the Zr barrier layer to the U–Mo foil during fabrication alters the microstructure of both the U–10Mo fuel meat and the U–Mo/Zr interface. This work studied the effects of post-rolling annealing treatmentmore » on the microstructure of the co-rolled U–Mo fuel meat and the U–Mo/Zr interaction layer. Microscopic characterization shows that the grain size of U–Mo fuel meat increases with the annealing temperature, as expected. The grain sizes were ~9, ~13, and ~20 μm for annealing temperature of 650, 750, and 850 °C, respectively. No abnormal grain growth was observed. The U–Mo/Zr interaction-layer thickness increased with the annealing temperature with an Arrhenius constant for growth of 184 kJ/mole, consistent with a previous diffusion-couple study. The interaction layer thickness was 3.2 ± 0.5 μm, 11.1 ± 2.1 μm, 27.1 ± 0.9 μm for annealing temperature of 650, 750, to 850 °C, respectively. The homogeneity of Mo improves with post rolling annealing temperature and with U–Mo coupon homogenization. The phases in the Zr/U–Mo interaction layer produced by co-rolling, however, differ from those reported in the previous diffusion couple studies.« less

Thermochemical Compatibility and Oxidation Resistance of Advanced LWR Fuel Cladding

DOE PAGES

Besmann, T. M.; Yamamoto, Y.; Unocic, K. A.

2016-06-21

We assessed the thermochemical compatibility of potential replacement cladding materials for zirconium alloys in light water reactors. Considered were FeCrAl steel (similar to Kanthal APMT), Nb-1%Zr (similar to PWC-11), and a hybrid SiC-composite with a metallic barrier layer. The niobium alloy was also seen as requiring an oxidation protective layer, and a diffusion silicide was investigated. Metallic barrier layers for the SiC-composite reviewed included a FeCrAl alloy, Nb-1%Zr, and chromium. Thermochemical calculations were performed to determine oxidation behavior of the materials in steam, and for hybrid SiC-composites possible interactions between the metallic layer and SiC. Additionally, experimental exposures of SiC-alloymore » reaction couples at 673K, 1073K, and 1273K for 168 h in an inert atmosphere were made and microanalysis performed. Whereas all materials were determined to oxidize under higher oxygen partial pressures in the steam environment, these varied by material with expected protective oxides forming. Finally, the computed and experimental results indicate the formation of liquid phase eutectic in the FeCrAl-SiC system at the higher temperatures.« less

Plasma-sprayed zirconia gas path seal technology: A state-of-the-art review

NASA Technical Reports Server (NTRS)

Bill, R. C.

1979-01-01

The benefits derived from application of ceramic materials to high pressure turbine gas path seal components are described and the developmental backgrounds of various approaches are reviewed. The most fully developed approaches are those employing plasma sprayed zirconium oxide as the ceramic material. Prevention of cracking and spalling of the zirconium oxide under cyclic thermal shock conditions imposed by the engine operating cycle is the most immediate problem to be solved before implementation is undertaken. Three promising approaches to improving cyclic thermal shock resistance are described and comparative rig performance of each are reviewed. Advanced concepts showing potential for performance improvements are described.

Improvements to the strength and corrosion resistance of aluminum-magnesium-manganese alloys of near-AA5083 chemistry

NASA Astrophysics Data System (ADS)

Carroll, Mark Christopher

Aluminum alloys of the 5000 series (AI-Mg-Mn) are extremely popular in a wide range of applications that call for a balance of moderately high strength, good corrosion resistance, and light weight, all at a moderate cost. One of the most popular 5000 series alloys is designated A1-5083, containing, in addition to aluminum, approximately 4 wt% magnesium and 0.7 wt% manganese. In order to increase the range of versatility of this particular alloy, a number of modifications have been examined that will potentially improve the strength and corrosion resistance characteristics while maintaining a chemical composition that is very close to the proven 5083 alloy. The strength of the 5083-based alloys under study are investigated with two goals in mind---to maximize the potential strength characteristics in a "standard" 5083 form through changes in minor processing parameters or through minor alloying additions. Increasing the standard alloy's potential is possible through improved efficiency of "preprocessing" heat treatments that maximize the homogeneous dispersion of secondary manganese-based particles. For the modified alloy study, additions of scandium and zirconium are shown to improve strength not only by forming secondary particles in the alloy, but also through substitutional solid solution strengthening, even when added at very small levels. Corrosion resistance of these 5083-based alloys is investigated once again through minor alloying additions; specifically zinc, copper, and silver. Zinc is particularly effective in that it changes the corrosion-susceptible binary aluminum-magnesium phase that would otherwise form on grain boundaries following exposure to moderately elevated temperatures for extended periods of time to a ternary aluminum-magnesium-zinc phase. This chemical composition of this ternary phase that forms following zinc additions can be further altered through minor additions of copper and silver. By determining threshold levels for these modifications while maintaining a chemical composition that is very near that of standard Al-5083, it can be shown that even minor modifications to processing and alloying parameters can have a favorable effect on the final bulk properties of the alloy. The increased range of strength and corrosion resistance of these lightly modified alloys make them more attractive in a broadened range of potential applications.

Strength and fracture of uranium, plutonium and several their alloys under shock wave loading

NASA Astrophysics Data System (ADS)

Golubev, V. K.

2012-08-01

Results on studying the spall fracture of uranium, plutonium and several their alloys under shock wave loading are presented in the paper. The problems of influence of initial temperature in a range of - 196 - 800∘C and loading time on the spall strength and failure character of uranium and two its alloys with molybdenum and both molybdenum and zirconium were studied. The results for plutonium and its alloy with gallium were obtained at a normal temperature and in a temperature range of 40-315∘C, respectively. The majority of tests were conducted with the samples in the form of disks 4 mm in thickness. They were loaded by the impact of aluminum plates 4 mm thick through a copper screen 12 mm thick serving as the cover or bottom part of a special container. The character of spall failure of materials and the damage degree of samples were observed on the longitudinal metallographic sections of recovered samples. For a concrete test temperature, the impact velocity was sequentially changed and therefore the loading conditions corresponding to the consecutive transition from microdamage nucleation up to complete macroscopic spall fracture were determined. The conditions of shock wave loading were calculated using an elastic-plastic computer program. The comparison of obtained results with the data of other researchers on the spall fracture of examined materials was conducted.

An overview of recent advances in designing orthopedic and craniofacial implants

PubMed Central

Mantripragada, Venkata P.; Lecka-Czernik, Beata; Ebraheim, Nabil A.; Jayasuriya, Ambalangodage C.

2016-01-01

Great deal of research is still going on in the field of orthopedic and craniofacial implant development to resolve various issues being faced by the industry today. Despite several disadvantages of the metallic implants, they continue to be used, primarily because of their superior mechanical properties. In order to minimize the harmful effects of the metallic implants and its by-products, several modifications are being made to these materials, for instance nickel-free stainless steel, cobalt-chromium and titanium alloys are being introduced to eliminate the toxic effects of nickel being released from the alloys, introduce metallic implants with lower modulus, reduce the cost of these alloys by replacing rare elements with less expensive elements etc. New alloys like tantalum, niobium, zirconium, and magnesium are receiving attention given their satisfying mechanical and biological properties. Non-oxide ceramics like silicon nitride and silicon carbide are being currently developed as a promising implant material possessing a combination of properties such as good wear and corrosion resistance, increased ductility, good fracture and creep resistance, and relatively high hardness in comparison to alumina. Polymer/magnesium composites are being developed to improve mechanical properties as well as retain polymer’s property of degradation. Recent advances in orthobiologics are proving interesting as well. This paper thus deals with the latest improvements being made to the existing implant materials and includes new materials being introduced in the field of biomaterials. PMID:23766134

Biocompatibility evaluation of sputtered zirconium-based thin film metallic glass-coated steels.

PubMed

Subramanian, Balasubramanian; Maruthamuthu, Sundaram; Rajan, Senthilperumal Thanka

2015-01-01

Thin film metallic glasses comprised of Zr48Cu36Al8Ag8 (at.%) of approximately 1.5 μm and 3 μm in thickness were prepared using magnetron sputtering onto medical grade 316L stainless steel. Their structural and mechanical properties, in vitro corrosion, and antimicrobial activity were analyzed. The amorphous thin film metallic glasses consisted of a single glassy phase, with an absence of any detectable peaks corresponding to crystalline phases. Elemental composition close to the target alloy was noted from EDAX analysis of the thin film. The surface morphology of the film showed a smooth surface on scanning electron microscopy and atomic force microscopy. In vitro electrochemical corrosion studies indicated that the zirconium-based metallic glass could withstand body fluid, showing superior resistance to corrosion and electrochemical stability. Interactions between the coated surface and bacteria were investigated by agar diffusion, solution suspension, and wet interfacial contact methods. The results indicated a clear zone of inhibition against the growth of microorganisms such as Escherichia coli and Staphylococcus aureus, confirming the antimicrobial activity of the thin film metallic glasses. Cytotoxicity studies using L929 fibroblast cells showed these coatings to be noncytotoxic in nature.

Biocompatibility evaluation of sputtered zirconium-based thin film metallic glass-coated steels

PubMed Central

Subramanian, Balasubramanian; Maruthamuthu, Sundaram; Rajan, Senthilperumal Thanka

2015-01-01

Thin film metallic glasses comprised of Zr48Cu36Al8Ag8 (at.%) of approximately 1.5 μm and 3 μm in thickness were prepared using magnetron sputtering onto medical grade 316L stainless steel. Their structural and mechanical properties, in vitro corrosion, and antimicrobial activity were analyzed. The amorphous thin film metallic glasses consisted of a single glassy phase, with an absence of any detectable peaks corresponding to crystalline phases. Elemental composition close to the target alloy was noted from EDAX analysis of the thin film. The surface morphology of the film showed a smooth surface on scanning electron microscopy and atomic force microscopy. In vitro electrochemical corrosion studies indicated that the zirconium-based metallic glass could withstand body fluid, showing superior resistance to corrosion and electrochemical stability. Interactions between the coated surface and bacteria were investigated by agar diffusion, solution suspension, and wet interfacial contact methods. The results indicated a clear zone of inhibition against the growth of microorganisms such as Escherichia coli and Staphylococcus aureus, confirming the antimicrobial activity of the thin film metallic glasses. Cytotoxicity studies using L929 fibroblast cells showed these coatings to be noncytotoxic in nature. PMID:26491304

Magnesium Alloy WE43 and WE43-T5 - Mechanical and Thermal Properties

NASA Astrophysics Data System (ADS)

Xiang, Chongchen

Magnesium alloys are promising in aerospace, automotive and electronic industries due to low density, high specific strength and excellent machinability. A rare earth element alloy (WE43) is studied in as cast and heat treated conditions. Multiscale characterization is conducted to understand the nanomechanical response using a nanoindentor and microscale behavior using tensile tests. Further, compressive characterization is conducted across six orders of strain rate magnitudes from 10-3 to 3x103 s -1 under the range of liquid nitrogen (-196°C) to room temperature (25°C). Based on the results, a constitutive model is developed to estimate the plastic behavior of as-cast WE43 and WE43-T5 at different strain rates and under different temperatures. In addition, dynamic properties are studied using a dynamic mechanical analyzer at 1-100 Hz loading frequencies and the temperature range from 35°C to 500°C. Only Yttrium-rich cuboidal phase and zirconium-rich phase were present in WE43-T5 alloy and the eutectic phase was absent. Also, the grain size was reduced due to the hot rolling process. The difference in microstructure reflects into the mechanical properties. WE43-T5 specimens have improved mechanical properties over the as-cast alloy. Two transition temperatures are found at 210 and 250°C based on the storage and loss moduli results. The Mg24Y5 peak is found in the high temperature x-ray diffraction results along with a new Mg12Nd peak at those two temperature points. The corrosion behavior, studied by 7-day immersion in 3.5% NaCl solution, shows that the heat treated alloy has significantly lower corrosion rate than the as-cast alloy due to the absence of the eutectic mixture in the microstructure. With rapidly growing applications of magnesium alloys, particularly with rare earth elements, this study is expected to provide critical data and structure-property correlations that will help the scientific community.

Microstructural Modeling of Dynamic Intergranular and Transgranular Fracture Modes in Zircaloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohammed, I.; Zikry, M.A.; Ziaei, S.

2017-04-01

In this time period, we have continued to focus on (i) refining the thermo-mechanical fracture model for zirconium (Zr) alloys subjected to large deformations and high temperatures that accounts for the cracking of ZrH and ZrH2 hydrides, (ii) formulating a framework to account intergranular fracture due to iodine diffusion and pit formation in grain-boundaries (GBs). Our future objectives are focused on extending to a combined population of ZrH and ZrH2 populations and understanding how thermo-mechanical behavior affects hydride reorientation and cracking. We will also refine the intergranular failure mechanisms for grain boundaries with pits.

Influence of Deformation Mechanisms on the Mechanical Behavior of Metals and Alloys: Experiments, Constitutive Modeling, and Validation

NASA Astrophysics Data System (ADS)

Gray, G. T.; Cerreta, E.; Chen, Shuh Rong; Maudlin, P. J.

2004-06-01

Jim Williams has made seminal contributions to the field of structure / property relations and its controlling effects on the mechanical behavior of metals and alloys. This talk will discuss experimental results illustrating the role of interstitial content, grain size, texture, temperature, and strain rate on the operative deformation mechanisms, mechanical behavior, and substructure evolution in titanium, zirconium, hafnium, and rhenium. Increasing grain size is shown to significantly decrease the dynamic flow strength of Ti and Zr while increasing work-hardening rates due to an increased incidence of deformation twinning. Increasing oxygen interstitial content is shown to significantly alter both the constitutive response and α-ω shock-induced phase transition in Zr. The influence of crystallographic texture on the mechanical behavior in Ti, Zr, and Hf is discussed in terms of slip system and deformation twinning activity. An example of the utility of incorporation of operative deformation mechanisms into a polycrystalline plasticity constitutive model and validation using Taylor cylinder impact testing is presented.

Chemical compatibility of cartridge materials

NASA Technical Reports Server (NTRS)

Wilcox, Roy C.; Zee, R. H.

1991-01-01

This twelve month progress report deals with the chemical compatibility of semiconductor crystals grown in zero gravity. Specifically, it studies the chemical compatibility between TZM, a molybdenum alloy containing titanium and zirconium, and WC 103, a titanium alloy containing Niobium and Hafnium, and Gallium arsenide (GaAs) and Cadmium Zinc Tellurite (CdZnTe). Due to the health hazards involved, three approaches were used to study the chemical compatibility between the semiconductor and cartridge materials: reaction retort, thermogravimetric analysis, and bulk cylindrical cartridge containers. A scanning electron microscope with an energy dispersive X-ray analyzer was used to examine all samples after testing. The first conclusion drawn is that reaction rates with TZM were not nearly as great as they were with WC 103. Second, the total reaction between GaAs and WC 103 was almost twice that with TZM. Therefore, even though WC 103 is easier to fabricate, at least half of the cartridge thickness will be degraded if contact is made with one of the semiconductor materials leading to a loss of strength properties.

Feasibility of EB Welded Hastelloy X and Combination of Refractory Metals

NASA Technical Reports Server (NTRS)

Martinez, Diana A.

2004-01-01

As NASA continues to expand its horizon, exploration and discovery creates the need of advancement in technology. The Jupiter Icy Moon Orbiter's (JIMO) mission to explore and document the outer surfaces, rate the possibility of holding potential life forms, etc. within the three moons (Callisto, Ganymede, and Europa) proves to be challenging. The orbiter itself consists of many sections including: the nuclear reactor and the power conversion system, the radiator panels, and the thrusters and antenna. The nuclear reactor serves as a power source, and if successfully developed, can operate for extended periods. During the duration of my tenure at NASA Glenn Research Center's (NASA GRC) Advanced Metallics Branch, I was assigned to assist Frank J. Ritzert on analyzing the feasibility of the Electron Beam Welded Hastelloy X (HX), a nickel-based superalloy, to Niobium- 1 %Zirconium (Nb-1 Zr) and other refractory metals/alloys including Tantalum, Molybdenum, Tungsten, and Rhenium alloys. This welding technique is going to be used for the nuclear reactor within JIMO.

Effects of Zr Addition on Strengthening Mechanisms of Al-Alloyed High-Cr ODS Steels.

PubMed

Ren, Jian; Yu, Liming; Liu, Yongchang; Liu, Chenxi; Li, Huijun; Wu, Jiefeng

2018-01-12

Oxide dispersion strengthened (ODS) steels with different contents of zirconium (denoted as 16Cr ODS, 16Cr-0.3Zr ODS and 16Cr-0.6Zr ODS) were fabricated to investigate the effects of Zr on strengthening mechanism of Al-alloyed 16Cr ODS steel. Electron backscatter diffraction (EBSD) results show that the mean grain size of ODS steels could be decreased by Zr addition. Transmission electron microscope (TEM) results indicate that Zr addition could increase the number density but decrease the mean diameter and inter-particle spacing of oxide particles. Furthermore, it is also found that in addition to Y-Al-O nanoparticles, Y-Zr-O oxides with finer size were observed in 16Cr-0.3Zr ODS and 16Cr-0.6Zr ODS steels. These changes in microstructure significantly increase the yield strength (YS) and ultimate tensile strength (UTS) of ODS steels through mechanisms of grain boundary strengthening and dispersion strengthening.

Fatigue crack propagation of nickel-base superalloys at 650 deg C

NASA Technical Reports Server (NTRS)

Gayda, J.; Gabb, T. P.; Miner, R. V.

1988-01-01

The 650 C fatigue crack propagation behavior of two nickel-base superalloys, Rene 95 and Waspaloy, is studied with particular emphasis placed on understanding the roles of creep, environment, and two key grain boundary alloying additions, boron and zirconium. Comparison of air and vacuum data shows the air environment to be detrimental over a wide range of frequencies for both alloys. More in-depth analysis on Rene 95 shows at lower frequencies, such as 0.02 Hz, failure in air occurs by intergranular, environmentally-assisted creep crack growth, while at higher frequencies, up to 5.0 Hz, environmental interaction are still evident but creep effects are minimized. The effect of B and Zr in Waspaloy is found to be important where environmental and/or creep interactions are presented. In those instances, removal of B and Zr dramatically increases crack growth and it is therefore plausible that effective dilution of these elements may explain a previously observed trend in which crack growth rates increase with decreasing grain size.

Fatigue crack propagation of nickel-base superalloys at 650 deg C

NASA Technical Reports Server (NTRS)

Gayda, J.; Gabb, T. P.; Miner, R. V.

1985-01-01

The 650 C fatigue crack propagation behavior of two nickel-base superalloys, Rene 95 and Waspaloy, is studied with particular emphasis placed on understanding the roles of creep, environment, and two key grain boundary alloying additions, boron and zirconium. Comparison of air and vacuum data shows the air environment to be detrimental over a wide range of frequencies for both alloys. More in-depth analysis on Rene 95 shows at lower frequencies, such as 0.02 Hz, failure in air occurs by intergranular, environmentally-assisted creep crack growth, while at higher frequencies, up to 5.0 Hz, environmental interactions are still evident but creep effects are minimized. The effect of B and Zr in Waspaloy is found to be important where environmental and/or creep interactions are presented. In those instances, removal of B and Zr dramatically increases crack growth and it is therefore plausible that effective dilution of these elements may explain a previously observed trend in which crack growth rates increase with decreasing grain size.

Nuclear fuel elements having a composite cladding

DOEpatents

Gordon, Gerald M.; Cowan, II, Robert L.; Davies, John H.

1983-09-20

An improved nuclear fuel element is disclosed for use in the core of nuclear reactors. The improved nuclear fuel element has a composite cladding of an outer portion forming a substrate having on the inside surface a metal layer selected from the group consisting of copper, nickel, iron and alloys of the foregoing with a gap between the composite cladding and the core of nuclear fuel. The nuclear fuel element comprises a container of the elongated composite cladding, a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container and a nuclear fuel material retaining means positioned in the cavity. The metal layer of the composite cladding prevents perforations or failures in the cladding substrate from stress corrosion cracking or from fuel pellet-cladding interaction or both. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy.

Effects of Zr Addition on Strengthening Mechanisms of Al-Alloyed High-Cr ODS Steels

PubMed Central

Ren, Jian; Yu, Liming; Liu, Yongchang; Liu, Chenxi; Li, Huijun; Wu, Jiefeng

2018-01-01

Oxide dispersion strengthened (ODS) steels with different contents of zirconium (denoted as 16Cr ODS, 16Cr-0.3Zr ODS and 16Cr-0.6Zr ODS) were fabricated to investigate the effects of Zr on strengthening mechanism of Al-alloyed 16Cr ODS steel. Electron backscatter diffraction (EBSD) results show that the mean grain size of ODS steels could be decreased by Zr addition. Transmission electron microscope (TEM) results indicate that Zr addition could increase the number density but decrease the mean diameter and inter-particle spacing of oxide particles. Furthermore, it is also found that in addition to Y-Al-O nanoparticles, Y-Zr-O oxides with finer size were observed in 16Cr-0.3Zr ODS and 16Cr-0.6Zr ODS steels. These changes in microstructure significantly increase the yield strength (YS) and ultimate tensile strength (UTS) of ODS steels through mechanisms of grain boundary strengthening and dispersion strengthening. PMID:29329260

Irradiation of Wrought FeCrAl Tubes in the High Flux Isotope Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Linton, Kory D.; Field, Kevin G.; Petrie, Christian M.

The Advanced Fuels Campaign within the Nuclear Technology Research and Development program of the Department of Energy Office of Nuclear Energy is seeking to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are one of the leading candidate materials for fuel cladding to replace traditional zirconium alloys because of the superior oxidation resistance of FeCrAl. However, there are still some unresolved questions regarding irradiation effects on the microstructure and mechanical properties of FeCrAl at end-of-life dose levels. In particular, there are concerns related to irradiation-induced embrittlement of FeCrAl alloys due to secondary phase formation. Tomore » address this issue, Oak Ridge National Laboratory has developed a new experimental design to irradiate shortened cladding tube specimens with representative 17×17 array pressurized water reactor diameter and thickness in the High Flux Isotope Reactor (HFIR) under relevant temperatures (300–350°C). Post-irradiation examination will include studies of dimensional change, microstructural changes, and mechanical performance. This report briefly summarizes the capsule design concept and the irradiation test matrix for six rabbit capsules. Each rabbit contains two FeCrAl alloy tube specimens. The specimens include Generation I and Generation II FeCrAl alloys with varying processing conditions, Cr concentrations, and minor alloying elements. The rabbits were successfully assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. Pictures of the rabbit assembly process and detailed dimensional inspection of select specimens are included in this report. The rabbits were inserted into HFIR starting in cycle 472 (May 2017).« less

Method for preparing hydrous zirconium oxide gels and spherules

DOEpatents

Collins, Jack L.

2003-08-05

Methods for preparing hydrous zirconium oxide spherules, hydrous zirconium oxide gels such as gel slabs, films, capillary and electrophoresis gels, zirconium monohydrogen phosphate spherules, hydrous zirconium oxide spherules having suspendable particles homogeneously embedded within to form a composite sorbent, zirconium monohydrogen phosphate spherules having suspendable particles of at least one different sorbent homogeneously embedded within to form a composite sorbent having a desired crystallinity, zirconium oxide spherules having suspendable particles homogeneously embedded within to form a composite, hydrous zirconium oxide fiber materials, zirconium oxide fiber materials, hydrous zirconium oxide fiber materials having suspendable particles homogeneously embedded within to form a composite, zirconium oxide fiber materials having suspendable particles homogeneously embedded within to form a composite and spherules of barium zirconate. The hydrous zirconium oxide spherules and gel forms prepared by the gel-sphere, internal gelation process are useful as inorganic ion exchangers, catalysts, getters and ceramics.

From Laboratory Research to a Clinical Trial

PubMed Central

Michels, Harold T.; Keevil, C. William; Salgado, Cassandra D.; Schmidt, Michael G.

2015-01-01

Objective: This is a translational science article that discusses copper alloys as antimicrobial environmental surfaces. Bacteria die when they come in contact with copper alloys in laboratory tests. Components made of copper alloys were also found to be efficacious in a clinical trial. Background: There are indications that bacteria found on frequently touched environmental surfaces play a role in infection transmission. Methods: In laboratory testing, copper alloy samples were inoculated with bacteria. In clinical trials, the amount of live bacteria on the surfaces of hospital components made of copper alloys, as well as those made from standard materials, was measured. Finally, infection rates were tracked in the hospital rooms with the copper components and compared to those found in the rooms containing the standard components. Results: Greater than a 99.9% reduction in live bacteria was realized in laboratory tests. In the clinical trials, an 83% reduction in bacteria was seen on the copper alloy components, when compared to the surfaces made from standard materials in the control rooms. Finally, the infection rates were found to be reduced by 58% in patient rooms with components made of copper, when compared to patients' rooms with components made of standard materials. Conclusions: Bacteria die on copper alloy surfaces in both the laboratory and the hospital rooms. Infection rates were lowered in those hospital rooms containing copper components. Thus, based on the presented information, the placement of copper alloy components, in the built environment, may have the potential to reduce not only hospital-acquired infections but also patient treatment costs. PMID:26163568

Microstructure and Mechanical Properties of Nano-Size Zirconium Carbide Dispersion Strengthened Tungsten Alloys Fabricated by Spark Plasma Sintering Method

NASA Astrophysics Data System (ADS)

Xie, Zhuoming; Liu, Rui; Fang, Qianfeng; Zhang, Tao; Jiang, Yan; Wang, Xianping; Liu, Changsong

2015-12-01

W-(0.2, 0.5, 1.0)wt% ZrC alloys with a relative density above 97.5% were fabricated through the spark plasma sintering (SPS) method. The grain size of W-1.0wt% ZrC is about 2.7 μm, smaller than that of pure W and W-(0.2, 0.5)wt% ZrC. The results indicated that the W-ZrC alloys exhibit higher hardness at room temperature, higher tensile strength at high temperature, and a lower ductile to brittle transition temperature (DBTT) than pure W. The tensile strength and total elongation of W-0.5wt% ZrC alloy at 700 °C is 535 MPa and 24.8%, which are respectively 59% and 114% higher than those of pure W (337 MPa, 11.6%). The DBTT of W-(0.2, 0.5, 1.0)wt% ZrC materials is in the range of 500°C-600°C, which is about 100 °C lower than that of pure W. Based on microstructure analysis, the improved mechanical properties of the W-ZrC alloys were suggested to originate from the enhanced grain boundary cohesion by ZrC capturing the impurity oxygen in tungsten and nano-size ZrC dispersion strengthening. supported by the Innovation Program of Chinese Academy of Sciences (No. KJCX2-YW-N35), the National Magnetic Confinement Fusion Science Program of China (No. 2011GB108004), National Natural Science Foundation of China (Nos. 51301164, 11075177, 11274305), and Anhui Provincial Natural Science Foundation of China (No. 1408085QE77)

Precipitate Evolution and Strengthening in Supersaturated Rapidly Solidified Al-Sc-Zr Alloys

NASA Astrophysics Data System (ADS)

Deane, Kyle; Kampe, S. L.; Swenson, Douglas; Sanders, P. G.

2017-04-01

Because of the low diffusivities of scandium and zirconium in aluminum, trialuminide precipitates containing these elements have been reported to possess excellent thermal stability at temperatures of 573 K (300 °C) and higher. However, the relatively low equilibrium solubilities of these elements in aluminum limit the achievable phase fraction and, in turn, strengthening contributions from these precipitates. One method of circumventing this limitation involves the use of rapid solidification techniques to suppress the initial formation of precipitates in alloys containing higher solute compositions. This work specifically discusses the fabrication of supersaturated Al-Sc, Al-Zr, and Al-Sc-Zr alloys via melt spinning, in which supersaturations of at least 0.55 at. pct Zr and 0.8 at. pct Sc are shown to be attainable through XRD analysis. The resulting ribbons were subjected to a multistep aging heat treatment in order to encourage a core-shell precipitate morphology, the precipitate evolution behavior was monitored with XRD and TEM, and the aging behavior was observed. While aging in these alloys is shown to follow similar trends to conventionally processed materials reported in literature, with phase fraction increasing until higher aging temperatures causing a competing dissolution effect, the onset of precipitation begins at lower temperatures than previously observed and the peak hardnesses occurred at higher temperature steps due to an increased aging time associated with increased solute concentration. Peaking in strength at a higher temperature doesn't necessarily mean an increase in thermal stability, but rather emphasizes the need for intelligently designed heat treatments to take full advantage of the potential strengthening of supersaturated Al-Sc-Zr alloys.

Three-dimensional ordered titanium dioxide-zirconium dioxide film-based microfluidic device for efficient on-chip phosphopeptide enrichment.

PubMed

Zhao, De; He, Zhongyuan; Wang, Gang; Wang, Hongzhi; Zhang, Qinghong; Li, Yaogang

2016-09-15

Microfluidic technology plays a significant role in separating biomolecules, because of its miniaturization, integration, and automation. Introducing micro/nanostructured functional materials can improve the properties of microfluidic devices, and extend their application. Inverse opal has a three-dimensional ordered net-like structure. It possesses a large surface area and exhibits good mass transport, making it a good candidate for bio-separation. This study exploits inverse opal titanium dioxide-zirconium dioxide films for on-chip phosphopeptide enrichment. Titanium dioxide-zirconium dioxide inverse opal film-based microfluidic devices were constructed from templates of 270-, 340-, and 370-nm-diameter poly(methylmethacrylate) spheres. The phosphopeptide enrichments of these devices were determined by matrix-assisted laser desorption/ionization time-of-flight (MALDI-TOF) mass spectrometry. The device constructed from the 270-nm-diameter sphere template exhibited good comprehensive phosphopeptide enrichment, and was the best among these three devices. Because the size of opal template used in construction was the smallest, the inverse opal film therefore had the smallest pore sizes and the largest surface area. Enrichment by this device was also better than those of similar devices based on nanoparticle films and single component films. The titanium dioxide-zirconium dioxide inverse opal film-based device provides a promising approach for the efficient separation of various biomolecules. Copyright © 2016 Elsevier Inc. All rights reserved.

Reactivity Initiated Accident Simulation to Inform Transient Testing of Candidate Advanced Cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas R; Wysocki, Aaron J; Terrani, Kurt A

2016-01-01

Abstract. Advanced cladding materials with potentially enhanced accident tolerance will yield different light water reactor performance and safety characteristics than the present zirconium-based cladding alloys. These differences are due to different cladding material properties and responses to the transient, and to some extent, reactor physics, thermal, and hydraulic characteristics. Some of the differences in reactors physics characteristics will be driven by the fundamental properties (e.g., absorption in iron for an iron-based cladding) and others will be driven by design modifications necessitated by the candidate cladding materials (e.g., a larger fuel pellet to compensate for parasitic absorption). Potential changes in thermalmore » hydraulic limits after transition from the current zirconium-based cladding to the advanced materials will also affect the transient response of the integral fuel. This paper leverages three-dimensional reactor core simulation capabilities to inform on appropriate experimental test conditions for candidate advanced cladding materials in a control rod ejection event. These test conditions are using three-dimensional nodal kinetics simulations of a reactivity initiated accident (RIA) in a representative state-of-the-art pressurized water reactor with both nuclear-grade iron-chromium-aluminum (FeCrAl) and silicon carbide based (SiC-SiC) cladding materials. The effort yields boundary conditions for experimental mechanical tests, specifically peak cladding strain during the power pulse following the rod ejection. The impact of candidate cladding materials on the reactor kinetics behavior of RIA progression versus reference zirconium cladding is predominantly due to differences in: (1) fuel mass/volume/specific power density, (2) spectral effects due to parasitic neutron absorption, (3) control rod worth due to hardened (or softened) spectrum, and (4) initial conditions due to power peaking and neutron transport cross sections in the equilibrium cycle cores due to hardened (or softened) spectrum. This study shows minimal impact of SiC-based cladding configurations on the transient response versus reference zirconium-based cladding. However, the FeCrAl cladding response indicates similar energy deposition, but with significantly shorter pulses of higher magnitude. Therefore the FeCrAl-based cases have a more rapid fuel thermal expansion rate and the resultant pellet-cladding interaction occurs more rapidly.« less

Wear mechanisms and improvements of wear resistance in cobalt-chromium alloy femoral components in artificial total knee joints

NASA Astrophysics Data System (ADS)

Que, Like

Wear is one of the major causes of artificial total knee arthroplasty (TKA) failure. Wear debris can cause adverse reactions to the surrounding tissue which can ultimately lead to loosening of the prosthesis. The wear behavior of UHMWPE tibial components have been studied extensively, but relatively little attention has been paid to the CoCrMo femoral component. The goal of the present study was to investigate the wear mechanisms of CoCrMo femoral components, to study the effect of CoCrMo alloy surface roughness on the wear of UHMWPE, and to determine the effect of heat treatments on the wear resistance of the CoCrMo implant alloys. The surface roughness of twenty-seven retrieved CoCrMo femoral components was analyzed. A multiple station wear testing machine and a wear fixture attached to an MTS 858 bionix system were built and used for in vitro wear studies of the CoCrMo/UHMWPE bearing couple. Solution and aging treatments were applied to the CoCrMo alloys. A white light interference surface profilometer (WLISP) and a scanning electron microscope (SEM) were used to measure the surface roughness and to study wear mechanisms of CoCrMo alloy. An optical microscope was used for alloy microstructure study. X-ray diffraction tests were performed to identify alloy phase transformation after aging. The micro-structure, hardness, and wear resistance of the alloys were studied. Surface roughness was used to quantify alloy wear, and the minimum number of surface roughness measurements required to obtain a reliable and repeatable characterization of surface roughness for a worn alloy surface was determined. The surfaces of the retrieved CoCrMo femoral components appeared to be damaged by metal particles embedded in the UHMWPE tibial component and metal-on-metal wear due to UHMWPE tibial component through-wear. Surface roughness of the femoral components was not correlated with patient age, weight, sex, or length of implantation. In vitro wear tests showed that when the CoCrMo alloy surface roughness was higher than 0.022 mum Ra (surface roughness average), UHMWPE wear increased with increasing CoCrMo alloy surface roughness. Bone and poly(methyl methacrylate) (PMMA) bone cement abrasive particles created scratches on the alloy via a ploughing mechanism, and resulted in significantly rougher surfaces than controls without particles (P < 0.01). Solution treatments at 1230sp°C and 1245sp°C reduced the hardness and wear resistance of the as-cast F75 CoCrMo alloy. Aging at 700sp°C caused recrystallization of the forged F799 alloy and improved wear resistance. Thermo-mechanical treatments have the potential to increase the lifetime of artificial joints by increasing the wear resistance of CoCrMo components.

Energy Saving Melting and Revert Reduction Technology (Energy-SMARRT): Light Metals Permanent Mold Casting

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fasoyinu, Yemi

2014-03-31

Current vehicles use mostly ferrous components for structural applications. It is possible to reduce the weight of the vehicle by substituting these parts with those made from light metals such as aluminum and magnesium. Many alloys and manufacturing processes can be used to produce these light metal components and casting is known to be most economical. One of the high integrity casting processes is permanent mold casting which is the focus of this research report. Many aluminum alloy castings used in automotive applications are produced by the sand casting process. Also, aluminum-silicon (Al-Si) alloys are the most widely used alloymore » systems for automotive applications. It is possible that by using high strength aluminum alloys based on an aluminum-copper (Al-Cu) system and permanent mold casting, the performance of these components can be enhanced significantly. This will also help to further reduce the weight. However, many technological obstacles need to be overcome before using these alloys in automotive applications in an economical way. There is very limited information in the open literature on gravity and low-pressure permanent mold casting of high strength aluminum alloys. This report summarizes the results and issues encountered during the casting trials of high strength aluminum alloy 206.0 (Al-Cu alloy) and moderate strength alloy 535.0 (Al-Mg alloy). Five engineering components were cast by gravity tilt-pour or low pressure permanent mold casting processes at CanmetMATERIALS (CMAT) and two production foundries. The results of the casting trials show that high integrity engineering components can be produced successfully from both alloys if specific processing parameters are used. It was shown that a combination of melt processing and mold temperature is necessary for the elimination of hot tears in both alloys.« less

Influence of zirconium additions on nitinol shape memory phase stability, transformation temperatures, and thermo-mechanical properties

NASA Astrophysics Data System (ADS)

Kornegay, Suzanne M.

This research focuses on exploring the influence of Zr additions in Ni-rich Nitinol alloys on the phase stability, transformation temperatures, and thermo-mechanical behavior using various microanalysis techniques. The dissertation is divided into three major bodies of work: (1) The microstructural and thermo-mechanical characterization of a 50.3Ni-32.2Ti-17.5Zr (at.%) Zr alloy; (2) The characterization and mechanical behavior of 50.3Ni-48.7Ti-1Zr and 50.3Ni-48.7Ti-1Hf alloys to determine how dilute additions alter the phases, transformation temperatures, and thermo-mechanical properties; and (3) The microstructural evolution and transformation behavior comparison of microstructure and transformation temperature for 50.3Ni-(49.7-X)Ti-XZr alloys, where X is 1,7, or 17.5% Zr aged at either 400°C and 550°C. The major findings of this work include the following: (1) In the dilute limit of 1% Zr, at 400°C aging, a spherical precipitate, denoted as the S-phase, was observed. This is the first report of this phase. Further aging resulted in the secondary precipitation event of the H-phase. Increasing the aging temperature to 550°C, resulted in no evident precipitation of the S- and H-phase precipitates suggestive this temperature is above the solvus boundary for these compositions. (2) For the 7% and 17.5% Zr alloys, aging at 400°C and 550°C resulted in the precipitation of the H-phase. For the lower temperature anneal, this phase required annealing up to 300 hours of aging to be observed for the 17.5% Zr alloy. Upon increasing the aging temperature, the H-phase precipitation was present in both alloys. The transformation behavior and thermo-mechanical properties are linked to the precipitation behavior.

Role of Hf on Phase Formation in Ti45Zr(38-x)Hf(x)Ni17 Liquids and Solids

NASA Technical Reports Server (NTRS)

Wessels, V.; Sahu, K. K.; Gangopadhyay, A. K.; Huett, V. T.; Canepari, S.; Goldman, A. I.; Hyers, R. W.; Kramer, M. J.; Rogers, J. R.; Kelton, K. F.;

Development of Alternative Technetium Waste Forms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Czerwinski, Kenneth

2013-09-13

The UREX+1 process is under consideration for the separation of transuranic elements from spent nuclear fuel. The first steps of this process extract the fission product technicium-99 ({sup 99}Tc) into an organic phase containing tributylphosphate together with uranium. Treatment of this stream requires the separation of Tc from U and placement into a suitable waste storage form. A potential candidate waste form involves immobilizing the Tc as an alloy with either excess metallic zirconium or stainless steel. Although Tc-Zr alloys seem to be promising waste forms, alternative materials must be investigated. Innovative studies related to the synthesis and behavior ofmore » a different class of Tc materials will increase the scientific knowledge related to development of Tc waste forms. These studies will also provide a better understanding of the behavior of {sup 99}Tc in repository conditions. A literature survey has selected promising alternative waste forms for further study: technetium metallic alloys, nitrides, oxides, sulfides, and pertechnetate salts. The goals of this project are to 1) synthesize and structurally characterize relevant technetium materials that may be considered as waste forms, 2) investigate material behavior in solution under different conditions of temperature, electrochemical potential, and radiation, and 3) predict the long-term behavior of these materials.« less

Report on the Synchrotron Characterization of U-Mo and U-Zr Alloys and the Modeling Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Okuniewski, Maria A.; Ganapathy, Varsha; Hamilton, Brenden

2016-09-01

ABSTRACT Uranium-molybdenum (U-Mo) and uranium-zirconium (U-Zr) are two promising fuel candidates for nuclear transmutation reactors which burn long-lived minor actinides and fission products within fast spectrum reactors. The objectives of this research are centered on understanding the early stages of fuel performance through the examination of the irradiation induced microstructural changes in U-Zr and U-Mo alloys subjected to low neutron fluences. Specimens that were analyzed include those that were previously irradiated in the Advanced Test Reactor at INL. This most recent work has focused on a sub-set of the irradiated specimens, specifically U-Zr and U-Mo alloys that were irradiated tomore » 0.01 dpa at temperatures ranging from (150-800oC). These specimens were analyzed with two types of synchrotron techniques, including X-ray absorption fine structure and X-ray diffraction. These techniques provide non-destructive microstructural analysis, including phase identification and quantitation, lattice parameters, crystallite sizes, as well as bonding, structure, and chemistry. Preliminary research has shown changes in the phase fractions, crystallite sizes, and lattice parameters as a function of irradiation and temperature. Future data analyses will continue to explore these microstructural changes.« less

Irradiation-induced effects of proton irradiation on zirconium carbides with different stoichiometries

DOE Office of Scientific and Technical Information (OSTI.GOV)

Y. Huang; B.R. Maier; T.R. Allen

2014-10-01

Zirconium carbide (ZrC) is being considered for utilization in deep burn TRISO fuel particles for hightemperature, gas-cooled reactors. Zirconium carbide has a cubic B1 type crystal structure along with a very high melting point (3420 ?C), exceptional hardness and good thermal and electrical conductivities. Understanding the ZrC irradiation response is crucial for establishing ZrC as an alternative component in TRISO fuel. Until now, very few studies on irradiation effects on ZrC have been released and fundamental aspects of defect evolution and kinetics are not well understood although some atomistic simulations and phenomenological studies have been performed. This work was carriedmore » out to understand the damage evolution in float-zone refined ZrC with different stoichiometries. Proton irradiations at 800 ?C up to doses of 3 dpa were performed on ZrCx (where x ranges from 0.9 to 1.2) to investigate the damage evolution. The irradiation-induced defects, such as density of dislocation loops, at different stoichiometries and doses which were characterized by transmission electron microscopy (TEM) is presented and discussed.« less

Net Shaped Component Fabrication of Refractory Metal Alloys using Vacuum Plasma Spraying

NASA Technical Reports Server (NTRS)

Sen, S.; ODell, S.; Gorti, S.; Litchford, R.

2006-01-01

The vacuum plasma spraying (VPS) technique was employed to produce dense and net shaped components of a new tungsten-rhenium (W-Re) refractory metal alloy. The fine grain size obtained using this technique enhanced the mechanical properties of the alloy at elevated temperatures. The alloy development also included incorporation of thermodynamically stable dispersion phases to pin down grain boundaries at elevated temperatures and thereby circumventing the inherent problem of recrystallization of refractory alloys at elevated temperatures. Requirements for such alloys as related to high temperature space propulsion components will be discussed. Grain size distribution as a function of cooling rate and dispersion phase loading will be presented. Mechanical testing and grain growth results as a function of temperature will also be discussed.

Effects of Alloying Elements on Room and High Temperature Tensile Properties of Al-Si Cu-Mg Base Alloys =

NASA Astrophysics Data System (ADS)

Alyaldin, Loay

In recent years, aluminum and aluminum alloys have been widely used in automotive and aerospace industries. Among the most commonly used cast aluminum alloys are those belonging to the Al-Si system. Due to their mechanical properties, light weight, excellent castability and corrosion resistance, these alloys are primarily used in engineering and in automotive applications. The more aluminum is used in the production of a vehicle, the less the weight of the vehicle, and the less fuel it consumes, thereby reducing the amount of harmful emissions into the atmosphere. The principal alloying elements in Al-Si alloys, in addition to silicon, are magnesium and copper which, through the formation of Al2Cu and Mg2Si precipitates, improve the alloy strength via precipitation hardening following heat treatment. However, most Al-Si alloys are not suitable for high temperature applications because their tensile and fatigue strengths are not as high as desired in the temperature range 230-350°C, which are the temperatures that are often attained in automotive engine components under actual service conditions. The main challenge lies in the fact that the strength of heat-treatable cast aluminum alloys decreases at temperatures above 200°C. The strength of alloys under high temperature conditions is improved by obtaining a microstructure containing thermally stable and coarsening-resistant intermetallics, which may be achieved with the addition of Ni. Zr and Sc. Nickel leads to the formation of nickel aluminide Al3Ni and Al 9FeNi in the presence of iron, while zirconium forms Al3Zr. These intermetallics improve the high temperature strength of Al-Si alloys. Some interesting improvements have been achieved by modifying the composition of the base alloy with additions of Mn, resulting in an increase in strength and ductility at both room and high temperatures. Al-Si-Cu-Mg alloys such as the 354 (Al-9wt%Si-1.8wt%Cu-0.5wt%Mg) alloys show a greater response to heat treatment as a result of the presence of both Mg and Cu. These alloy types display excellent strength values at both low and high temperatures. Additions of Zr, Ni, Mn and Sc would be expected to maintain the performance of these alloys at still higher temperatures. Six alloys were prepared using 0.2 wt% Ti grain-refined 354 alloy, comprising alloy R (354 + 0.25wt% Zr) considered as the base or reference alloy, and five others, viz., alloys S, T, U, V, and Z containing various amounts of Ni, Mn, Sc and Zr, added individually or in combination. For comparison purposes, another alloy L was prepared from 398 (Al-16%Si) alloy, reported to give excellent high temperature properties, to which the same levels of Zr and Sc additions were made, as in alloy Z. Tensile test bars were prepared from the different 354 alloys using an ASTM B-108 permanent mold. The test bars were solution heat treated using a one-step or a multi-step solution heat treatment, followed by quenching in warm water, and then artificial aging employing different aging treatments (T5, T6, T62 and T7). The one-step (or SHT 1) solution treatment consisted of 5 h 495 °C) and the multi-step (or SHT 2) solution treatment comprised 5 h 495°C + 2 h 515°C + 2 h 530°C. Thermal analysis of the various 354 alloy melts was carried out to determine the sequence of reactions and phases formed during solidification under close-to-equilibrium cooling conditions. The main reactions observed comprised formation of the alpha-Al dendritic network at 598°C followed by precipitation of the Al-Si eutectic and post-eutectic beta-Al5FeSi phase at 560°C; Mg2Si phase and transformation of the beta-phase into pi-Al8Mg 3FeSi6 phase at 540°C and 525°C; and lastly, precipitation of Al2Cu and Q-Al5Mg8Cu2Si 6 almost simultaneously at 498°C and 488°C. Larger sizes of AlFeNi and AlCuNi phase particles were observed in T alloy with its higher Ni content of 4 wt%, when compared to those seen in S alloy at 2% Ni content. Mn addition in Alloy U helps in reducing the detrimental effect of the beta-iron phase by replacing it with the less-detrimental Chinese script alpha-Al 15(Fe,Mn)3Si2 phase and sludge particles.

Optical properties and emissivities of liquid metals and alloys

NASA Technical Reports Server (NTRS)

Krishnan, Shankar; Nordine, Paul C.

1993-01-01

This paper presents the results from our on-going program to investigate the optical properties of liquid metals and alloys at elevated temperatures. Ellipsometric and polarimetric techniques have been used to investigate the optical properties of materials in the 1000 - 3000 K temperature range and in the 0.3 - 0.1 mu m wavelength range. The ellipsometric and polarimetric techniques are described and the characteristics of the instruments are presented. The measurements are conducted by reflecting a polarized laser beam from an electromagnetically levitated liquid metal or alloy specimen. A Rotating Analyzer Ellipsometer (RAE) or a four-detector Division-of-Amplitude Photopolarimeter (DOAP) is used to determine the polarimetric properties of the light reflected at an angle of incidence of approximately 68 deg. Optical properties of the specimen which are calculated from these measurements include the index of refraction, extinction coefficient, normal spectral emissivity, and spectral hemispherical emissivity. These properties have been determined at various wavelengths and temperatures for liquid Ag, Al, Au, Cu, Nb, Ni, Pd, Pt, Si, Ti, Ti-Al alloys, U, and Zr. We also describe new experiments using pulsed-dye laser spectroscopic ellipsometry for studies of the wavelength dependence of the emissivities and optical properties of materials at high temperature. Preliminary results are given for liquid Al. The application of four-detector polarimetry for rapid determination of surface emissivity and true temperature is also described. Characteristics of these devices are presented. An example of the accuracy of this instrument in measurements of the melting point of zirconium is illustrated.

Electrical insulator assembly with oxygen permeation barrier

DOEpatents

Van Der Beck, R.R.; Bond, J.A.

1994-03-29

A high-voltage electrical insulator for electrically insulating a thermoelectric module in a spacecraft from a niobium-1% zirconium alloy wall of a heat exchanger filled with liquid lithium while providing good thermal conductivity between the heat exchanger and the thermoelectric module. The insulator has a single crystal alumina layer (SxAl[sub 2]O[sub 3], sapphire) with a niobium foil layer bonded thereto on the surface of the alumina crystal facing the heat exchanger wall, and a molybdenum layer bonded to the niobium layer to act as an oxygen permeation barrier to preclude the oxygen depleting effects of the lithium from causing undesirable niobium-aluminum intermetallic layers near the alumina-niobium interface. 3 figures.

Parametric and experimentally informed BWR Severe Accident Analysis Utilizing FeCrAl - M3FT-17OR020205041

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ott, Larry J.; Howell, Michael; Robb, Kevin R.

Iron-chromium-aluminum (FeCrAl) alloys are being considered as advanced fuel cladding concepts with enhanced accident tolerance. At high temperatures, FeCrAl alloys have slower oxidation kinetics and higher strength compared with zirconium-based alloys. FeCrAl could be used for fuel cladding and spacer or mixing vane grids in light water reactors and/or as channel box material in boiling water reactors (BWRs). There is a need to assess the potential gains afforded by the FeCrAl accident-tolerant-fuel (ATF) concept over the existing zirconium-based materials employed today. To accurately assess the response of FeCrAl alloys under severe accident conditions, a number of FeCrAl properties and characteristicsmore » are required. These include thermophysical properties as well as burst characteristics, oxidation kinetics, possible eutectic interactions, and failure temperatures. These properties can vary among different FeCrAl alloys. Oak Ridge National Laboratory has pursued refined values for the oxidation kinetics of the B136Y FeCrAl alloy (Fe-13Cr-6Al wt %). This investigation included oxidation tests with varying heating rates and end-point temperatures in a steam environment. The rate constant for the low-temperature oxidation kinetics was found to be higher than that for the commercial APMT FeCrAl alloy (Fe-21Cr-5Al-3Mo wt %). Compared with APMT, a 5 times higher rate constant best predicted the entire dataset (root mean square deviation). Based on tests following heating rates comparable with those the cladding would experience during a station blackout, the transition to higher oxidation kinetics occurs at approximately 1,500°C. A parametric study varying the low-temperature FeCrAl oxidation kinetics was conducted for a BWR plant using FeCrAl fuel cladding and channel boxes using the MELCOR code. A range of station blackout severe accident scenarios were simulated for a BWR/4 reactor with Mark I containment. Increasing the FeCrAl low-temperature oxidation rate constant (3 times and 10 times that of the rate constant for APMT) had a negligible impact on the early stages of the accident and minor impacts on the accident progression after the first relocation of the fuel. At temperatures below 1,500°C, increasing the rate constant for APMT by a factor of 10 still resulted in only minor FeCrAl oxidation. In general, the gains afforded by the FeCrAl enhanced ATF concept with respect to accident sequence timing and combustible gas generation are consistent with previous efforts. Compared with the traditional Zircaloy-based cladding and channel box system, the FeCrAl concept could provide a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. For example, a station blackout was simulated in which cooling water injection was lost 36 hours after shutdown. The timing to first fuel relocation was delayed by approximately 5 h for the FeCrAl ATF concept compared with that of the traditional Zircaloy-based cladding and channel box system.« less

The adhesion performance of epoxy coating on AA6063 treated in Ti/Zr/V based solution

NASA Astrophysics Data System (ADS)

Zhu, Wen; Li, Wenfang; Mu, Songlin; Yang, Yunyu; Zuo, Xi

2016-10-01

An environment-friendly titanium/zirconium/vanadium-based (Ti/Zr/V) conversion coating was prepared on aluminum alloy 6063 (AA6063). The epoxy powder coatings were applied on the AA6063 samples with/without Ti/Zr/V conversion coatings via electrostatic spraying. The morphology and composition of the conversion coating were studied by scanning electron microscope (SEM) and X-ray photoelectron spectroscopy (XPS), respectively. The surface free energy components of AA6063 samples were measured by a static contact angle measuring device with Owens method. The adhesion properties of the epoxy coating on AA6063 treated with different conversion times were evaluated using a pull-off tester. The Ti/Zr/V conversion coating was mainly composed of metal oxide (TiO2, ZrO2, V2O5, Al2O3, etc.), metal fluoride (ZrF4, AlF3, etc.) and metal organic complex. The formation time of this conversion coating was reduced to 50 s. After such surface treatment, the samples' surface roughness was increased and the contact angle with water was decreased. Both the surface free energy and the work of adhesion were increased. The adhesion strength between the epoxy coating and AA6063 was enhanced significantly.

Atomic mixing induced by swift heavy ion irradiation of Fe/Zr multilayers

NASA Astrophysics Data System (ADS)

Jaouen, C.; Michel, A.; Pacaud, J.; Dufour, C.; Bauer, Ph.; Gervais, B.

1999-01-01

The mechanism of ion induced mixing and phase change was studied for Fe/Zr multilayers, and specifically for the case of swift heavy ions giving rise to a very large electronic excitation of the target. The multilayers had a modulation of 7.6 nm and an overall composition Fe 69Zr 31. The Zr layers were amorphous whereas the Fe ones were crystalline (bcc) with a very strong (1 1 0) texture in the growth direction. The phase transformation and the composition changes were analysed using the structural and magnetic properties of the Fe component by means of a detailed analysis of the X-ray diffraction profiles and with the aid of backscattering Mössbauer spectroscopy. A complete mixing was observed at a fluence of 10 13 U/cm 2. Both phenomena, the dose dependence of the ion beam mixed amorphous non-magnetic phase and the quantitative evolution of the crystalline iron layer thickness, suggest that mixing occurs in a two-stage process. At an initial stage, an anisotropic diffusion of iron atoms in the amorphous zirconium layers takes place along the interface, while subsequent ion bombardment leads to a generalised transformation through the whole of the Fe layer. Finally, the implications of these observations are discussed in comparison to the plastic deformation phenomena reported for amorphous alloys.

Fusion boundary microstructure evolution in aluminum alloys

NASA Astrophysics Data System (ADS)

Kostrivas, Anastasios Dimitrios

2000-10-01

A melting technique was developed to simulate the fusion boundary of aluminum alloys using the GleebleRTM thermal simulator. Using a steel sleeve to contain the aluminum, samples were heated to incremental temperatures above the solidus temperature of a number of alloys. In alloy 2195, a 4wt%Cu-1wt%Li alloy, an equiaxed non-dendritic zone (EQZ) could be formed by heating in the temperature range from approximately 630 to 640°C. At temperatures above 640°C, solidification occurred by the normal epitaxial nucleation and growth mechanism. Fusion boundary behavior was also studied in alloys 5454-H34, 6061-T6, and 2219-T8. Additionally, experimental alloy compositions were produced by making bead on plate welds using an alloy 5454-H32 base metal and 5025 or 5087 filler metals. These filler metals contain zirconium and scandium additions, respectively, and were expected to influence nucleation and growth behavior. Both as-welded and welded/heat treated (540°C and 300°C) substrates were tested by melting simulation, resulting in dendritic and EQZ structures depending on composition and substrate condition. Orientation imaging microscopy (OIM(TM)) was employed to study the crystallographic character of the microstructures produced and to verify the mechanism responsible for EQZ formation. OIM(TM) proved that grains within the EQZ have random orientation. In all other cases, where the simulated microstructures were dendritic in nature, it was shown that epitaxy was the dominant mode of nucleation. The lack of any preferred crystallographic orientation relationship in the EQZ supports a theory proposed by Lippold et al that the EQZ is the result of heterogeneous nucleation within the weld unmixed zone. EDS analysis of the 2195 on STEM revealed particles with ternary composition consisted of Zr, Cu and Al and a tetragonal type crystallographic lattice. Microdiffraction line scans on EQZ grains in the alloy 2195 showed very good agreement between the measured Cu composition within the interior of the non-dendritic grains and the corresponding value the Scheil equation predicts for the first solid to form upon solidification for a binary Al-Cu alloy with identical Cu composition. In the context of the alloys, compositions and substrate conditions examined a mechanistic model for EQZ zone formation is proposed, helpful in adjusting base metal compositions and/or substrate conditions to control fusion boundary microstructure.

Long-term strategies for increased recycling of automotive aluminum and its alloying elements.

PubMed

Løvik, Amund N; Modaresi, Roja; Müller, Daniel B

2014-04-15

Aluminum recycling currently occurs in a cascading fashion, where some alloys, used in a limited number of applications, absorb most of the end-of-life scrap. An expected increase in scrap supply in coming decades necessitates restructuring of the aluminum cycle to open up new recycling paths for alloys and avoid a potential scrap surplus. This paper explores various interventions in end-of-life management and recycling of automotive aluminum, using a dynamic substance flow analysis model of aluminum and its alloying elements with resolution on component and alloy level (vehicle-component-alloy-element model). It was found that increased component dismantling before vehicle shredding can be an effective, so far underestimated, intervention in the medium term, especially if combined with development of safety-relevant components such as wheels from secondary material. In the long term, automatic alloy sorting technologies are most likely required, but could at the same time reduce the need for magnesium removal in refining. Cooperation between the primary and secondary aluminum industries, the automotive industry, and end-of-life vehicle dismantlers is therefore essential to ensure continued recycling of automotive aluminum and its alloying elements.

Biomechanical comparison of the strength of adhesion of polymethylmethacrylate cement to zirconia ceramic and cobalt-chromium alloy components in a total knee arthroplasty.

PubMed

Kumahashi, Nobuyuki; Uchio, Yuji; Kitamura, Nobuto; Satake, Shigeru; Iwamoto, Mikio; Yasuda, Kazunori

2014-11-01

The purpose of this study was to clarify the biomechanical characteristics of cement-material interfaces for the zirconia ceramic and cobalt-chromium (Co-Cr) alloy femoral components used for total knee arthroplasty. In the first sub-study, we compared the strength of adhesion of the cement to flat plates, by tensile testing under dry and moistened conditions. In the second sub-study, we compared the maximum load of the cement-component complex by tensile testing. In the third sub-study, we compared the fatigue characteristics of the cement-component complex by use of a dynamic tensile testing machine. Under dry conditions, the maximum strength of adhesion to the zirconia ceramic plate was the same as that to the Co-Cr alloy plate. Under moistened conditions, however, the strength of adhesion to the zirconia ceramic plate was significantly lower (p = 0.0017) whereas the strength of adhesion to the Co-Cr alloy plate was not reduced. Maximum load for the cement-component complexes for zirconia ceramic and Co-Cr alloy was no different under both dry and moistened conditions. Fatigue testing showed that cement-zirconia adhesion was stronger than cement-Co-Cr alloy adhesion (p = 0.0161). The strength of adhesion of cement to zirconia ceramic is substantially weaker under wet conditions than under dry conditions. The mechanical properties of cement-zirconia ceramic component complexes and cement-Co-Cr alloy component complexes are equivalent.

SU-F-T-426: Measurement of Dose Enhancement Due to Backscatter From Modern Dental Materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hurwitz, M; Margalit, D; Williams, C

Purpose: High-density materials used in dental restoration can cause significant localized dose enhancement due to electron backscatter in head-and-neck radiotherapy, increasing the risk of mucositis. The materials used in prosthetic dentistry have evolved in the last decades from metal alloys to ceramics. We aim to determine the dose enhancement caused by backscatter from currently-used dental materials. Methods: Measurements were performed for three different dental materials: lithium disilicate (Li{sub 2}Si{sub 2}O{sub 5}), zirconium dioxide (ZrO{sub 2}), and gold alloy. Small thin squares (2×2×0.15 cm{sup 3}) of the material were fabricated, and placed into a phantom composed of tissue-equivalent material. The phantommore » was irradiated with a single 6 MV photon field. A thin-window parallel-plate ion chamber was used to measure the dose at varying distances from the proximal interface between the material and the plastic. Results: The dose enhancement at the interface between the high-density and tissue-equivalent materials, relative to a homogeneous phantom, was 54% for the gold alloy, 31% for ZrO{sub 2}, and 9% for Li{sub 2}Si{sub 2}O{sub 5}. This enhancement decreased rapidly with distance from the interface, falling to 11%, 5%, and 0.5%, respectively, 2 mm from the interface. Comparisons with the modeling of this effect in treatment planning systems are performed. Conclusion: While dose enhancement due to dental restoration is smaller with ceramic materials than with metal alloys, it can still be significant. A spacer of about 2–3 mm would be effective in reducing this enhancement, even for metal alloys.« less

Aluminum-Silicon Alloy Having Improved Properties At Elevated Temperatures and Process for Producing Cast Articles Therefrom

NASA Technical Reports Server (NTRS)

Lee, Jonathan A. (Inventor); Chen, Po-Shou (Inventor)

2002-01-01

A process for making a cast article from an aluminum alloy includes first casting an article from an alloy having the following composition, in weight percent: Silicon 11.0-14.0, Copper 5.6-8.0, Iron 0-0.8, Magnesium 0.5-1.5, Nickel 0.05-0.9, Manganese 0-1.0, Titanium 0.05-1.2, Zirconium 0.12-1.2, Vanadium 0.05-1.2, Zinc 0.05-0.9, Strontium 0.001-0.1, Aluminum balance . In this alloy the ratio of silicon to magnesium is 10 to 25, and the ratio of copper to magnesium is 4 to 15. After an article is cast from the alloy, the cast article is aged at a temperature within the range of 400F to 500F for a time period within the range of four to 16 hours. It has been found especially advantageous if the cast article is first exposed to a solutionizing step prior to the aging step. This solutionizing step is carried out by exposing the cast article to a temperature within the range of 900F to 1000F for a time period of fifteen minutes to four hours. It has also been found to be especially advantageous if the solutionizing step is followed directly with a quenching step, wherein the cast article is quenched in a quenching medium such as water at a temperature within the range of 120F to 300F. The resulting cast article is suitable in a number of high temperature applications, such as heavy-duty pistons for internal combustion engines.

Process for Producing a Cast Article from a Hypereutectic Aluminum-Silicon Alloy

NASA Technical Reports Server (NTRS)

Lee, Jonathan A. (Inventor); Chen, Po-Shou (Inventor)

2003-01-01

A process for making a cast article from an aluminum alloy includes first casting an article from an alloy having the following composition, in weight percent: Silicon (Si) 14.0-25.0, Copper (CU) 5.5-8.0, Iron (Fe) 0-0.8, Magnesium (Mg) 0.5-1.5, Nickel (Ni) 0.05-1.2, Manganese (Mn) 0-1.0, Titanium (Ti) 0.05-1.2, Zirconium (Zr) 0.12-1.2, Vanadium (V) 0.05-1.2, Zinc (Zn) 0-0.9, Phosphorus (P) 0.001-0.1, Aluminum, balance. In this alloy the ration of Si:Mg is 15-35, and the ratio of Cu:Mg is 4-15. After an article is cast from the alloy, the cast article is aged at a temperature within the range of 400 F to 500 F for a time period within the range of four to 16 hours. It has been found especially advantageous if the cast article is first exposed to a solutionizing step prior to the aging step. This solutionizing step is carried out by exposing the cast article to a temperature within the range of 875 F to 1025 F for a time period of fifteen minutes to four hours. It has also been found to be especially advantageous if the solutionizing step is followed directly with a quenching step, wherein the cast article is quenched in a quenching medium such as water at a temperature within the range of 120 F to 300 F. The resulting cast article is highly suitable in a number of high temperature applications, such as heavy-duty pistons for internal combustion engines.

An Impact of Zirconium Doping of Zn-Al Braze on the Aluminum-Stainless Steel Joints Integrity During Aging

NASA Astrophysics Data System (ADS)

Yang, Jinlong; Xue, Songbai; Sekulic, Dusan P.

2017-01-01

This work offers an analysis of the microstructure and the growth rate of an intermetallic compound within the aged AA 6061 aluminum alloy-304 stainless steel joint brazed with Zn-15Al and Zn-15Al-0.2Zr filler metals. The effect of zirconium addition on mechanical integrity of the brazed joint was studied. The experimental results confirm that the thickness of the Fe-Al intermetallic layer formed at the brazed seam/stainless steel interface increases with the increase of the aging time. Furthermore, it is established that the growth rate of the intermetallic layer for the Zn-15Al-0.2Zr brazed joint was lower than that for Zn-15Al. The results also indicate that the shear strength of both Zn-15Al and Zn-15Al-0.2Zr brazed joints decreases monotonously during aging. The value of the strength after aging lasting for 800 h for Zn-15Al and Zn-15Al-0.2Zr has decreased by 20 and 17%, respectively. The fracture of joints occurred at the interface between the brazed seam and the Fe4Al13 intermetallic layer. The morphology of the surfaces exhibits a cleavage fracture.

Nanomechanical Characterization of Temperature-Dependent Mechanical Properties of Ion-Irradiated Zirconium with Consideration of Microstructure and Surface Damage

NASA Astrophysics Data System (ADS)

Marsh, Jonathan; Zhang, Yang; Verma, Devendra; Biswas, Sudipta; Haque, Aman; Tomar, Vikas

2015-12-01

Zirconium alloys for nuclear applications with different microstructures were produced by manufacturing processes such as chipping, rolling and annealing. The two Zr samples, rolled and rolled-annealed were subjected to different levels of irradiation, 1 keV and 100 eV, to study the effect of irradiation dosages. The effect of microstructure and irradiation on the mechanical properties (reduced modulus, hardness, indentation yield strength) was analyzed with nanoindentation experiments, which were carried out in the temperature range of 25°C to 450°C to investigate temperature dependence. An indentation size effect analysis was performed and the mechanical properties were also corrected for the oxidation effects at high temperatures. The irradiation-induced hardness was observed, with rolled samples exhibiting higher increase compared to rolled and annealed samples. The relevant material parameters of the Anand viscoplastic model were determined for Zr samples containing different level of irradiation to account for viscoplasticity at high temperatures. The effect of the microstructure and irradiation on the stress-strain curve along with the influence of temperature on the mechanisms of irradiation creep such as formation of vacancies and interstitials is presented. The yield strength of irradiated samples was found to be higher than the unirradiated samples which also showed a decreasing trend with the temperature.

MEMS-Based Waste Vibrational Energy Harvesters

DTIC Science & Technology

2013-06-01

7 1. Lead Zirconium Titanate ( PZT ) .........................................................7 2. Aluminum...Laboratory PiezoMUMPS Piezoelectric Multi-User MEMS Processes PZT Lead Zirconate Titanate SEM Scanning Electron Microscopy SiO2 Silicon...titanate ( PZT ) possess high 4 coupling between the electrical and mechanical domains [11]. The output voltage, V, is related to the z-component

Thermal barrier coating

DOEpatents

Bowker, Jeffrey Charles; Sabol, Stephen M.; Goedjen, John G.

2001-01-01

A thermal barrier coating for hot gas path components of a combustion turbine based on a zirconia-scandia system. A layer of zirconium scandate having the hexagonal Zr.sub.3 Sc.sub.4 O.sub.12 structure is formed directly on a superalloy substrate or on a bond coat formed on the substrate.

Modification in band gap of zirconium complexes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sharma, Mayank, E-mail: mayank30134@gmail.com; Singh, J.; Chouhan, S.

2016-05-06

The optical properties of zirconium complexes with amino acid based Schiff bases are reported here. The zirconium complexes show interesting stereo chemical features, which are applicable in organometallic and organic synthesis as well as in catalysis. The band gaps of both Schiff bases and zirconium complexes were obtained by UV-Visible spectroscopy. It was found that the band gap of zirconium complexes has been modified after adding zirconium compound to the Schiff bases.

Properties of zirconium silicate and zirconium-silicon oxynitride high-k dielectric alloys for advanced microelectronic applications: Chemical and electrical characterizations

NASA Astrophysics Data System (ADS)

Ju, Byongsun

2005-11-01

As the microelectronic devices are aggressively scaled down to the 1999 International Technology Roadmap, the advanced complementary metal oxide semiconductor (CMOS) is required to increase packing density of ultra-large scale integrated circuits (ULSI). High-k alternative dielectrics can provide the required levels of EOT for device scaling at larger physical thickness, thereby providing a materials pathway for reducing the tunneling current. Zr silicates and its end members (SiO2 and ZrO2) and Zr-Si oxynitride films, (ZrO2)x(Si3N 4)y(SiO2)z, have been deposited using a remote plasma-enhanced chemical vapor deposition (RPECVD) system. After deposition of Zr silicate, the films were exposed to He/N2 plasma to incorporate nitrogen atoms into the surface of films. The amount of incorporated nitrogen atoms was measured by on-line Auger electron spectrometry (AES) as a function of silicate composition and showed its local minimum around the 30% silicate. The effect of nitrogen atoms on capacitance-voltage (C-V) and leakage-voltage (J-V) were also investigated by fabricating metal-oxide-semiconductor (MOS) capacitors. Results suggested that incorporating nitrogen into silicate decreased the leakage current in SiO2-rich silicate, whereas the leakage increased in the middle range of silicate. Zr-Si oxynitride was a pseudo-ternary alloy and no phase separation was detected by x-ray photoelectron spectroscopy (XPS) analysis up to 1100°C annealing. The leakage current of Zr-Si oxynitride films showed two different temperature dependent activation energies, 0.02 eV for low temperature and 0.3 eV for high temperature. Poole-Frenkel emission was the dominant leakage mechanism. Zr silicate alloys with no Si3N4 phase were chemically separated into the SiO2 and ZrO2 phase as annealed above 900°C. While chemical phase separation in Zr silicate films with Si 3N4 phase (Zr-Si oxynitride) were suppressed as increasing the amount of Si3N4 phase due to the narrow bonding network m Si3N4 phase. (3.4 bonds/atom for Si3 N4 network, 2.67 bonds/atom for SiO2 network).

Physical and Mechanical Metallurgy of Zirconium Alloys for Nuclear Applications: A Multi-Scale Computational Study

NASA Astrophysics Data System (ADS)

Glazoff, Michael Vasily

In the post-Fukushima world, thermal and structural stability of materials under extreme conditions is an important issue for the safety of nuclear reactors. Because the nuclear industry will continue using zirconium (Zr) cladding for the foreseeable future, it becomes critical to gain a fundamental understanding of several interconnected problems. First, what are the thermodynamic and kinetic factors affecting oxidation and hydrogen pick-up by these materials at normal, off-normal conditions, and in long-term storage? Secondly, what protective coatings could be used in order to gain valuable time at off-normal conditions (temperature exceeds ~1200°C (2200°F)? Thirdly, the kinetics of the coating's oxidation must be understood. Lastly, one needs automated inspection algorithms allowing identifying cladding's defects. This work attempts to explore the problem from a computational perspective, utilizing first principles atomistic simulations, computational thermodynamics, plasticity theory, and morphological algorithms of image processing for defect identification. It consists of the four parts dealing with these four problem areas preceded by the introduction. In the 1st part, computational thermodynamics and ab initio calculations were used to shed light upon the different stages of zircaloy oxidation and hydrogen pickup, and microstructure optimization to increase thermal stability. The 2 nd part describes the kinetic theory of oxidation of the several materials considered to be perspective coatings for Zr alloys: SiC and ZrSiO4. The 3rd part deals with understanding the respective roles of the two different plasticity mechanisms in Zr nuclear alloys: twinning (at low T) and crystallographic slip (higher T's). For that goal, an advanced plasticity model was proposed. In the 4th part projectional algorithms for defect identification in zircaloy coatings are described. Conclusions and recommendations are presented in the 5th part. This integrative approach's value is in developing multi-faceted understanding of complex processes taking place in nuclear fuel rods. It helped identify several problems pertaining to the safe operations with nuclear fuel: limits of temperature that should be strictly obeyed in storage to retard zircaloy hydriding; understanding the benefits and limitations of coatings; developing in-depth understanding of Zr plasticity; developing original algorithms for defect identification in SiC-braided zircaloy. The obtained results will be useful for the nuclear industry.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wood, Elizabeth Sooby; Parker, Stephen Scott; Nelson, Andrew Thomas

The Fuel Cycle Research and Development program’s Advanced Fuels Campaign is currently supporting a range of experimental efforts aimed at the development and qualification of ‘accident tolerant’ nuclear fuel forms. One route to enhance the accident tolerance of nuclear fuel is to replace the zirconium alloy cladding, which is prone to rapid oxidation in steam at elevated temperatures, with a more oxidation-resistant cladding. Several cladding replacement solutions have been envisaged. The cladding can be completely replaced with a more oxidation resistant alloy, a layered approach can be used to optimize the strength, creep resistance, and oxidation tolerance of various materials,more » or the existing zirconium alloy cladding can be coated with a more oxidation-resistant material. Molybdenum is one candidate cladding material favored due to its high temperature creep resistance. However, it performs poorly under autoclave testing and suffers degradation under high temperature steam oxidation exposure. Development of composite cladding architectures consisting of a molybdenum core shielded by a molybdenum disilicide (MoSi 2) coating is hypothesized to improve the performance of a Mo-based cladding system. MoSi 2 was identified based on its high temperature oxidation resistance in O 2 atmospheres (e.g. air and “wet air”). However, its behavior in H 2O is less known. This report presents thermogravimetric analysis (TGA), scanning electron microscopy (SEM), and x-ray diffraction (XRD) results for MoSi 2 exposed to 670-1498 K water vapor. Synthetic air (80-20%, Ar-O 2) exposures were also performed, and those results are presented here for a comparative analysis. It was determined that MoSi 2 displays drastically different oxidation behavior in water vapor than in dry air. In the 670-1498 K temperature range, four distinct behaviors are observed. Parabolic oxidation is exhibited in only 670-773 K water vapor, a temperature range in which the material pests in dry O 2 environments. From 877-1084 K in water vapor, MoSi 2 undergoes rapid mass gain resulting in oxidation throughout the bulk of the sample at 980 K and 1084 K. The resulting material displays swelling and warping after the 980-1084 K exposures. A pre-passivation heat treatment performed at 1395 K was found capable of producing a coarse SiO 2 layer that limited pesting at lower temperatures in water vapor over the time periods investigated.« less

Cold worked ferritic alloys and components

DOEpatents

Korenko, Michael K.

1984-01-01

This invention relates to liquid metal fast breeder reactor and steam generator precipitation hardening fully ferritic alloy components which have a microstructure substantially free of the primary precipitation hardening phase while having cells or arrays of dislocations of varying population densities. It also relates to the process by which these components are produced, which entails solution treating the alloy followed by a final cold working step. In this condition, the first significant precipitation hardening of the component occurs during high temperature use.

Physicochemical analysis of urinary stones from Dharmapuri district

NASA Astrophysics Data System (ADS)

Aslin Shamema, A.; Thanigai Arul, K.; Senthil Kumar, R.; Narayana Kalkura, S.

2015-01-01

Nephrolithiasis is a common disease caused by the multifactorial components such as geographical location, bacterial infection, low urine volume, and low intake of water. This disease induces severe metabolic abnormalities in the human body. As the prevalence of this disease was high in Dharmapuri district located in Tamil Nadu, urinary stones removed from the patients pertaining to this district were collected and to identify the toxic elements present in the stones. The presence of functional groups and phases of the stones were analyzed using X-ray diffraction (XRD), Fourier transform Raman spectroscopy and Fourier transform infrared spectroscopy (FT-IR). The majority of stones were found to be calcium oxalate monohydrate (COM) and mixed stones having minor existence of struvite and uric acid. Hexagonal shaped COM crystals, needle shaped uric acid crystals and layered arrangement of struvite crystals in the core region were revealed by Scanning Electron Microscopy (SEM). Thermo Gravimetric Analysis (TGA) was used to determine the thermal stability and the hardness of the stone which was measured using Vickers hardness (HV). The presence of toxic elements in stones such as zirconium and mercury was identified using Energy Dispersive X-ray Spectroscopy (EDS). The EDS analysis showed higher concentration of zirconium in the core region compared to the periphery. The percentage of zirconium was relatively high compared to other toxic elements in the stones. The Vickers hardness results indicated that high HV values in the core region than the periphery and this might be due to the presence of zirconium.

[Research on fast classification based on LIBS technology and principle component analyses].

PubMed

Yu, Qi; Ma, Xiao-Hong; Wang, Rui; Zhao, Hua-Feng

2014-11-01

Laser-induced breakdown spectroscopy (LIBS) and the principle component analysis (PCA) were combined to study aluminum alloy classification in the present article. Classification experiments were done on thirteen different kinds of standard samples of aluminum alloy which belong to 4 different types, and the results suggested that the LIBS-PCA method can be used to aluminum alloy fast classification. PCA was used to analyze the spectrum data from LIBS experiments, three principle components were figured out that contribute the most, the principle component scores of the spectrums were calculated, and the scores of the spectrums data in three-dimensional coordinates were plotted. It was found that the spectrum sample points show clear convergence phenomenon according to the type of aluminum alloy they belong to. This result ensured the three principle components and the preliminary aluminum alloy type zoning. In order to verify its accuracy, 20 different aluminum alloy samples were used to do the same experiments to verify the aluminum alloy type zoning. The experimental result showed that the spectrum sample points all located in their corresponding area of the aluminum alloy type, and this proved the correctness of the earlier aluminum alloy standard sample type zoning method. Based on this, the identification of unknown type of aluminum alloy can be done. All the experimental results showed that the accuracy of principle component analyses method based on laser-induced breakdown spectroscopy is more than 97.14%, and it can classify the different type effectively. Compared to commonly used chemical methods, laser-induced breakdown spectroscopy can do the detection of the sample in situ and fast with little sample preparation, therefore, using the method of the combination of LIBS and PCA in the areas such as quality testing and on-line industrial controlling can save a lot of time and cost, and improve the efficiency of detection greatly.

Nanocrystalline diamond protects Zr cladding surface against oxygen and hydrogen uptake: Nuclear fuel durability enhancement.

PubMed

Škarohlíd, Jan; Ashcheulov, Petr; Škoda, Radek; Taylor, Andrew; Čtvrtlík, Radim; Tomáštík, Jan; Fendrych, František; Kopeček, Jaromír; Cháb, Vladimír; Cichoň, Stanislav; Sajdl, Petr; Macák, Jan; Xu, Peng; Partezana, Jonna M; Lorinčík, Jan; Prehradná, Jana; Steinbrück, Martin; Kratochvílová, Irena

2017-07-25

In this work, we demonstrate and describe an effective method of protecting zirconium fuel cladding against oxygen and hydrogen uptake at both accident and working temperatures in water-cooled nuclear reactor environments. Zr alloy samples were coated with nanocrystalline diamond (NCD) layers of different thicknesses, grown in a microwave plasma chemical vapor deposition apparatus. In addition to showing that such an NCD layer prevents the Zr alloy from directly interacting with water, we show that carbon released from the NCD film enters the underlying Zr material and changes its properties, such that uptake of oxygen and hydrogen is significantly decreased. After 100-170 days of exposure to hot water at 360 °C, the oxidation of the NCD-coated Zr plates was typically decreased by 40%. Protective NCD layers may prolong the lifetime of nuclear cladding and consequently enhance nuclear fuel burnup. NCD may also serve as a passive element for nuclear safety. NCD-coated ZIRLO claddings have been selected as a candidate for Accident Tolerant Fuel in commercially operated reactors in 2020.

SPITFIRE-1

NASA Technical Reports Server (NTRS)

Ryntz, Edward F.

1995-01-01

The purpose of this effort is to develop low-cost rapid forming superplastic aluminum that will be evaluated in pilot production trials for automotive SPF components. The alloy development study conducted under SPITFIRE-1 showed that the addition of CU or Mn to the base 5083 aluminum alloy refined the grain size, leading to enhanced superplastic properties. In SPITFIRE-2, these alloy variants will be further refined and studied to meet the target properties established earlier in the program. Mechanical properties, component forming and post-forming properties will be evaluated. Also, the alloy production process, including thermomechanical processing (TMP) optimization to reduce production cost, will be investigated during SPITFIRE-2. After identifying preferred compositions and production processing, the most promising alloy will be manufactured into production coils for verification during SPITFIRE-3. Components will be produced from these coils in SPITFIRE-4, and the process and component performance will be assessed.

Ceramic Coatings for Clad (The C 3 Project): Advanced Accident-Tolerant Ceramic Coatings for Zr-Alloy Cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sickafus, Kurt E.; Wirth, Brian; Miller, Larry

The goal of this NEUP-IRP project is to develop a fuel concept based on an advanced ceramic coating for Zr-alloy cladding. The coated cladding must exhibit demonstrably improved performance compared to conventional Zr-alloy clad in the following respects: During normal service, the ceramic coating should decrease cladding oxidation and hydrogen pickup (the latter leads to hydriding and embrittlement). During a reactor transient (e.g., a loss of coolant accident), the ceramic coating must minimize or at least significantly delay oxidation of the Zr-alloy cladding, thus reducing the amount of hydrogen generated and the oxygen ingress into the cladding. The specific objectivesmore » of this project are as follows: To produce durable ceramic coatings on Zr-alloy clad using two possible routes: (i) MAX phase ceramic coatings or similar nitride or carbide coatings; and (ii) graded interface architecture (multilayer) ceramic coatings, using, for instance, an oxide such as yttria-stabilized zirconia (YSZ) as the outer protective layer. To characterize the structural and physical properties of the coated clad samples produced in 1. above, especially the corrosion properties under simulated normal and transient reactor operating conditions. To perform computational analyses to assess the effects of such coatings on fuel performance and reactor neutronics, and to perform fuel cycle analyses to assess the economic viability of modifying conventional Zr-alloy cladding with ceramic coatings. This project meets a number of the goals outlined in the NEUP-IRP call for proposals, including: Improve the fuel/cladding system through innovative designs (e.g. coatings/liners for zirconium-based cladding) Reduce or eliminate hydrogen generation Increase resistance to bulk steam oxidation Achievement of our goals and objectives, as defined above, will lead to safer light-water reactor (LWR) nuclear fuel assemblies, due to improved cladding properties and built-in accident resistance, as well as the possibilities for enhanced fuel/clad system performance and longevity.« less

SEPARATION PROCESS FOR ZIRCONIUM AND COMPOUNDS THEREOF

DOEpatents

Crandall, H.W.; Thomas, J.R.

1959-06-30

The separation of zirconium from columbium, rare earths, yttrium and the alkaline earth metals, such mixtures of elements occurring in zirconium ores or neutron irradiated uranium is described. According to the invention a suitable separation of zirconium from a one normal acidic aqueous solution containing salts, nitrates for example, of tetravalent zirconium, pentavalent columbium, yttrium, rare earths in the trivalent state and alkaline earths can be obtained by contacting the aqueous solution with a fluorinated beta diketonc alone or in an organic solvent solution, such as benzene, to form a zirconium chelate compound. When the organic solvent is present the zirconium chelate compound is directly extracted; otherwise it is separated by filtration. The zirconium may be recovered from contacting the organic solvent solution containing the chelated compound by back extraction with either an aqueous hydrofluoric acid or an oxalic acid solution.

Advanced oxide dispersion strengthened sheet alloys for improved combustor durability

NASA Technical Reports Server (NTRS)

Henricks, R. J.

1981-01-01

Burner design modifications that will take advantage of the improved creep and cyclic oxidation resistance of oxide dispersion strengthened (ODS) alloys while accommodating the reduced fatigue properties of these materials were evaluated based on preliminary analysis and life predictions, on construction and repair feasibility, and on maintenance and direct operating costs. Two designs - the film cooled, segmented louver and the transpiration cooled, segmented twin Wall - were selected for low cycle fatigue (LCF) component testing. Detailed thermal and structural analysis of these designs established the strain range and temprature at critical locations resulting in predicted lives of 10,000 cycles for MA 956 alloy. The ODs alloys, MA 956 and HDA 8077, demonstrated a 167 C (300 F) temperature advantage over Hastelloy X alloy in creep strength and oxidation resistance. The MA 956 alloy was selected for mechanical property and component test evaluations. The MA 956 alloy was superior to Hastelloy X in LCF component testing of the film cooled, segmented louver design.

BRAZING ALLOYS

DOEpatents

Donnelly, R.G.; Gilliland, R.G.; Slaughter, G.M.

1963-02-26

A brazing alloy which, in the molten state, is characterized by excellent wettability and flowability, said alloy being capable of forming a corrosion resistant brazed joint wherein at least one component of said joint is graphite and the other component is a corrosion resistant refractory metal, said alloy consisting essentially of 20 to 50 per cent by weight of gold, 20 to 50 per cent by weight of nickel, and 15 to 45 per cent by weight of molybdenum. (AEC)

Zirconium determination by cooling curve analysis during the pyroprocessing of used nuclear fuel

NASA Astrophysics Data System (ADS)

Westphal, B. R.; Price, J. C.; Bateman, K. J.; Marsden, K. C.

2015-02-01

An alternative method to sampling and chemical analyses has been developed to monitor the concentration of zirconium in real-time during the casting of uranium products from the pyroprocessing of used nuclear fuel. The method utilizes the solidification characteristics of the uranium products to determine zirconium levels based on standard cooling curve analyses and established binary phase diagram data. Numerous uranium products have been analyzed for their zirconium content and compared against measured zirconium data. From this data, the following equation was derived for the zirconium content of uranium products:

Zirconium Recycle Test Equipment for Hot Cell Operations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collins, Emory D.; DelCul, Guillermo Daniel; Spencer, Barry B.

2015-01-30

The equipment components and assembly support work were modified for optimized, remote hot cell operations to complete this milestone. The modifications include installation of a charging door, Swagelok connector for the off-gas line between the reactor and condenser, and slide valve installation to permit attachment/replacement of the product salt collector bottle.

Dislocation structure in textured zirconium tensile-deformed along rolling and transverse directions determined by X-ray diffraction line profile analysis

NASA Astrophysics Data System (ADS)

Fan, Zhijian; Jóni, Bertalan; Xie, Lei; Ribárik, Gábor; Ungár, Tamás

2018-04-01

Specimens of cold-rolled zirconium were tensile-deformed along the rolling (RD) and the transverse (TD) directions. The stress-strain curves revealed a strong texture dependence. High resolution X-ray line profile analysis was used to determine the prevailing active slip-systems in the specimens with different textures. The reflections in the X-ray diffraction patterns were separated into two groups. One group corresponds to the major and the other group to the random texture component, respectively. The dislocation densities, the subgrain size and the prevailing active slip-systems were evaluated by using the convolutional multiple whole profile (CMWP) procedure. These microstructure parameters were evaluated separately in the two groups of reflections corresponding to the two different texture components. Significant differences were found in both, the evolution of dislocation densities and the development of the fractions of and type slip systems in the RD and TD specimens during tensile deformation. The differences between the RD and TD stress-strain curves are discussed in terms of the differences of the microstructure evolution.

Compressive Creep Behaviour of Extruded Mg Alloys at 150 °C

NASA Astrophysics Data System (ADS)

Fletcher, M.; Bichler, L.; Sediako, D.; Klassen, R.

Wrought magnesium alloy bars, sections and tubes have been extensively used in the aerospace, electronics and automotive industries, where component weight is of concern. The operating temperature of these components is typically limited to below 100°C, since appreciable creep relaxation of the wrought alloys takes place above this temperature.

Spall fracture and strength of uranium, plutonium and their alloys under shock wave loading

NASA Astrophysics Data System (ADS)

Golubev, Vladimir

2015-06-01

Numerous results on studying the spall fracture phenomenon of uranium, two its alloys with molybdenum and zirconium, plutonium and its alloy with gallium under shock wave loading are presented in the paper. The majority of tests were conducted with the samples in the form of disks 4mm in thickness. They were loaded by the impact of aluminum plates 4mm thick through a copper screen serving as the cover or bottom part of a special container. The initial temperature of samples was changed in the range of -196 - 800 C degree for uranium and 40 - 315 C degree for plutonium. The character of spall failure of materials and the degree of damage for all tested samples were observed on the longitudinal metallographic sections of recovered samples. For a concrete test temperature, the impact velocity was sequentially changed and therefore the loading conditions corresponding to the consecutive transition from microdamage nucleation up to complete macroscopic spall fracture were determined. Numerical calculations of the conditions of shock wave loading and spall fracture of samples were performed in the elastoplastic approach. Several two- and three-dimensional effects of loading were taken into account. Some results obtained under conditions of intensive impulse irradiation and intensive explosive loading are presented too. The rather complete analysis and comparison of obtained results with the data of other researchers on the spall fracture of examined materials were conducted.

An investigation of the oxidation behaviour of zirconium alloys using isotopic tracers and high resolution SIMS

NASA Astrophysics Data System (ADS)

Yardley, Sean S.; Moore, Katie L.; Ni, Na; Wei, Jang Fei; Lyon, Stuart; Preuss, Michael; Lozano-Perez, Sergio; Grovenor, Chris R. M.

2013-11-01

High resolution secondary ion mass spectrometry (SIMS) analysis has been used to study the oxidation mechanisms when commercial low tin ZIRLO™Low tin ZIRLO™ is a trademark of Westinghouse Electric Company LLC in the United States and may be registered in other countries throughout the world. Unauthorized use is strictly prohibited.1 and Zircaloy 4 materials are exposed to corroding environments containing both 18O and 2H isotopes. Clear evidence has been shown for different characteristic distributions of 18O before and after the kinetic transitions, and this behaviour has been correlated with the development of porosity in the oxide which allows the corroding medium to penetrate locally to the metal/oxide interface.

Electrical insulator assembly with oxygen permeation barrier

DOEpatents

Van Der Beck, Roland R.; Bond, James A.

1994-01-01

A high-voltage electrical insulator (21) for electrically insulating a thermoelectric module (17) in a spacecraft from a niobium-1% zirconium alloy wall (11) of a heat exchanger (13) filled with liquid lithium (16) while providing good thermal conductivity between the heat exchanger and the thermoelectric module. The insulator (21) has a single crystal alumina layer (SxAl.sub.2 O.sub.3, sapphire) with a niobium foil layer (32) bonded thereto on the surface of the alumina crystal (26) facing the heat exchanger wall (11), and a molybdenum layer (31) bonded to the niobium layer (32) to act as an oxygen permeation barrier to preclude the oxygen depleting effects of the lithium from causing undesirable niobium-aluminum intermetallic layers near the alumina-niobium interface.

Dynamic decoupling and local atomic order of a model multicomponent metallic glass-former.

PubMed

Kim, Jeongmin; Sung, Bong June

2015-06-17

The dynamics of multicomponent metallic alloys is spatially heterogeneous near glass transition. The diffusion coefficient of one component of the metallic alloys may also decouple from those of other components, i.e., the diffusion coefficient of each component depends differently on the viscosity of metallic alloys. In this work we investigate the dynamic heterogeneity and decoupling of a model system for multicomponent Pd43Cu27Ni10P20 melts by using a hard sphere model that considers the size disparity of alloys but does not take chemical effects into account. We also study how such dynamic behaviors would relate to the local atomic structure of metallic alloys. We find, from molecular dynamics simulations, that the smallest component P of multicomponent Pd43Cu27Ni10P20 melts becomes dynamically heterogeneous at a translational relaxation time scale and that the largest major component Pd forms a slow subsystem, which has been considered mainly responsible for the stabilization of amorphous state of alloys. The heterogeneous dynamics of P atoms accounts for the breakdown of Stokes-Einstein relation and also leads to the dynamic decoupling of P and Pd atoms. The dynamically heterogeneous P atoms decrease the lifetime of the local short-range atomic orders of both icosahedral and close-packed structures by orders of magnitude.

Alloy 740H Component Manufacturing Development

NASA Astrophysics Data System (ADS)

de Barbadillo, J. J.; Baker, B. A.; Gollihue, R. D.; Patel, S. J.

Alloy 740H was developed specifically for use in A-USC power plants. This alloy has been intensively evaluated in collaborative programs throughout the world, and the key properties have been verified and documented. In 2011 the alloy was approved for use in welded construction under ASME Code Case 2702. At present, alloy 740H is the only age-hardened nickel-base alloy that is ASME code approved. The emphasis for A-USC materials development is now on verification of the metalworking industry's capability to make the full range of mill product forms and sizes and to produce fittings and fabrications required for construction of a power plant. This paper presents the results of recent developments in component manufacture and evaluation.

40 CFR 721.9973 - Zirconium dichlorides (generic).

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 30 2010-07-01 2010-07-01 false Zirconium dichlorides (generic). 721... Substances § 721.9973 Zirconium dichlorides (generic). (a) Chemical substance and significant new uses subject to reporting. (1) The chemical substances identified generically as zirconium dichlorides (PMNs P...

Nuclear-grade zirconium prepared by combining combustion synthesis with molten-salt electrorefining technique

NASA Astrophysics Data System (ADS)

Li, Hui; Nersisyan, Hayk H.; Park, Kyung-Tae; Park, Sung-Bin; Kim, Jeong-Guk; Lee, Jeong-Min; Lee, Jong-Hyeon

2011-06-01

Zirconium has a low absorption cross-section for neutrons, which makes it an ideal material for use in nuclear reactor applications. However, hafnium typically contained in zirconium causes it to be far less useful for nuclear reactor materials because of its high neutron-absorbing properties. In the present study, a novel effective method has been developed for the production of hafnium-free zirconium. The process includes two main stages: magnesio-thermic reduction of ZrSiO 4 under a combustion mode, to produce zirconium silicide (ZrSi), and recovery of hafnium-free zirconium by molten-salt electrorefining. It was found that, depending on the electrorefining procedure, it is possible to produce zirconium powder with a low hafnium content: 70 ppm, determined by ICP-AES analysis.

Increasing Ti-6Al-4V brazed joint strength equal to the base metal by Ti and Zr amorphous filler alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ganjeh, E., E-mail: navidganjehie@sina.kntu.ac.ir; Sarkhosh, H.; Bajgholi, M.E.

Microstructural features developed along with mechanical properties in furnace brazing of Ti-6Al-4V alloy using STEMET 1228 (Ti-26.8Zr-13Ni-13.9Cu, wt.%) and STEMET 1406 (Zr-9.7Ti-12.4Ni-11.2Cu, wt.%) amorphous filler alloys. Brazing temperatures employed were 900-950 Degree-Sign C for the titanium-based filler and 900-990 Degree-Sign C for the zirconium-based filler alloys, respectively. The brazing time durations were 600, 1200 and 1800 s. The brazed joints were evaluated by ultrasonic test, and their microstructures and phase constitutions analyzed by metallography, scanning electron microscopy and X-ray diffraction analysis. Since microstructural evolution across the furnace brazed joints primarily depends on their alloying elements such as Cu, Ni andmore » Zr along the joint. Accordingly, existence of Zr{sub 2}Cu, Ti{sub 2}Cu and (Ti,Zr){sub 2}Ni intermetallic compounds was identified in the brazed joints. The chemical composition of segregation region in the center of brazed joints was identical to virgin filler alloy content which greatly deteriorated the shear strength of the joints. Adequate brazing time (1800 s) and/or temperature (950 Degree-Sign C for Ti-based and 990 Degree-Sign C for Zr-based) resulted in an acicular Widmanstaetten microstructure throughout the entire joint section due to eutectoid reaction. This microstructure increased the shear strength of the brazed joints up to the Ti-6Al-4V tensile strength level. Consequently, Ti-6Al-4V can be furnace brazed by Ti and Zr base foils produced excellent joint strengths. - Highlights: Black-Right-Pointing-Pointer Temperature or time was the main factors of controlling braze joint strength. Black-Right-Pointing-Pointer Developing a Widmanstaetten microstructure generates equal strength to base metal. Black-Right-Pointing-Pointer Brittle intermetallic compounds like (Ti,Zr){sub 2}Ni/Cu deteriorate shear strength. Black-Right-Pointing-Pointer Ti and Zr base filler alloys were the best choice for brazing Ti-6Al-4V.« less

Method for making fine and ultrafine spherical particles of zirconium titanate and other mixed metal oxide systems

DOEpatents

Hu, Michael Z.

2006-05-23

Disclosed is a method for making amorphous spherical particles of zirconium titanate and crystalline spherical particles of zirconium titanate comprising the steps of mixing an aqueous solution of zirconium salt and an aqueous solution of titanium salt into a mixed solution having equal moles of zirconium and titanium and having a total salt concentration in the range from 0.01 M to about 0.5 M. A stearic dispersant and an organic solvent is added to the mixed salt solution, subjecting the zirconium salt and the titanium salt in the mixed solution to a coprecipitation reaction forming a solution containing amorphous spherical particles of zirconium titanate wherein the volume ratio of the organic solvent to aqueous part is in the range from 1 to 5. The solution of amorphous spherical particles is incubated in an oven at a temperature .ltoreq.100.degree. C. for a period of time .ltoreq.24 hours converting the amorphous particles to fine or ultrafine crystalline spherical particles of zirconium titanate.

40 CFR 471.90 - Applicability; description of the zirconium-hafnium forming subcategory.

Code of Federal Regulations, 2010 CFR

2010-07-01

... zirconium-hafnium forming subcategory. 471.90 Section 471.90 Protection of Environment ENVIRONMENTAL... POINT SOURCE CATEGORY Zirconium-Hafnium Forming Subcategory § 471.90 Applicability; description of the zirconium-hafnium forming subcategory. This subpart applies to discharges of pollutants to waters of the...

40 CFR 421.330 - Applicability: Description of the primary zirconium and hafnium subcategory.

Code of Federal Regulations, 2010 CFR

2010-07-01

... primary zirconium and hafnium subcategory. 421.330 Section 421.330 Protection of Environment ENVIRONMENTAL... CATEGORY Primary Zirconium and Hafnium Subcategory § 421.330 Applicability: Description of the primary zirconium and hafnium subcategory. The provisions of this subpart are applicable to discharges resulting...

Ablation Resistant Zirconium and Hafnium Ceramics

NASA Technical Reports Server (NTRS)

Bull, Jeffrey (Inventor); White, Michael J. (Inventor); Kaufman, Larry (Inventor)

1998-01-01

High temperature ablation resistant ceramic composites have been made. These ceramics are composites of zirconium diboride and zirconium carbide with silicon carbide, hafnium diboride and hafnium carbide with silicon carbide and ceramic composites which contain mixed diborides and/or carbides of zirconium and hafnium. along with silicon carbide.

Optimization of stress relief heat treatment of PHWR pressure tubes (Zr 2.5Nb alloy)

NASA Astrophysics Data System (ADS)

Choudhuri, Gargi; Srivastava, D.; Gurumurthy, K. R.; Shah, B. K.

2008-12-01

The micro-structure of cold worked Zr-2.5%Nb pressure tube material consists of elongated grains of α-zirconium enclosed by a thin film of β-zirconium phase. This β-Zr phase is unstable and on heating, progressively decomposes to α-Zr phase and β-phase enriched with Nb and ultimately form β Nb. Meta-stable ω-phase precipitates as an intermediate step during decomposition depending on the heat treatment schedule, β→α+β→α+ω+β→α+β→α+β Morphological changes occur in the β-zirconium phase during the decomposition. The continuous ligaments of β Zr phase turn into a discontinuous array of particles followed by globulization of the β-phase. The morphological changes impose a significant effect on the creep rate and on the delayed hydride cracking velocity due to reduction in the hydrogen diffusion coefficient in α Zr. If the continuity of β-phase is disrupted by heat treatment, the effective diffusion coefficient decreases with a concomitant reduction in DHC velocity. The pressure tubes for the Indian PHWRs are made by a process of hot extrusion followed by cold pilgering in two stages and an intermediate annealing. Autoclaving at 400 °C for 36 h ensures stress relieving of the finished tubes. In the present studies, autoclaving duration at 400 °C was varied from 24 h to 96 h at 12 h-steps and the micro-structural changes in the β-phase were observed by TEM. Dislocation density, hardness and the micro-structural features such as thickness of β-phase, inter-particle spacing and volume fraction of the phases were measured at each stage. Autoclaving for a longer duration was found to change the morphology of β-phase and increase the inter-particle spacing. Progressive changes in the aspect ratio of the β-phase and their size and distribution are documented and reported. These micro-structural modifications are expected to decrease DHC velocity during reactor operation.

Criteria for predicting the formation of single-phase high-entropy alloys

DOE PAGES

Troparevsky, M Claudia; Morris, James R..; Kent, Paul R.; ...

2015-03-15

High entropy alloys constitute a new class of materials whose very existence poses fundamental questions. Originally thought to be stabilized by the large entropy of mixing, these alloys have attracted attention due to their potential applications, yet no model capable of robustly predicting which combinations of elements will form a single-phase currently exists. Here we propose a model that, through the use of high-throughput computation of the enthalpies of formation of binary compounds, is able to confirm all known high-entropy alloys while rejecting similar alloys that are known to form multiple phases. Despite the increasing entropy, our model predicts thatmore » the number of potential single-phase multicomponent alloys decreases with an increasing number of components: out of more than two million possible 7-component alloys considered, fewer than twenty single-phase alloys are likely.« less

Multilayer bioactive glass/zirconium titanate thin films in bone tissue engineering and regenerative dentistry

PubMed Central

Mozafari, Masoud; Salahinejad, Erfan; Shabafrooz, Vahid; Yazdimamaghani, Mostafa; Vashaee, Daryoosh; Tayebi, Lobat

2013-01-01

Surface modification, particularly coatings deposition, is beneficial to tissue-engineering applications. In this work, bioactive glass/zirconium titanate composite thin films were prepared by a sol-gel spin-coating method. The surface features of the coatings were studied by scanning electron microscopy, atomic force microscopy, and spectroscopic reflection analyses. The results show that uniform and sound multilayer thin films were successfully prepared through the optimization of the process variables and the application of carboxymethyl cellulose as a dispersing agent. Also, it was found that the thickness and roughness of the multilayer coatings increase nonlinearly with increasing the number of the layers. This new class of nanocomposite coatings, comprising the bioactive and inert components, is expected not only to enhance bioactivity and biocompatibility, but also to protect the surface of metallic implants against wear and corrosion. PMID:23641155

Preparation of monotectic alloys having a controlled microstructure by directional solidification under dopant-induced interface breakdown

NASA Technical Reports Server (NTRS)

Parr, R. A.; Johnston, M. H.; Mcclure, J. C.

1980-01-01

Monotectic alloys having aligned spherical particles of rods of the minor component dispersed in a matrix of the major component are prepared by forming a melt containing predetermined amounts of the major and minor components of a chosen monotectic system, providing in the melt a dopant capable of breaking down the liquid solid interface for the chosen alloy, and directionally solidfying the melt at a selected temperature gradient and a selected rate of movement of the liquid-solid interface (growth rate). Shaping of the minor component into spheres or rods and the spacing between them are controlled by the amount of dopant and the temperature gradient and growth rate values. Specific alloy systems include Al Bi, Al Pb and Zn Bi, using a transition element such as iron.

Fatigue Crack Growth Threshold Testing of Metallic Rotorcraft Materials

NASA Technical Reports Server (NTRS)

Newman, John A.; James, Mark A.; Johnson, William M.; Le, Dy D.

2008-01-01

Results are presented for a program to determine the near-threshold fatigue crack growth behavior appropriate for metallic rotorcraft alloys. Four alloys, all commonly used in the manufacture of rotorcraft, were selected for study: Aluminum alloy 7050, 4340 steel, AZ91E Magnesium, and Titanium alloy Ti-6Al-4V (beta-STOA). The Federal Aviation Administration (FAA) sponsored this research to advance efforts to incorporate damage tolerance design and analysis as requirements for rotorcraft certification. Rotorcraft components are subjected to high cycle fatigue and are typically subjected to higher stresses and more stress cycles per flight hour than fixed-wing aircraft components. Fatigue lives of rotorcraft components are generally spent initiating small fatigue cracks that propagate slowly under near-threshold cracktip loading conditions. For these components, the fatigue life is very sensitive to the near-threshold characteristics of the material.

40 CFR 721.10089 - Modified salicylic acid, zirconium complex (generic).

Code of Federal Regulations, 2011 CFR

2011-07-01

... 40 Protection of Environment 31 2011-07-01 2011-07-01 false Modified salicylic acid, zirconium... Specific Chemical Substances § 721.10089 Modified salicylic acid, zirconium complex (generic). (a) Chemical... as modified salicylic acid, zirconium complex (PMN P-00-552) is subject to reporting under this...

40 CFR 721.10089 - Modified salicylic acid, zirconium complex (generic).

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 30 2010-07-01 2010-07-01 false Modified salicylic acid, zirconium... Specific Chemical Substances § 721.10089 Modified salicylic acid, zirconium complex (generic). (a) Chemical... as modified salicylic acid, zirconium complex (PMN P-00-552) is subject to reporting under this...

Surface composition of alloys

NASA Astrophysics Data System (ADS)

Sachtler, W. M. H.

1984-11-01

In equilibrium, the composition of the surface of an alloy will, in general, differ from that of the bulk. The broken-bond model is applicable to alloys with atoms of virtually equal size. If the heat of alloy formation is zero, the component of lower heat of atomization is found enriched in the surface. If both partners have equal heats of sublimination, the surface of a diluted alloy is enriched with the minority component. Size effects can enhance or weaken the electronic effects. In general, lattice strain can be relaxed by precipitating atoms of deviating size on the surface. Two-phase alloys are described by the "cherry model", i.e. one alloy phase, the "kernel" is surrounded by another alloy, the "flesh", and the surface of the outer phase, the "skin" displays a deviating surface composition as in monophasic alloys. In the presence of molecules capable of forming chemical bonds with individual metal atoms, "chemisorption induced surface segregation" can be observed at low temperatures, i.e. the surface becomes enriched with the metal forming the stronger chemisorption bonds.

Anticorrosive Behavior and Porosity of Tricationic Phosphate and Zirconium Conversion Coating on Galvanized Steel

NASA Astrophysics Data System (ADS)

Velasquez, Camilo S.; Pimenta, Egnalda P. S.; Lins, Vanessa F. C.

2018-05-01

This work evaluates the corrosion resistance of galvanized steel treated with tricationic phosphate and zirconium conversion coating after painting, by using electrochemical techniques, accelerated and field corrosion tests. A non-uniform and heterogeneous distribution of zirconium on the steel surface was observed due to preferential nucleation of the zirconium on the aluminum-rich sites on the surface of galvanized steel. The long-term anti-corrosion performance in a saline solution was better for the phosphate coating up to 120 days. The coating capacitance registered a higher increase for the zirconium coatings than the phosphate coatings up to 120 days of immersion. This result agrees with the higher porosity of zirconium coating in relation to the phosphate coating. After 3840 h of accelerated corrosion test, and after 1 year of accelerated field test, zirconium-treated samples showed an average scribe delamination length higher than the phosphate-treated samples.

Anticorrosive Behavior and Porosity of Tricationic Phosphate and Zirconium Conversion Coating on Galvanized Steel

NASA Astrophysics Data System (ADS)

Velasquez, Camilo S.; Pimenta, Egnalda P. S.; Lins, Vanessa F. C.

2018-04-01

This work evaluates the corrosion resistance of galvanized steel treated with tricationic phosphate and zirconium conversion coating after painting, by using electrochemical techniques, accelerated and field corrosion tests. A non-uniform and heterogeneous distribution of zirconium on the steel surface was observed due to preferential nucleation of the zirconium on the aluminum-rich sites on the surface of galvanized steel. The long-term anti-corrosion performance in a saline solution was better for the phosphate coating up to 120 days. The coating capacitance registered a higher increase for the zirconium coatings than the phosphate coatings up to 120 days of immersion. This result agrees with the higher porosity of zirconium coating in relation to the phosphate coating. After 3840 h of accelerated corrosion test, and after 1 year of accelerated field test, zirconium-treated samples showed an average scribe delamination length higher than the phosphate-treated samples.

Analysis of the influence of the macro- and microstructure of dental zirconium implants on osseointegration: a minipig study.

PubMed

Mueller, Cornelia Katharina; Solcher, Philipp; Peisker, Andrè; Mtsariashvilli, Maia; Schlegel, Karl Andreas; Hildebrand, Gerhard; Rost, Juergen; Liefeith, Klaus; Chen, Jiang; Schultze-Mosgau, Stefan

2013-07-01

It was the aim of this study to analyze the influence of implant design and surface topography on the osseointegration of dental zirconium implants. Six different implant designs were tested in the study. Nine or 10 test implants were inserted in the frontal skull in each of 10 miniature pigs. Biopsies were harvested after 2 and 4 months and subjected to microradiography. No significant differences between titanium and zirconium were found regarding the microradiographically detected bone-implant contact (BIC). Cylindric zirconium implants showed a higher BIC at the 2-month follow-up than conic zirconium implants. Among zirconium implants, those with an intermediate Ra value showed a significantly higher BIC compared with low and high Ra implants 4 months after surgery. Regarding osseointegration, titanium and zirconium showed equal properties. Cylindric implant design and intermediate surface roughness seemed to enhance osseointegration. Copyright © 2013 Elsevier Inc. All rights reserved.